ACCN
        TYPE
        DB
        MTI
        TRACED
        MYDATE
        CALLNO
        BRANCH
        LOCATION
1   3712
2   CAT
3   ELAL                        	—
4   Radiation  safety at  superfund  sites
5   Environmental  Response  Training Program
6   0094
7   6886-9
8   back
9   elad
                                   FOREWORD
This manual is for reference use of participants enrolled in scheduled training courses of the
U.S. Environmental Protection Agency (EPA).  While it will be useful to anyone who needs
information on the subjects covered,  it will have its greatest value as an adjunct to classroom
presentations involving discussion among the participants and the instructional staff.

Individual instructors may provide  additional  materials to cover special aspects  of their
presentations.

Because of the limited availability of the manual, it should not be cited in bibliographies or other
publications.

References  to products and  manufacturers are for illustration  only;  they  do  not imply
endorsement by EPA.

Constructive suggestions for improving content and format of the manual are welcome.
                                                    U.S. Environmental Protection Agency
                                                    Region 5, Library 'PL-12J)
                                                    77 West Jackson Boulevard, 12th Floor
                                                    Chicago, IL  60604-359(k

-------
         RADIATION SAFETY AT SUPERFUND SITES
                               (165.11)

                                5 Days
This is a basic course in radiation safety for individuals who, in the course of
their work, become involved with Superfund sites that have radioactive material
concerns.

This course is designed to provide participants with an understanding of the
fundamental principles of radiation safety, with emphasis on radiation detection
and contamination control.

Upon completion of this course, participants will be able to:

      •      Discuss  fundamental  concepts  of  atomic   structure,
             radiation, and radioactive decay

      •      Identify the biological effects of radiation exposure and the
             existing rules and regulations that establish the protection
             criteria for exposure

      •      Discuss  radiation  detection,  including  the  theory of
             operation, use, and selection  of radiation monitoring
             instruments

      •      Conduct  radiation  surveys  using  proper methods  and
             techniques

      •      Discuss  contamination  surveys  and   the  setup  of
             contaminated areas

      •      Review regulations regarding the transport of radioactive
             material

      •      Discuss radioactive waste disposal and remedial options for
             radioactive cleanup.
          U.S. ENVIRONMENTAL PROTECTION AGENCY
               Office of Emergency and Remedial Response
                     Environmental Response Team

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             RADIATION SAFETY AT SUPERFUND SITES (165.11)




                                      CONTENTS






                                                                                 Section




LECTURES




       Atomic Structure and Radioactive Decay	  1




       Interaction of Radiation with Matter  	  2




       Radiation Exposure and Biological Effects	  3




       Radiation Exposure Limits and Methods to Control Exposure	  4




       Basic Concepts in Radiation Detection and Measurement	  5




       Radiation Detection Instruments	  6




       Surveying for Radioactive Materials  	  7




       Radiation Signs and Labels	  8




       Contamination Control	  9




       Anti-contamination Clothing and Respiratory Protection Devices	10




       Demonstration:  Radiological Control Area	11




       Decontamination  	12




       Radioactive Material Packaging, Labeling, and Shipping	13




       Radioactive Soil and Water Sampling	14




       Regulations and Guidance on Radioactive Waste Disposal	15




       Remedial and Disposal Options  	16




REFERENCES	17




       Glossary




       Excerpt, Radiological Health Handbook, January 1970




       10 CFR 20—Standards for Protection Against Radiation




       OSHA 1910.96—Ionizing Radiation

-------
CONTENTS (cont.)

       Federal Radiation Protection Guidance for Occupational Exposure

       Regulatory Guide 1.86—Termination of Operating Licenses for Nuclear Reactors

       Policy and Guidance Directive FC 83-23

       Regulatory Guide 8.13—Instruction Concerning Prenatal Radiation Exposure

       Fundamentals of Health Physics for the Radiation Protection Officer. September 1983. Darcom
       P385-1.  Prepared  for the Department of the Army, U.S. Army Materiel Development and
       Readiness Command,  Alexandria,  Virginia.    Pacific Northwest Laboratory,  Richland,
       Washington.

       A Review of the Department of Transportation Regulations for Transportation of Radioactive
       Materials

       49 CFR  173.403-Definitions

       U.S. Nuclear Regulatory Commission and Agreement State Offices, September 9, 1990

       Suppliers
                                             VI

-------
                        SECTION 1

  ATOMIC STRUCTURE AND RADIOACTIVE DECAY
         After completing this unit, participants will be able to:

         •    Describe the theoretical structure of the atom and the
              chemical notation system.

         •    Define radioactive decay and explain its cause.

         •    Characterize the types of radiation emissions as a result
              of radioactive decay.
5/93

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ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                                                                NOTES
I.
Atomic Structure

A.    Nucleus
             -positively charged central portion of an
             atom that comprises nearly all of the
             atomic mass and that consists of protons
             and neutrons

             1.     Proton
                    -positively  charged
                    -mass of 1  AMU (1.673 x  10'24
                     gram) (1  AMU = 1/12 the mass
                     of a carbon-12 atom)
                    -determines the element  (Figure 1)

             2.     Neutron
                    -no charge
                    -mass of 1  AMU
                    -determines the isotope
                    (hydrogen-1, hydrogen-2,
                    hydrogen-3)

B.    Electrons
             -negatively charged particles that orbit
             the nucleus and comprise nearly all of the
             volume of an atom.  These can be thought
             of as a "cloud" around the nucleus.

             -negatively charged
                    (1.602  x 10-19 coulomb)

             -mass of 1/1832 AMU
                    (9.109534  x 10-28gram)

             -the electrons can also be thought of as
              filling discrete orbital shells around the
              nucleus.  Each shell represents an energy
              level. This is called the Bohr  model and
              each shell is identified by a letter  (K, L,
              M, N....)
 5/93

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ATOMIC STRUCTURE AND RADIOACTIVE DECAY
NOTES
      C.    Chemical Notation

            1.    Element Symbol (X) - indicates what basic
                  element the atom is (Figure 1)

            2.    Proton Number (Z) - number of protons
                  within the nucleus

            3.    Neutron Number (N) - number of neutrons
                  within the nucleus

            4.    Mass Number (A) - represents the total
                  mass of the atom. It is the sum of the
                  proton number and the neutron number (A
                   = Z + N).
             5.     Convention
                   Examples:   "c = 12C = C-12
                      (see Figure 1)
       D.    Definitions
             1.     Isotope - atoms of one element that have
                   the same atomic number but differ in
                   neutron  number

             2.     Nuclide - a species of atom characterized
                   by the constitution of its nucleus

             3.     Radionuclide - a radioactive nuclide
                   Example:      ^    2    g
                                 j#,  j//,  ^H
  5/93

-------
ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                                                                            NOTES
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                             FIGURE 1

              PERIODIC TABLE OF THE ELEMENTS
5/93

-------
ATOMIC STRUCTURE AND RADIOACTIVE DECAY
NOTES
II.    Radioactivity
       A.    Radioactive Decay - the disintegration of the
             nucleus of an unstable nuclide by spontaneous
             emission of particles or photons

       B.    Types of Unstable Nuclides

             1.     Excess binding energy in the nucleus

             2.     Unstable pro ton'.neutron ratio

       C.    Mechanism of Radioactive Decay

             1.     Particle Emission
                   a.     Alpha
                   b.     Beta

             2.     Electron Capture

             3.     Photon Emission (Gamma)

       D.    Definitions

             1.     Radioactivity - the property of
                   spontaneously emitting particles or photons

             2.     Natural Radioactivity - radioactivity
                   exhibited by more than 50 naturally
                   occurring radionuclides

             3.     Artificial Radioactivity - man-made
                   radioactivity produced by particle
                   bombardment or electromagnetic
                   irradiation

             4.     Induced Radioactivity  - radioactivity
                   produced in a substance after
                   bombardment with particles
5/93

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ATOMIC STRUCTURE AND RADIOACTIVE DECAY
NOTES
III.   Radiation
      A.     Definitions

             1.     Radiation - energy emitted in the form of
                   waves or particles

             2.     Ionizing Radiation - radiation that has a
                   high enough energy level to strip electrons
                   from atoms

      B.     Types of Radiation

             1.     Particles - alpha, beta, neutron

             2.     Electromagnetic ray - gamma, x-ray

      C.     Characteristics  of Each Radiation

             1.     Alpha

                   a.    Make up

                         -2 protons and 2 neutrons
                         -mass = 4 AMUs
                         -charge = +2
                         -energy level = 4-7 MeV

                           kinetic energy = — m0v2

                         - expressed  in units of electron
                           volts (eV)

                   b.    Emitted from very heavy nuclei

                           Z > 82 such as gg^w

                   c.    Decay scheme

                               F~* 17 f\ * ~*~ f^flC
                                £s~Z,    £
5/93

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ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                                                       NOTES
            2.     Beta
                   a.     Make up
                         -electron or positron
                         -mass = 1/1832 AMU
                         -charge =  +1 or —1
                         -energy level =  .1 - 2.5 MeV
kinetic energy = — r
                                            -1
                         - expressed in units of electron
                          volts (eV)

                   b.     Emission is due to an unstable
                         proton:neutron ratio

                   c.     Decay schemes

                         -Normal beta decay

                          ~N - ,P + B' ejected

                         -Positron decay

                          \P - \N + 5+ ejected

                         - Electron capture
                          1
                           P + e- orbital
5/93
             3.     Gamma Ray

                   a.     Make up
                         -electromagnetic ray (pure energy)
                         -no mass
                         -no charge
                         -energy level measured in units of
                          electron volts (eV)

-------
ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                                                         NOTES
                          Emission is from a nucleus that has
                          been left in an excited state
                          following a particle emission or
                          capture

                          The gamma is usually emitted
                          immediately after particle ejection
                          but may be several hours later
IV.    Decay Pathways

       A.    Definitions

             1.
       B.
       C.
             2.
      Radioactive Decay - the disintegration of
      the nucleus of an unstable nuclide by
      spontaneous emission of particles or
      photons

      Decay Pathway - consists  of particle
      emission or capture, possibly followed by
      one or more gamma rays
Notes of Interest

1.
                    Some radionuclides can decay via more
                    than one decay pathway
2.    The pathways differ with the energy of
      decay and subsequent gamma rays

3.    The number and energy of the emissions
      are characteristic of a given radionuclide

4.    A single nucleus can decay by only one of
      the various pathways

Rate of Decay

1.    Half-Life (T 1/2) - that amount of time
      required for one-half of a radionuclide
      population to decay away

2.    Decay Constant (X)

            In2
        A — 	
5/93

-------
ATOMIC STRUCTURE AND RADIOACTIVE DECAY
NOTES
                   See Figures 2 and 3
                12.3 y
                          He
              (MTW)
              5730 y
                          N
                            2.60 y
 TRITIUM, CARBON, AND SODIUM DECAY SCHEME
                  FIGURE 2
5/93

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ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                                NOTES
                           3.82d
                                 222
                                    Rn
                                           226
     1602y
                                    ^(4.60 MeV (5%)
                                    a \4.78 MeV (95%)

                                    y 10.19 MeV (4% + 1C)
                                    '«( 5.49 MeV (100%)
                                    CM 6.00 MeV (100%)
                                      , 0.69 MeV (~47%)
                                    /3 | 0.74 MeV (44%)
                                      (1.03 MeV (6%)
                                       0.05 MeV (1%-f 1C)
                                       0.24 MeV (4%)
                                       0.29 MeV (19%)
                                       0.35 MeV (36%)

                                       < 2 MeV (—76%)
                                       3.26 MeV (~19%)
                                       0.61 MeV (47%)
                                       0.77 MeV (5%)
                                       0.93 MeV (3%)
                                       1.12 MeV (17%)
                                       1.24 MeV (6%)
                                       1.3 8 MeV (5%)
                                       1.76 MeV (17%)
                                       2.20 MeV (5%)
                                       2.44 MeV (2%)
                                 (RaC1)

                                     a I 7.69 MeV (100%)

                                       0.01 MeV (81%)
                                     .ft
                                     P
                                        0.06 MeV (19%)

                                       I 0.05 MeV (4% 4 1C)

                                       1 1.1 6 MeV (100%)
                                138.4d
       Stable
                                     a  I 5.31 MeV (100%)
                RADIUM DECAY SCHEME
                          FIGURE 3
5/93
10

-------
ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                                                                NOTES
V.
Measuring Radioactive Material

Radioactive material is measured by counting the number
of disintegrations that occur over some period of time.
This is call its "Activity."

       Activity (A)  = disintegrations/time
                    (i.e., dps or dpm)

A.     Traditional Units

          1 curie (Ci)= 3.7 X  1010 dps
                    =  2.2 x 1012 dpm

B.     Subunits

       1 microcurie  frtCi) = 1  X 10"6 Ci
       1 picocurie (pCi)  =  1  x W12 Ci

C.     International System (SI)

           1 Becquerel (Bq) = 1 dps

D.     Activity Determination

                   Xt
      E.
              A = A0 e -

             where:
                   A = activity after time t
                   A0 = original activity
                   e = base of natural logarithm
                   (2.718)
                   X = decay constant (In2/half life)
                   t = f»1an«p/i tim«»
             t = elapsed time

Radioactive Contamination
             1.     Definition
                          Radioactive contamination is a fine
                          form of radioactive material in a
                          place that it is not wanted.
5/93
                                  11

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ATOMIC STRUCTURE AND RADIOACTIVE DECAY
                          NOTES
           2.    Radioactive Contamination Measurement
                 Applied Units

                 a.    Surface area - dpm/100 cm2

                 b.    Air -

                 c.    Water -

                 d.    Solids (specific activity) - Ci/g
5/93
12

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                       SECTION 2

    INTERACTION OF RADIATION WITH MATTER
         After completing this unit, participants will be able to:

         •   Describe the mechanisms by which alpha, beta, gamma,
             and neutron radiations interact with matter.
5/93

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INTERACTION OF RADIATION WITH MATTER
                                                                   NOTES
I.
Basic Interaction Results

A.     Excitation - the forced movement of an electron
       from an inner shell to some outer shell of an atom
       (Note:  when the electron jumps back to its
       original shell, a photon is emitted)
       B.
       D.
       lonization - the stripping of an orbital electron
       from an atom

       Secondary lonization - the ionization caused by a
       particle which is itself a result of ionization
       (sometimes termed "delta rays")

       Linear Energy Transfer - the amount of energy
       deposited by a specific radiation over a specific
       distance

                            dE,
                     LET = —'-
                             dl
       (Note: Ultimately, the energy transferred either
       to tissue or to any other material is dissipated as
       heat)
II.     Specific Radiation Interactions

       A.    Alpha - because of its high charge (+2) and large
             mass, it actually pulls electrons off as it goes past
             atoms or molecules
             - it has a very high LET (tens of thousands of
             ions per centimeter in air)
       B.    Beta
              1.
             Electron

             a.     lonization/Excitation - the single
                    negative charge and small mass
                    interacts by a repelling "collision,1
                    where the two electrons do not
                    actually touch
5/93

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INTERACTION OF RADIATION WITH MATTER
                                                     NOTES
                    b.
       The beta pushes the orbital electron
       out of its shell
       - it has a medium LET (hundreds
       of ions per centimeter in air)

       Bremsstrahlung (Braking Radiation)
       - x-rays are given off during the
       course of electron deceleration
       around the nucleus of an atom
             2.
Positron
                    a.     lonization/Excitation - single
                          positive charge and small mass
                          ionizes by attraction

                    b.     When the positron has slowed
                          down, it combines with an electron,
                          they annihilate one another,
                          generating two gamma rays with
                          energies equivalent to the mass of
                          the original particles (0.51 MeV
                          each)

             Gamma Rays

             1.      Photoelectric Effect

                    a.     The photon interacts with a tightly
                          bound electron and ejects it from its
                          orbit
                          i.      the photon disappears
                          ii.     ajl of the photon's energy
                                 goes into breaking the
                                 electron's bond and the
                                 resulting  kinetic energy of
                                 the electron. The electron
                                 then becomes the primary
                                 ionizer.

             2.      Compton Scattering

                    a.     Elastic collision  between a "free"
                          electron (one whose binding energy
5/93

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INTERACTION OF RADIATION WITH MATTER
NOTES
                           is far lower than the photon's
                           energy), where the electron gets
                           some kinetic energy and the photon
                           still exists but at a lower energy
                           (longer wavelength)

             3.     Pair Production - this is the primary
                    interaction for photons greater than 1.02
                    MeV

                    a.     the photon passes near a nucleus
                           and disappears

                    b.     an electron and a positron appear
                           and are attenuated in their normal
                           manner

             4.     Absorption (photodisintegration)

                    a.     the photon is captured in the
                           nucleus, which then emits a neutron

                    b.     usually a very high energy is
                           required for this (except Beryllium-
                           9). Only  a 1.666 MeV gamma is
                           required to give a 9Be(7,n)8Be
                           reaction.
 III.    Neutron Radiation
       A.     Neutron Characteristics
              1 .      Neutrons are not part of normal atom
                     decay schemes;  an interaction must occur
                     for a neutron to be ejected

              2.      mass = 1 AMU

              3.      charge = 0

              4.      energy level is expressed in units of
                     electron volts
 5/93

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INTERACTION OF RADIATION WITH MATTER
                                                    NOTES
       B.    Neutron-Generating Mechanisms
             1.
             2.
Nuclear Reaction - when either a uranium
or a plutonium atom is split, there are an
average of 2.54 neutrons generated

Cyclotron Bombardment - deuterons are
accelerated and bombarded into a
beryllium target
                    Alpha Bombardment - mix the fine
                    powders of beryllium and an alpha emitter
                    (radium, polonium, or plutonium) together
                    and seal it in a capsule

                     9     4      13  »  12    1
                    Photodisintegration - a photon is absorbed
                    by the nucleus and a neutron is ejected
             Classification of Neutrons

             1 .     Fast Neutron - a high kinetic energy
                   neutron with a KE > 0. 1 MeV - all
                   neutrons are fast upon generation

             2.     Thermal Neutrons - these have the same
                   average kinetic energy as the gas
                   molecules in their environment (which is
                   set by temperature)

             3.     Middle Range - this covers all of the area
                   between  fast and thermal neutrons.
                   Depending on the author, they could be
                   called Intermediate. Resonance., or Slow
                   Neutrons; these terms are used very
                   loosely.
5/93

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INTERACTION OF RADIATION WITH MATTER
                                             NOTES
      D.     Neutron Interaction with Matter
                   Scattering

                   a.
Inelastic scattering - this is actually
two steps:

i.     Neutron is captured by a
      nucleus

ii.    Neutron is re-emitted, at a
      lower energy, by that
      nucleus with a gamma
      photon
                   b.     Elastic scattering

                          i.     occurs between fast neutrons
                                and low atomic number
                                absorbers (such as
                                hydrogen-rich material like
                                water or a poly material)

                          ii.     This is a billiard ball type
                                collision

                   Absorption

                   a.     A thermal neutron can be captured
                          by a nucleus
                           1,
                           10B(n,a)7Li
 5/93

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INTERACTION OF RADIATION WITH MATTER
                           NOTES
                 b.     Activation is when the capture of a
                       thermal neutron results in the atom
                       becoming radioactive
                        50
                          0
51
  Cr
                                    60
                                      Co +
5/93

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                       SECTION 3

RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
         After completing this unit, participants will be able to:

         •    Define radiation dose and exposure.

         •    Define the units of measurement for radiation dose and
              exposure.

         •    Describe the toxicological or biological impacts due to
              radiation exposure.
5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
NOTES
4
I.      Radiation Exposure

       A.    Measurement Concepts

             1.     Exposure

                    The measurement of radiation passing
                    through air that an individual would be
                    subjected to if they were to stand in that
                    spot.  This is normally measured in units
                    of Roentgen (R).

             2.     Dose

                    The deposition of energy into soft tissue
                    (human body) by a specific form and
                    energy level of radiation.  This is normally
                    measured in units of Rad (an acronym for
                    Radiation Absorbed Dose).

             3.     Dose Equivalent

                    The estimation of the biological risk
                    associated with radiation exposure
                    regardless of the type of radiation or its
                    energy level. This is normally measured
                    in units of Rem.

       B.    Units of Measurement

             1.     Roentgen (R)

                    The quantity of x-ray or gamma radiation
                    producing one electrostatic unit of charge
                    in one cubic centimeter (cc)  of dry air at
                    standard temperature and pressure.

                    Notes:
                    - absorption of 1R in Ice of air results in
                     formation of 2.08 X  109 ion pairs.
                    - in terms of energy per unit mass of dry
                     air, this converts to 87.8 ergs/gram
5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                   NOTES
                   - when applied to muscle tissue (instead of
                     dry air), it leads to the absorption of 95
                     ergs/gram of muscle tissue.

                   Rad

                   A measure of the dose of any ionizing
                   radiation to body tissue in terms of the
                   energy absorbed per unit mass of tissue
                   (1 Rad = ^00 erg/gram of body tissue).

                   Rem (Roentgen Equivalent Man)

                   The amount of radiation that causes
                   damage equivalent to the damage  done by
                   the absorption of 100 ergs X (or 7)
                   radiation per gram of soft body tissue.

                   Quality Factor (QF)

                   QF accounts for the differences in the
                   biological effect of different types of
                   radiation as compared to that of X-
                   radiation.
                          Radiation
                          7
                          X
                             >  .03 MeV
                             <  .03 MeV
                          Nf
                          P
                          a
                          Heavy Ions
QE

1
1
1
1.7
3
10
10
20
20
       C.    Subunits
              1.     Milli (m) - one thousandth of the indicated
                    unit.
                    ex:    1 millirem = 10"3 rem
 5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                                                     NOTES
             2.     Micro (/x) - one millionth of the indicated
                    unit.
                    ex:    1  microrem = 10"6 rem
             3.     Kilo (k) - one thousand of the indicated
                    unit
                    ex:    1  keV  = 103 eV

             4.     Mega (M) - one million of the indicated
                    unit
                    ex:    1  MeV = 106 eV
II.     Biological Effects

       A.     Chemical Toxicity
              1.
              2.
              3.
Chemical toxicity is the harm that can be
caused by an element due to its chemical
nature.

Most chemicals have a level at which they
become toxic (poisonous).

Radioactive chemicals interact by the same
chemical interaction as their nonradioactive
counterparts.
       B.     Radio toxicity
              1.     Radioactivity indicates the relative
                    radiological hazard associated with
                    internally deposited radionuclides.

              2.     Alpha particles or high-energy beta
                    particles present the greatest hazard when
                    they are emitted internally.

              Interaction of Radiation with Tissue

              1.     Effects of damage to cells whether through
                    direct or indirect action of radiation are:
 5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                                                    NOTES
             2.
                    a. cell dies
                    b. cell lives normally
                    c. cell produces daughter cells that die
                    d. cell produces mutated daughter cells
Direct Action

a.      ~ 1 atom in 10 million is affected
      by a dose of 450 rads

      Proof:
                 is
            34-^
              ion
                         = 7.35 X 10
                                   "
                  g of tissue
                          i.
                          11.
             estimate 9 atoms excited for
             each 1 ionized

             7.35 x 1018 atoms/kg of
             tissue are directly affected.
             There are  -9.5  x  1025
             atoms/kg in soft tissue.
                                  1018
                                  1025
                     1
                     io7
                                 Note:  The reason that this
                                 is of interest is LD50  =
                                 450 rads

                          If the affected atom is on the DNA
                          molecule, genetic information may
                          not be transferred.

                          Germinal cells - mutation is passed
                          on to next individual

                          Somatic cells - mutation is passed
                          on to daughter cell
5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                                                         NOTES
       D.
             3.     Indirect Action

                   a.     Most of the body is water;
                          therefore, most of radiation's direct
                          action is  on water.
                              * - H* + OH
                           H2Q- - H +
             High LET radiation:

              OH + OH - H2O2

              H + H - H2

             If dissolved O2 is present:

              H

              H
                     i*     t,  Jr

Radiosensitivity

1.     Law of Bergonie and Tribondeau

      "The radiosensitivity of a tissue is directly
      proportional to its reproductive capacity
      and inversely  proportional to its degree of
      differentiation."
 5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
NOTES
                    (i.e., cells most active in reproducing
                    themselves and cells not fully mature will
                    be most harmed by radiation)

             2.     Specific Classifications of Mammalian Cell
                    Sensitivity

                    Group 1 (Extremely Radiosensitive)
                          Mature lymphocytes - a major
                          class ox circulating white blood cell
                          Erythroblast - red blood cell
                          precursor
                          Spermatogonia -  most primitive
                          cell in the spermatogenic series

                    Group 2  (Slightly less radiosensitive than
                          group 1)
                          Granulosa cells - cells surrounding
                          ovum which develop and mature in
                          the ovarian follicles
                          Myelocytes - (in bone marrow)
                          precursor to a leukocyte (colorless
                          ameboid/white blood cell)
                          Intestinal crypt cells - part of the
                          intestine lining
                          Germinal cells of the epidermal
                          layer of skin - primitive
                          development level of the cell

                    Group 3 (Radiosensitive)
                          Gastric gland cells - stomach gland
                          cells
                          Endothelial cells - lining of small
                          blood vessels

                    Group 4 (Moderately  Radiosensitive)
                          Osteoblasts - bone-forming cells
                          Osteoclasts - bone-absorbing cells
                          Chondroblasts - precursors to
                          cartilage cells
                          Spermatocytes
                          Spermatids
5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
NOTES
                    Group 5 (Slightly Radioresistant)
                          Granulocytes - white blood cells
                          Osteocytes - bone cells
                          Sperm
                          Superficial cells of the
                          Gastrointestinal tract

                    Group 6 (Moderately Radioresistant)
                          Parenchymal and duct cells of
                          glands
                          Fibroblasts - form intercellular
                          fibrous matrix
                          Endothelial cells of large  blood
                          vessels
                          Erythrocytes - red blood  cells

                    Group 7 (Radioresistant)
                          Fibrocytes - connective tissue cells
                          Reticular cells - fixed hemato-
                          poietic stem cells
                          Chondrocytes - cartilage  cells
                          Phagocytes - scavengers

                    Group 8 (Very Radioresistant)
                          Muscle cells and nerve cells - fully
                          diffeientiated, incapable of division

              Exposure Rate

              1.     Acute Exposure  - a radiation exposure
                    received in a short period of time, usually
                    considered to be less than 1 day

              2.     Chronic Exposure - a radiation exposure
                    spread over a long period of time, usually
                    considered to be over a period  of years

              Overall (Gross) Biological Effect

              1.     Acute Radiation  Syndrome (immediate
                    effects)

                    25 Rem - blood  changes
 5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                                                   NOTES
             2.
200 Rem - Hemopoietic Syndrome
         symptoms that develop within
         several hours following
         exposure:
             nausea, vomiting, malaise,
             fatigue, fever, blood
             changes

         symptoms that develop 2 to 3
         weeks after exposure:
         epilation (hair loss)

         symptoms that develop 1 to 2
         months after exposure: death

400 - 600 Rem - Bone marrow  destruction
               (reversible)

700 Rem - Irreversible destruction of bone
          marrow

1000 Rem - Gastrointestinal Syndrome
           immediately after exposure:
           severe nausea, vomiting
           and diarrhea 1 to 2
           weeks after exposure:
           death

2000 Rem - Central Nervous  System
           Syndrome
           minutes after exposure:
           unconsciousness  hours after
           exposure:  death

Latent Effects

a.     May be due to either a single large
      overexposure or continuing low-
      level overexposure

b.     Exposure may be due  to external
      radiation fields or internally
      deposited radioisotopes
5/93

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                              NOTES
                    c.     Genetic mutations - mutations of
                          germ cells
                          - direct evidence of radiation-
                          induced mutation in man is lacking
                          - largest group available for study
                          are descendants of Hiroshima and
                          Nagasaki
                          - no detectable effect on frequency
                          of prenatal or neonatal deaths or
                          malformations
                          - not enough time has passed to
                          reveal  recessive damage
                          - doubling dose for spontaneous
                          mutation rate is probably:

                                 15-30 rads: acute
                                 100 rads: chronic

                    d.     somatic mutations - mutations that
                          are produced in cells which are not
                          germ cells

                          i.     Cataract Formation - critical
                                 dose estimated between 20
                                 and 40 rads

                                 Fractionation of dose delays
                                 time of on set and decreases
                                 the incidence of severity

                          ii.     Life Shortening
                                       - slightly less than
                                       1 % per 100 rads
                                       chronic
                                        - between  1 and
                                       1.5% per 100 rads
                                       acute
                          iii.    Cancer
                                       - Leukemia Japanese
                                       survivors had an
                                       increase in incidence
                                       of leukemia @ 100
                                       to 900 rads.
5/93
10

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                                                   NOTES
             3.
Teratoeenic Effects
                                       Average rate of
                                       increase was linear,
                                       giving between 1 and
                                       2 cases/ 106/yr/rad.
                                       Latent period is
                                       shorter at higher
                                       dose.
                                       - Skin Cancer
                                       common in early
                                       radiologists and
                                       dermatology patients

                                       - Bone Tumors
                                       radium has increased
                                       incidence of bone
                                       tumors in individuals
                                       w/skeletal burdens
                                       such as the dial
                                       painters. Latent
                                       period of 20 to 30
                                       years.

                                       - Lung Cancer in
                                       mine workers
                                       exposed to radon gas
                                       and its daughter
                                       products (high
                                       percentage develop
                                       bronchogenic
                                       carcinoma within 15
                                       years after beginning
                                       work in mines).
                   a.     Implantation occurs at about 11
                          days after fertilization, with major
                          organogenesis extending to about
                          day 38

                          i.     irradiation during the first 2
                                weeks of pregnancy results
5/93
                      11

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RADIATION EXPOSURE AND BIOLOGICAL EFFECTS
                            NOTES
                               in spontaneous abortions and
                               gross abnormalities

                         ii.    irradiation between the 3rd
                               and 6th weeks may produce
                               gross abnormalities

                         iii.    beyond day 40 the embryo
                               is more radioresistant
                   b.     Study at Nagasaki

                         i.     30 women exposed within
                               2,000 meters of hypocenter
                               estimated at >20 rads total
                               dose each

                         ii.    16 children survived,  4 with
                               mental retardation

                         iii.    therapeutic abortions have
                               been suggested following
                               exposures to 10R
                                                 4
5/93
12

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                        SECTION 4

         RADIATION EXPOSURE LIMITS AND
         METHODS TO CONTROL EXPOSURE
         After completing this unit, participants will be able to:

         •    Describe the base criteria for occupational radiation
              exposure limits.

         •    Define the OSHA exposure limits for external radiation.

         •    Explain the basis of the new recommended limits
              (Federal Register, 1/27/87).

         •    Describe how time, distance, and shielding can be used
              to minimize radiation exposure.
5/93

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RADIATION EXPOSURE LIMITS AND METHODS
TO CONTROL EXPOSURE
NOTES
I.     Base Criteria

      A.    There should not be any occupational exposure to
             workers from ionizing radiation without the
             expectation of an overall benefit from the activity
             causing the exposure.

      B.    A sustained effort should be made to ensure that
             collective dose, as well as individual annual,
             committed, and cumulative lifetime doses, are
             maintained as low as reasonably achievable
             (ALARA).

      C.    Radiation dose received as a result of
             occupational exposure should not  exceed the
             limiting values for assessed dose to individual
             workers.
II.     OSHA Exposure Limits to External Radiation

       A.    29 CFR 1910.96 (b)

             1.    Standard Limits

                   a.     1.25 Rem/calendar quarter

                          whole body; head and trunk, active
                          blood-forming organs, lens of eyes,
                          or gonads

                   b.     18.75 Rem/calendar quarter

                          hands and forearm, feet and ankles

                   c.     7.5 Rem/calendar quarter

                          skin of whole body
5/93

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                   c.
RADIATION EXPOSURE LIMITS AND METHODS
TO CONTROL EXPOSURE

             2.    Extension of Limit

                   The whole-body limit can be exceeded if:

                   a.    the dose to the whole body does not
                         exceed 3 Rem during any calendar
                         quarter

                   b.    the cumulative dose to the whole
                         body shall not exceed 5(N-18) Rem
                         where N is the individual age, in
                         years, at his or her last birthday

                         the employer maintains adequate
                         past and current exposure records
                         to prove 5(N-18) Rem is not
                         exceeded.

             3.    Individuals under 18 years of age are not
                   allowed to receive a dose in excess of 10
                   percent of the standard's limits.

      B.     Regulatory Guide 8.13—Instruction Concerning
             Prenatal Radiation Exposure

             1.    Sites in National Commission on Radiation
                   Protection's (NCRP) recommendation that
                   a fetus should not be exposed to more than
                   500 mrem during the gestation period.
III.   New Recommended Limits - Federal Register dated
      1/27/87

      A.     Basis of Limits

             1.     Cancer and genetic effect risks are limited
                   by the effective dose equivalent (He).
             2.     Other health effects are limited by dose
                   equivalent (HT) to individual organs.
                                                                            NOTES
5/93

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RADIATION EXPOSURE LIMITS AND METHODS
TO CONTROL EXPOSURE
      B.
Limits
                                                               NOTES
             1.     Adult Worker - External Exposure
                   a.     Annual effective dose equivalent,
                         HE, should not exceed 5 rems
                         where HP  is defined as:
                                     WTHT
 5/93
                         WT is the weighing factor and HT is
                         the annual dose equivalent averaged
                         over organ tissue T.

                         Values of WT and their
                         corresponding tissues are:

                         gonads              0.25
                         breasts              0.15
                         red bone marrow     0.12
                         lungs               0.12
                         thyroid              0.03
                         bone surfaces        0.03
                         remainder           0.30

                         "Remainder" means the five other
                         organs with the highest doses. The
                         weighing factor for each such organ
                         is 0.06.

                   b.    The dose equivalent, HT,  received
                         in any year should not exceed 15
                         rems to the lens of the eye, and  50
                         rems to any other organ tissue, or
                         extremity of the body.

             2.    Adult Worker - Internal Exposure

                   a.    Committed dose equivalent HT)50 is
                         the sum  of all dose equivalents to
                         organ or tissue  that may accumulate
                         over an individual anticipated
                            4

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RADIATION EXPOSURE LIMITS AND METHODS
TO CONTROL EXPOSURE
                                                               NOTES
             remaining lifetime (taken as 50
             years) from radionuclides that are
             retained in the body.
             defined as:
                                             HT 50  s
3.
4.
                   b.     The committed effective dose
                          equivalent from any radionuclide's
                          intake plus any annual effective
                          dose equivalent from external
                          exposure will not exceed 5 rems.

                   c.     The dose equivalent to any organ
                          or tissue from any radionuclide
                          intake plus any annual dose
                          equivalent from external exposure
                          will not exceed 50 rems.

                   Occupational dose equivalent to individuals
                   under the age of 18 should  be limited to
                   one-tenth of the values for adults.

                   Exposure of an unborn child should not
                   exceed 500 mR during the entire gestation
                   period.
             Notes clarifying application of the
             recommendations

             1 .     Occupational exposure does not include
                   background or medical radiation
                   exposures.

             2.     The numerical values provided by these
                   recommendations do not apply to workers
                   responsible for the management of
                   response emergencies.
5/93

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RADIATION EXPOSURE LIMITS AND METHODS
TO CONTROL EXPOSURE

             3.     Emergency exposures are controlled by the
                   individual federal agencies having
                   jurisdiction.
IV.   Methods to Control Radiation Exposures

      A.     Time
                                                              NOTES
       B.
             1.
             2.
      Radiation exposure is a function of time.
      The longer an individual stays in a
      radiation area, the more exposure they will
      receive.

      Performance of work in an efficient
      manner reduces time.  This does not mean
      "do it fast"; it means to do it right the first
      time in the least time that is necessary.

      a.     pre-job planning

      b.     mock ups

      c.     tool inventories
      d.     walk-throughs

Distance
             1.
      Radiation levels decrease with distance
      from the source.
                   a.
                   b.
                   c.
             reach rods

             move the work

             simply work at arms length if
             possible
             Shielding
             1.     Shielding is simply a method of putting
                    something between you and the source.
 5/93

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RADIATION EXPOSURE LIMITS AND METHODS
TO CONTROL EXPOSURE
NOTES
             2.    Tenth Value - that amount of shielding
                   required to reduce the radiation levels to
                   one-tenth its original value.

                   a.     gamma       2" lead, 4"  steel, 24"
                                       H2O

                   b.     neutron      10" H2O, 10" poly
                                       material

                   c.     alpha         1 sheet of paper

                   d.     beta          1 sheet of aluminum
                                       foil

             3.    Considerations when shielding

                   a.     The very act of installing and
                          removing shielding can result in
                          radiation exposure. It may take
                          more exposure to put it  up and  take
                          it down than is saved by having it
                          there for the job.

                   b.     Weight of the shielding  materials,
                          such as lead, is very heavy.  Two
                          inches of lead for any size
                          container is a significant weight
                          problem.  Structural consideration
                          of the floor should be made  for any
                          shielding project.

             D.    Awareness

                   1.     You cannot see, smell, taste, or
                          feel radiation.  When people start
                          to work, they tend to forget about
                          the radiation exposure that is
                          occurring.  You can reduce your
                          exposure just by paying attention to
                          what is going on and use time,
                          distance, and shielding to  your
                          advantage.
5/93

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                        SECTION 5

           BASIC CONCEPTS IN RADIATION
           DETECTION AND MEASUREMENT
         After completing this unit, participants will be able to:

         •    Describe the basic components of radiation detection
              instruments.

         •    Define energy resolution, dead time, absolute
              efficiency, and intrinsic efficiency.

         •    Describe how source characteristics affect the
              measurement of radiation.
5/93

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BASIC CONCEPTS IN RADIATION DETECTION
AND MEASUREMENT
NOTES
I.  Components of Instruments

       A.    Sensing element (detector): responds to the
             radiation and provides a measurable signal to the
             indicating element

       B.    Indicating element

             1. meter
             2. recorder
             3. sealer
             4. speaker
II.  Characteristics of Instruments

       A.     Radiation Interaction with the Detector

              1.     lonization

              2.     Excitation

              3.     Both ionization and excitation (directly or
                    indirectly) result in the formation of
                    electrical charges

              4.     If an electrical field is applied across the
                    detector, the electrical charges generated
                    in the detector can be collected

       B.     Operating Mode of the Indicating Element

              1.     Pulse mode: records an output pulse for
                    each individual interaction between the
                    detector and the radioactive particle or
                    photon

              2.     Current mode: records  the amount of ion
                    pairs produced  by the radiation in the
                    detector.
5/93

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BASIC CONCEPTS IN RADIATION DETECTION
AND MEASUREMENT
NOTES
      C.     Recording Modes of Meters

             1.     Rate meters: record the pulse or current
                   rate; units include cpm, mR/hr, or
                   mrem/hr

             2.     Integrating instruments: tally the pulses or
                   total current for the duration of the
                   measurement

      D.     Advantages of Pulse Meters

             1.     Greater sensitivity; lower limit of
                   detection

             2.     Can measure pulse height (amplitude)

                   a.     amplitude is proportional to energy
                          of the radiation

                   b.     knowledge of the radiation energy
                          and detector design can identify the
                          type of emission

                   c.     specific  radionuclide may be
                          identified

             3.     Discriminator: allows only pulses of a
                   specific amplitude to pass to the indicating
                   element

                   a.     used to differentiate pulse heights

                   b.     sometimes can identify radionuclide

      E.     Other Characteristics of Instruments

             1.     Dead Time:  the minimum amount of time
                   required, after radiation interaction with
                   the detector, in order for the next
                   interaction to register as a separate pulse.
5/93

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BASIC CONCEPTS IN RADIATION DETECTION
AND MEASUREMENT
                                                      NOTES
                   a.    an ionizing event occurring during
                         the dead time will not produce a
                         pulse

                   b.    counts recorded can be corrected
                         for dead-time losses

             2.     Energy Resolution: ability of a detection
                   system to distinguish between two pulses
                   of slightly different sizes

             3.     Counting Efficiency

                   a. some radiation will pass through the
                   detector without interacting with it

                   b. all radiation emitted from the source
                   does not pass through the probe

             4.     Absolute Efficiency (ae): indicates how
                   well the radiation detector counts all of the
                   radiation emitted from the source.
                  _  number of pulses recorded
                     number  of source emissions
             5.    Intrinsic Efficiency (ie): accounts for the
                   fact that all of the radiation emitted by the
                   source may not reach the detector. It also
                   indicates how well the radiation detector
                   counts all of the radiation that passes
                   through the detector.
               ie =
  number of pulses recorded
number  of incident radiations
                    Note:  incident radiations are those that
                          reach the detector
 5/93

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BASIC CONCEPTS IN RADIATION DETECTION
AND MEASUREMENT
NOTES
III. Source Characteristics

       A.    Isotropism: radiation is emitted by the source in
             all directions with equal frequency

       B.    Source Geometry: the physical shape of the
             source

             1.     Point source:  a source that is very small
                    compared to the distance from the source
                    to the detector
             2.     Line source:  finite line sources extending
                    along a single axis, such as small pipes

       C.    Geometry Factor: the fraction of the source
             sphere that actually intercepts the detector

             1.     4 TT Geometry: completely enclosing the
                    source within the sensitive volume of the
                    detector

             2.     2 T Geometry: enclosing one-half of the
                    source sphere within the sensitive volume
                    of the detector (the situation for nearly all
                    instruments)

       D.    Self-absorption: the absorption of radiation within
             the source itself

             1.     encapsulated alpha and beta sources

             2.     samples of large mass

       E.    Attenuation: unless the sample is placed under a
             vacuum, the air space between the source and the
             detector will absorb a portion of the radioactive
             emissions
5/93

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                        SECTION 6

       RADIATION DETECTION INSTRUMENTS
         After completing this unit, participants will be able to:

         •    Explain the ionization curve.

         •    Describe the operation of a direct reading dosimeter.

         •    List the requirements that are considered in the
              selection of radiation instruments.
5/93

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RADIATION DETECTION INSTRUMENTS
NOTES
I.      Gas lonization Detectors
       A.    Principle of operation
             1.     Detector: usually consists of a power
                    supply and a closed, electrically
                    conductive cylinder filled with a gas
                    (Figure 1)

                    a.     metal chamber walls: penetrated by
                          photons and high-energy beta
                          particles

                    b.     chamber wall "window": a portion
                          of the chamber wall may be made
                          of mylar or mica, which is easily
                          penetrated by alpha and low-energy
                          beta particles

                    c.     the chamber may be made
                          "directional" by adding a window
                          and/or shielding

                    d.     anode: a thin wire in the center of
                         ^the chamber, positively charged

                    e.     cathode: the chamber wall,
                          negatively charged

                    f.     incident radiation causes ionization
                          of the gas, producing an ion pair (a
                          free electron and a positively
                          charged gas molecule

                    g.     number of ion pairs produced
                          depends on type of gas as well as
                          the type and energy of the radiation

                    h.     on average, one ion pair is
                          produced for every 30 to 35 eV of
                          energy transferred to the gas
5/93

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RADIATION DETECTION INSTRUMENTS
                                                               NOTES
                   i.     when voltage is applied across the
                         chamber, the ion pairs produced
                         move to their respective electrodes
                             Collecting Electrode
                                  (anode)
 Ion Chamber -
                  •f
                                                   Pulse
                T
               Wall
             (cathode)
                                  Power Supply
      B.
           FIGURE 1
BASIC DETECTOR SCHEMATIC

Relationship between applied voltage and the
number of electrons  collected at the anode
(Figure!)

1.     Recombination Region

       a.     voltage across the electrodes is low

       b.     attraction between ion pairs may be
             greater than that between ion pairs
             and electrodes

       c.     no radiation detectors operate in
             this region
5/93

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RADIATION DETECTION INSTRUMENTS
NOTES
                    lonization Chamber Region (saturation
                    region)

                    a.     voltage across electrodes sufficient
                           to cause the collection of all
                           electrons

                    b.     moderate increases in voltage do
                           not increase the electron current

                    c.     different types of radiation can be
                           distinguished from each other
                           because of the different pulse
                           heights produced

                    Proportional Region

                    a.     voltage increased such  that the ion
                           pairs collected are  greater than the
                           number of primary ion pairs
                           formed (accelerated electrons cause
                           secondary ionizations in gas)

                    b.     gas amplification factor, or
                           multiplication factor:  a measure of
                           secondary ions produced

                    c.     multiplication factor is constant
                           over small voltage ranges

                    d.     proportional detectors can
                           distinguish among  alpha, beta, and
                           gamma radiation

                    Limited  Proportional Region: no useful
                    purpose  for radiation measurement

                    Geiger Region: voltage high enough to
                    cause avalanche along entire length of
                    anode
 5/93

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RADIATION DETECTION INSTRUMENTS
NOTES
                    a.     all pulses are the same size,
                           regardless of the type of radiation
                           (detector cannot distinguish among
                           different types of radiation)

                    b.     cation removes electron from
                           detector wall producing x-rays

                    c.     quenching gas, which supplies
                           electrons,  is added to the chamber
                           to prevent continuous discharge

                    d.     typical quenching gases include
                           bromine, chlorine, ethanol, and
                           methane

              6.     Continuous Discharge Region: voltage
                    increased until arcing occurs across  the
                    electrodes (operation  in this region can
                    permanently damage detectors)  (Figure 2)

       C.     lonization Chambers: instruments designed to
              operate in the ionization chamber region

              1.     Passive ion chambers: voltage is applied
                    by charging a capacitor. Ions formed by
                    incident radiation neutralize the charge

                    a.     drop in voltage proportional  to dose

                    b.     loss of charge due to leakage
                           results in  false reading

                    c.     pocket ionization chamber:

                           i.     integrating instrument

                           ii.     quartz  fiber attached  to a
                                  rigid metal electrode

                           iii.    positive charge is applied to
                                  the electrode, causing the
                                  repulsion of the quartz fiber
 5/93

-------
RADIATION DETECTION INSTRUMENTS
                                                           NOTES
                    SIMPLE
                   IONIZATION
                                GAS AMPLIFICATION
cr
LU
s
cz.
UJ
Q.
     o
              10NIZATION
               CHAMBER
                REGION
     o
     cr
     fc
     o
     c:
     LLJ
     CD
                   PROPORTIONAL
                      REGION
LIMITED
PROPOR-
TIONAL
REGION
  5

GEIGER
REGION
 REGION Of
CONTINUOUS
 DISCHARGE
                                    VOLTAGE
                               FIGURE 2
    RELATIONSHIP BETWEEN APPLIED VOLTAGE AND THE NUMBER OF
                 ELECTRONS COLLECTED AT THE ANODE
 5/93

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RADIATION DETECTION INSTRUMENTS
                                                          NOTES
             2.
       iv.    electrons produced by
             incident radiation neutralize
             the charge on the electrode
             and the quartz fiber moves
             toward the electrode

       v.     available in direct reading
             and non-self-reading models
             (Figure 3)

Active lonization Chamber: internal high
voltage supply
                    a.     integrating meter measures total
                          current

                    b.     nonintegrating functions as a rate
                          meter

                    c.     "cutie pie" is prototype

                          i.      thin window with beta shield

                          ii.     yields gross beta-gamma and
                                 gross gamma

                          iii.    beta-gamma minus gamma
                                 equals beta

                          iv.    relative response as a
                                 function of energy flat from
                                 10 to 1000 keV (RR = 1
                                 for gamma and x-rays)

                          v.     seldom used  for alpha
                                 radiation

       D.    Proportional Counters: a gas ionization detector
             that operates in the proportional region of the
             pulse-height voltage curve

             1.     lonization gases provide stable operating
                    characteristics and high amplification.
                    Examples include P-10 (10% methane and
5/93

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RADIATION DETECTION INSTRUMENTS
NOTES
                  90% argon) or 4% isobutylene and 96%
                  helium.
              Window
              assembly


              Eye lens
                               Polvihene
                               end cap
          Pocket clip
             Field lens
                            5	Sleeve
                               Graticule
                            • i	Cower tube
             Polythene
              end cao
                    7-94nvm/7-62ni/m
                     •3127.30CTDIA.
                      FIGURES
  CROSS-SECTION OF A DIRECT-READING POCKET
                     DOSIMETER
5/93

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RADIATION DETECTION INSTRUMENTS
                                                          NOTES
             2.
Easily differentiate alpha from beta
radiation (Figure 4)

a.     alpha particles have a higher
      specific ionization than beta
      particles and require lower voltage
      to collect ions formed by incident
      radiation

b.     plotting the count rate as a function
      of operating voltage yields a graph
      with two plateaus

      i.     only alpha particles are
             collected at low voltage
             (alpha plateau)

      ii.     increasing voltage allows the
             collection of beta particles
             and the development of the
             beta plateau (the beta
             plateau is not as flat as the
             alpha plateau because of the
             wide variations in beta
             energies)
       20.000
        10.000
     S
     c
                   ALPHA PUOIAU
             II \f  1   111   .1,1,1.1
              CD   1COD   1JOD    )«C   )MO   )CD    20X

                        COUMTlRVaTACCft'aTSI
                        FIGURE 4
 PLATEAUS FOR TYPICAL PROPORTIONAL COUNTER
5/93

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RADIATION DETECTION INSTRUMENTS
                                    NOTES
       3.     Sealed proportional counters: have limited life due
             to degradation of ionizing gas

       4.     Gas-flow proportional detectors: longer life due to
             replacement of ionizing gas

       5.     Neutron measurement using proportional counters

             a.     neutrons do not interact directly with the
                    ionization gas

             b.     chamber is filled with boron trifluoride gas

             c.     thermal (low-energy) neutrons interact
                    with the boron trifluoride and produce
                    alpha particles, which interact with the
                    ionization gas

             d.     Fast (high-energy) neutrons are measured
                    by wrapping the boron trifluoride tube
                    with polyethylene, paraffin, or some other
                    hydrogen-rich material. The wrapping
                    moderates (decreases the energy) the fast
                    neutrons.

             e.     BF-3 tubes measure only neutron radiation
                    (Figure 5)
5/93
10

-------
RADIATION DETECTION INSTRUMENTS
                                             NOTES
 Moderator Shield
                        f ////// //////// /////
                Br, Coating      ,n>  + ,B'° - GB")'-
                        FIGURES
                       BF-3 TUBE

             Geiger-Mueller Counters: ionization detectors
             operate in the Geiger region of the pulse height-
             voltage curve

             1.     Characteristics

                   a.     used to count alpha, beta and
                         gamma radiation

                   b.     cannot distinguish among different
                         types of radiation (all pulses same
                         size in Geiger region)

                   c.     wall and window thicknesses
                         i.
                         n.
30 mg/cm2 for gamma and
high energy beta

0.4 to 1.4 mg/cm2 for alpha
and low-energy beta
5/93
         11

-------
RADIATION DETECTION INSTRUMENTS
                                    NOTES
                    d.     most versatile radiation detectors;
                           inexpensive, easy to operate,
                           sensitive, and reliable

                    e.     subject to continuous discharge in
                           high radiation fields

              2.     Operating modes

                    a.     pulse counter

                    b.     portable survey meters

                           i.      activity (CPM)

                           ii.     dose rate (mR/hr)

              3.     Common detector configurations

                    a.     stainless steel tube with 30 mg/cm2
                           wall housed in ABS plastic as
                           optional beta shield (HP-270)

                    b.     "Pancake" GM tube, 5 cm diameter
                           1.4 to 2.0 mg/cm2 mica
                           window.  Sensitive to alpha  > 3
                           MeV.


II.     Scintillation Detectors:  radiation interacting with the
       detector creates visible  light  photons
       (Figure 6)

       A.     Principles of operation

              1.     Incident  radiation transfers energy to the
                    phosphor by ionization or excitation

              2.     Excited electrons  move into defects  or
                    gaps in the lattice called "traps"

              3.     "Trapped"  electrons return  to lower energy
                    levels and emit visible light
5/93
12

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RADIATION DETECTION INSTRUMENTS
                                  NOTES
            4.    Signal is amplified with a photomultiplier
                  tube and an electron amplifier
                       FIGURE 6
              DIAGRAM OF DETECTOR
      B.    Peripheral Circuitry

            1.    Survey meters

                  a.    pulse shaper

                  b.    rate meter

            2.    Pulse Height Analyzers (Figure 7)

                  a.    Spectrometry: gamma and alpha
                        radiation is emitted at discrete
                        energy levels. Identification of the
                        radionuclide is possible by analysis
                        of the energy spectrum.
5/93
13

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RADIATION DETECTION INSTRUMENTS
                                    NOTES
                   b.     Pulse height analyzer: pulse height
                          is proportional to the energy of the
                          incident radiation. The pulse
                          height analyzer sorts the detector
                          signals by height (energy).
                        FIGURE 7
              PULSE HEIGHT HISTOGRAM
                    c.     Single channel analyzer: manually
                          operated, analyzes only one channel

                    d.     Multichannel analyzer: several
                          hundred to thousand single channel
                          analyzers. Accumulated data
                          displayed as a spectrum with
                          channel number (photon energy) on
                          the x-axis versus counts per channel
                          on the y-axis.

             Inorganic Scintillators

             1.     Crystals of alkali halides (Nal) with
                    thallium inclusions
 5/93
14

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RADIATION DETECTION INSTRUMENTS
                                                            NOTES
             2.
a.      used for gamma and x-ray detection

b.      poor energy resolution

Zinc sulfide (ZnS) with silver inclusions   ~
5/93
                    a.      used for heavy charged particles
                           such as alpha

                    b.      often applied to the back of thin
                           window or painted on the face of
                           photomultiplier tube for alpha
                           survey meters

                    Note:  The following detection principles
                           will not be applied in this course
                           but you  will hear reference to them
                           when dealing with lab results.

       D.    Liquid Organic Scintillators

             1.     Produced by dissolving an organic
                    scintillator material in an organic solvent

             2.     Counting efficiencies approach 100%

             3.     Particularly advantageous for counting
                    low-energy beta emitters such as C-14 and
                    H-3

             4.     Can be used for alpha emitters

             5.     Cocktail (scintillator plus sample) is placed
                    in glass or plastic vial

             6.     Sample counted with photomultiplier tube
                    in light-tight enclosure

       E.    Semiconductor Detectors

             1.     Principle of operation

                    a.      detector made of a solid crystalline
                           material
                      15

-------
RADIATION DETECTION INSTRUMENTS
                                    NOTES
                    b.     "impurities" are added to the
                          crystalline material

                    c.     similar to gas ionization detector;
                          ionizations in the sensitive volume
                          of the detector cause a voltage
                          pulse which is amplified and
                          counted on a sealer  (secondary ions
                          are not produced)

                    d.     primary use is gamma spectroscopy
                          because of high degree of
                          resolution of energy peaks

                    e.     semiconductor detectors have lower
                          counting efficiencies than sodium
                          iodide crystals, but  better energy
                          resolution

                    f.     semiconductor detectors are
                          expensive and fragile

                    g.     there are two predominant types in
                          use: germanium-lithium (GeLi)
                          and selenium-lithium (SeLi)

                          i.     GeLi detectors require
                                 cooling with liquid nitrogen
                                 for proper operation

                          ii.     SeLi detectors can be
                                 operated at room
                                 temperature, but have lower
                                 counting efficiencies than
                                 GeLi detectors
III.    Instrument Selection Criteria

       A.    Purpose of Monitoring

             1.     Exposure rate

             2.     Count rate
5/93
16

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RADIATION DETECTION INSTRUMENTS
                                  NOTES
      B.     Degree of Accuracy and Precision




             1.     Detection vs. measurement




             2.     Sensitivity and lower limit of detection




      C.     Types of Radiation



             1.     The principle factor in instrument selection




      D.     Source Form




             1.     Physical state and matrix




      E.     Radiation Field Characteristics




             1.     Intensity of radiation field




             2.     Uniformity of radiation field
 5/93
17

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                        SECTION 7

     SURVEYING FOR RADIOACTIVE MATERIALS
          After completing this unit, participants will be able to:

          •    Explain the purpose of surveying for  radioactive
              materials.

          •    Define different types of sampling protocols and how
              sampling data are reported.

          •    Define and compare different sampling methods.
5/93

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SURVEYING FOR RADIOACTIVE MATERIALS
                                                                     NOTES
I.
III.
Surface Contamination.

A.    Loose Surface Survey.

      1.     Dry smear.

      2.     Wet smear.

      3.     Solvent smear.
       B.
       4.     Large area wipes.

       Fixed Surface Survey.

       1.     Direct frisk.
II.     Soil.
       A.    Considerations.
             1.    Accurate determination of mass.
             2.    Self attenuation must be taken into account.
             3.    Usually loaded into small planchets.
Water.

A.  Bulk.

       1.

       2.
                   Liter bottles.

                   Small samples taken and boiled off, counting
                   the residue.
IV.    Air.
5/93
A.     Paniculate Contaminants.

       1.     High-volume "grab"  samples.

                                   2

-------
SURVEYING FOR RADIOACTIVE MATERIALS
NOTES
                   a.     5-10 minutes sample time.

                   b.     Used for:

                                Quick check of an area for
                                entry.

                                Monitoring airborne causing
                                evolutions.

                                Routine air monitoring.

             2.     Low-volume Samples.

                   a.     1-12 hours sample time.

                   b.     Used  as   "proof   that  nothing
                          happened or  to  assess damage if
                          something did.

                   c.     Placed  in   work  areas  based on
                          expected   airborne   contamination
                          levels.

                   d.     Routine air monitoring.

      B.     Gaseous Contaminants.

             1.     Noble  gases -  Maranelli sampler or liter
                   bottle  (nonreactive,  needs  to  be volume
                   collected).

             2.     Iodine  - Charcoal filter cartridges
                          (reactive, charcoal used for most).

      C.     CAM (Constant Air Monitor).  Primarily used for
             particulates,   but  can   also   detect   gaseous
             contaminants.

             1.     Continuous air monitoring.

             2.     Alarm  capability.
5/93

-------
SURVEYING FOR RADIOACTIVE MATERIALS
NOTES
             3.     Some have isotopic identification capability..

             4.     Placed   in   areas   where   airborne
                   contamination may occur, early warning is
                   desired, or continuous documentation of air
                   quality is required.
V.     Dose Rate Survey  (1 mR/hr EPA action limit).

       A.    General Area.

             1.      Waist level
                    (Center of the "whole body").

             2.      > 12 inches  away from  any surface  (18
                    inches for commercial power plants).

       B.    Contact Dose Rates.

             1.      On contact with an object or surface.

       C.    Hot spot (contact reading).

             1.      5 times  the  general  area.   Used as  a
                    thumbrule.

             2.      > 150 mR/hr. Used as a thumbrule.

                    NOTE:      The thumbrules are general
                                guidance.    Hot  spots  are
                                normally small  areas where
                                radiation   levels   are
                                significantly  higher   than
                                normal area radiation levels.
                                This is  of primary interest
                                when  controlling  personnel
                                exposures.
 5/93

-------
SURVEYING FOR RADIOACTIVE MATERIALS
NOTES
VI.   Documentation.

      A.     Survey sheet.

             1.     Surveyor name(s).
             2.     Date and time of samples (isolation time is
                   critical).

             3.     Locations of samples.

             4.     Conditions  of samples.

             5.     Instruments used  (with  serial numbers).
                   Ensure instrument is in calibration at time of
                   use.

             6.     Area or item description.

             7.     Type of survey performed.

                   Note:

                   Any additional  information  that  may be
                   useful  in supplementing a survey should also
                   be documented  (e.g., survey after decon-
                   tamination,  initial  survey,   subsequent
                   survey, and incident survey)
5/93

-------
                          SECTION 8

            RADIATION SIGNS AND LABELS
          After completing this unit, participants will be able to:

          •    Identify the OSHA regulation that provides posting
               requirements for ionizing radiation.

          •    Define, according to OSHA, the following: radiation
               area, high radiation area, airborne radioactivity, and
               radioactive material.

          •    Describe one example of a posting practice found in
               industry.
5/93

-------
RADIATION SIGNS AND LABELS
NOTES
I.     29 CFR 1910.96 Posting Requirements

      A.     Conventions

             1.     The radiation caution color scheme is
                   magenta on a yellow background.

             2.  '   The radiation symbol is a three blade
                   design called a trefoil (see Figure  1).
                           Figure 1
     CONVENTIONAL RADIATION CAUTION SYMBOL

       B.     Definitions and Posting Requirements

             1.     Radiation Area

                   a.      "Radiation Area" means any area,
                         accessible to personnel, in which
                         there exists radiation at such levels
                         that a major portion of the body
                         could receive in any  1 hour a dose
                         in excess of 5 millirem, or in any 5
                         consecutive days a dose in excess
                         of 100 millirem.
 5/93

-------
RADIATION SIGNS AND LABELS
NOTES
             2.     Each radiation area shall be
                    conspicuously posted with a sign or signs
                    bearing the radiation caution symbol and
                    the words:

                                        Caution
                                   Radiation Area

             3.     High Radiation Area

                    a.     "High Radiation Area" means any
                          area, accessible to personnel, in
                          which there exists radiation at such
                          levels that a major portion of the
                          body could receive in any  1 hour a
                          dose in excess of 100 millirem.

                    b.     High radiation areas  must be locked
                          or guarded.

                    c.     High radiation areas  shall be
                          conspicuously posted with  a sign or
                          signs bearing the radiation caution
                          symbol  and the words:

                                         Caution
                                 High Radiation Area

             4.     Airborne Radioactivity Area

                    a.     "Airborne Radioactivity Area"
                          means any room,  enclosure, or
                          operating area in which airborne
                          radioactive material exists in
                          concentrations:

                          i.     in excess of the amounts
                                 specified in 10 CFR 20
                                 Appendix B,  Table I,
                                 Column I, or;
5/93

-------
RADIATION SIGNS AND LABELS
NO1
                        ii.    which averaged over the
                              number of hours in any
                              week during which
                              individuals are in the
                              area, exceed 25 %  of the
                              amounts specified  in 10
                              CFR 20 Appendix B, Table
                              I, Column I  (see Figure 2).
5/93

-------
 RADIATION SIGNS AND LABELS
                            NOTES
       Port 20, App. B
10 CFR Ch. I (1-1-91 Edition)
        APPENDIX B—CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND—Continued
                           *  IS*« loOlnoUt «l * u

Co 60..._..~..«. .„

Cu 64 	 	 	

Cm 242 	

Cm 243 	

Cm 244 	

Cm 245 	 . 	

Cm 246 	

Cm J47 	 _.._

Cm ?4( 	

Cm2<9. .



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3- 10"
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3 • 10"
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6 • 10 '
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3. 10' '
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1 • 10 •
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4 .. 10-
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1 . 10
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                                   FIGURE!
                      EXCERPT:  10 CFR 20, APPENDIX B
5/93

-------
RADIATION SIGNS AND LABELS
NOTES
                   b.     Each airborne radioactivity area
                          shall be conspicuously posted with
                          a sign or signs bearing the radiation
                          caution symbol and the words:

                                Caution
                          Airborne Radioactivity Area

             4.    Radioactive Materials Area

                   a.     Each area or room in which
                          radioactive material is used or
                          stored and which contains any
                          radioactive material (other than
                          natural uranium or thorium) in any
                          amount exceeding 10 times the
                          quantity of such material specified
                          in Appendix C to 10 CFR Part 20
                          (see Figure 3) shall be
                          conspicuously posted with a sign or
                          signs bearing the radiation caution
                          symbol and the words:

                                       Caution
                                 Radioactive Materials

                   b.     Each area or room  in which natural
                          uranium or thorium is used or
                          stored in any amount exceeding 100
                          times the quantity of such material
                          specified in Appendix C to 10 CFR
                          20 shall be conspicuously posted
                          with a sign or signs bearing the
                          radiation caution symbol and the
                          words:

                                       Caution
                                 Radioactive Materials
 5/93

-------
 RADIATION  SIGNS AND  LABELS
                                                                                                     NOTES
                      APPENDIX C—Continued
                                                           APPENDIX C—Continued
Naooymum-147.
Naodymtunvl^
NK*»(.S9
N*k«t-«,
Osmium-1B1
Ownium-193
Paladmm-103
Palladium-10S.,
PftocpfaOfus*32<
Ptatinum.191
Platinum- 193m
Platinum. I S3
Platinun>197nv
PlaUnun>197_,
Potoraum-ZtO-
Pra:
PraMOdymium-IO
Promathium-14 7 „.
Piom«thJum-149_
Radium.228
Rhodwm-103m
RnodiunvlOS
RubUum-M
Rubioum-a7_..
Ruthanium-97,

Ruthenium-1 OS...
Ruthanum-106.,
Samarium-1 Si „.
Sam»num-153...
Scandmm-46
Scanoum-47_.,
Scandum-48
SbOntium-65..
Suonlium-89.
SlronUjm-eo.
Suonbum-91
SlronUum-92	
Sulphur-3S	
Tantalunvi82	
Tachn*uunv9«....
T*chn*tium-97m
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                                                      10
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TtHu>ium.t31n<_
T«Uunum.132	
7«*1>unvl60	
Thafcum-200	
TnaiUum-201	
TfulUum-202	
                                                Thonum (nalural)*,
                                                Thulium. 170	
                                                Thulium-171	
                                                Tm-113	
Tungilan-181	
Tung»lan-1BS		
Uranium (natural)1	
OYanium-233		
Uranium-234—Uranum-235..
Vanadmm-48		
Xanon-131 m		
Xenon-133.		
X*non-135—
Ynarbium.17S.
Ynnum-90	
Yllnum-91.
                                                Any alpha «rinng ildooocl-dt nol failed «bovt
                                                  or  miitur«l  at alptu *mill«fs ol unknown
                                                  composition __ _ „ ............... „ ...... „.„...„..«..._«..„
                                                Any radionucfoM othvf tran alpha crmilmg ra*
                                                  dionuetidci. no: k>i«d above or muturei ol
                                                  txla •iratlart ol unknown compovtion -----------
                                                                                           10
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                                                                                           10
                                                                                           10
                                                                                          100
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                                                                                           10
                                                                                         1.000
                                                                                          IOO
                                                                                          100
                                                                                          100
                                                                                           10
                                                                                           10
                                                                                          IOO
                                                                                          IOO
                                                                                           10
                                                                                          100
                                                                                         1.000
                                                                                           10
                                                                                           10
                                                                                           10
                                           .0 1
                                                                                            .1
                                                  'Basad on alpha disinitoiaion me ol Th-232. Th-230 and
                                                ih€« daughitf p>oduci>
                                                  •Based on alpha dmnltoiaian ral* ol U-238. U-234. and
                                                U-23S.
  NOTE  For  purposes  of  i 20.303,  where
there Is  Involved a combination of Isotopes
In known amounts, the  limit for the combi-
nation should be derived as follows: Deter-
mine, for each isotope  In the  combination.
the ratio between  the  quantity present in
the combination and the limit otherwise es-
tablished for  the specific isotope when not
In combination. The sum of such ratios for
all the Isotopes In the combination may not
exceed "1" (i.e.. "unity").
(35 PR 6425. Apr. 22. 1970, as amended at 36
FR 16898, Aug. 26.  1971; 38 PR 29314. Oct.
24.  1973: 39 FR 23991. June  28. 1974: 45 FR
71763. Oct. 30. 1980)
                                                     FIGURE 3
                                  EXCERPT:   10  CFR 20, APPENDIX C
5/93

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RADIATION SIGNS AND LABELS
NOTES
             5.     Containers

                    a.      Each container in which is
                           transported, stored, or used a
                           quantity of any radioactive material
                           (other than natural uranium or
                           thorium) greater than the quantity
                           of such material specified in
                           Appendix C to 10 CFR Part 20
                           shall bear a durable, clearly visible
                           label bearing the radiation caution
                           symbol and the words:

                                        Caution
                                 Radioactive Materials

                    b.     Each container in which natural
                           uranium or thorium is transported,
                           stored, or used in a quantity greater
                           than 10 times the quantity specified
                           in Appendix C to 10  CFR Part 20
                           shall bear a durable, clearly visible
                           label bearing the radiation caution
                           symbol and the words:

                                        Caution
                                 Radioactive Materials

                    c.     A label shall not be required:

                           i.     If the concentration of the
                                 material in the container
                                 does not exceed that
                                 specified in Column 2 of
                                 Table I of Appendix B to
                                  10 CFR Part 20, or

                           ii.    For laboratory containers,
                                 such as beakers, flasks, and
                                 test tubes, used transiently
                                 in laboratory procedures,
                                 when the user is present.
 5/93

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RADIATION SIGNS AND LABELS
NOTES
                    d.     Where containers are used for
                           storage, the labels required in
                           this subparagraph shall state also
                           the quantities and kinds of
                           radioactive materials in the
                           containers and the date of
                           measurement of the quantities.

             Exceptions from Posting Requirements

             1.     A room or area is not required to be
                    posted with a caution sign because of the
                    presence of a sealed source, provided the
                    radiation level  12 inches from the surface
                    of the source container or housing does not
                    exceed 5 millirem per hour.

             2.     Rooms or other areas in  onsite medical
                    facilities are not required to be posted with
                    caution signs because of the presence of
                    patients containing radioactive material,
                    provided that there are personnel in
                    attendance who shall take the precautions
                    necessary to prevent the  exposure of any
                    individual to radiation or radioactive
                    material in excess of the limits established
                    in 29 CFR 1910.96 Section (b).

             3.     Caution signs are not required to be posted
                    in areas or rooms containing radioactive
                    materials for periods of less than 8 hours,
                    provided that:

                    a.     The  materials are constantly
                           attended during such periods  by an
                           individual who shall take the
                           precautions necessary to prevent the
                           exposure of any individual to
                           radiation or radioactive
                           materials in excess of the limits
                           established in 29  CFR  1910.96
                           Section  (b).
5/93

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RADIATION SIGNS AND LABELS
                                  NOTES
4
                   b.    Such area or room is subject to the
                         employer's control.
II.    Industry Accepted Conventions

      NOTE: THE FOLLOWING POSTING PRACTICES
      ARE SITE SPECIFIC

      A.    "Very High Radiation Area"

             1.     Posted when an area in the room has a
                   general area dose rate > 1 R/hr

             2.     Normally a sign is conspicuously posted
                   that bears the radiation caution symbol and
                   the words:

                         Caution
                 Very High Radiation Area

      B.    "Surface Contaminated Area"

             1.     Posted when radioactive loose surface
                   contamination exceeds the limits for an
                   uncontrolled area.

             2.     Normally a sign is conspicuously posted
                   that bears the radiation caution symbol and
                   the words:

                          Caution
                Surface Contaminated Area

             3.     Step-off pads are posted areas that
                   designate the point of entry and exit for a
                   contaminated area.

             4.     Yellow and magenta boundary tape is used
                   to establish the perimeter of a
                   contaminated area.
 5/93
10

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RADIATION SIGNS AND LABELS
                                    NOTES
             MultiPocket Signs
             1.    Used to identify areas that:
                   a.     are temporary
                   b.     have changing requirements
                   c.     that fall into more than one
                          classification (i.e. "High Radiation
                          Area" which is also a
                          "Contaminated Area")

       D.    Radiation Area Ribbon/Rope

             1.    Yellow and magenta ribbon or rope, set
                   waist high, used to separate a radiological
                   control area from an noncontrolled area.
                   Applicable signs are hung on the rope.
5/93
11

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                        SECTION 9

             CONTAMINATION CONTROL
         After completing this unit, participants will be able to:

         •    Define contamination and give one example of a
              contamination problem.

         •    Describe contamination surveying techniques.

         •    Describe radwaste reduction techniques.
5/93

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CONTAMINATION CONTROL
                                             NOTES
I.      Definitions and Units of Measure

       A.    Radioactive contamination is a fine form of
             radioactive material in a place that it is not
             wanted.

             1.     Loose surface contamination is
                    contamination that comes off of a surface
                    when it is wiped by a dry filter paper; this
                    is called a dry smear.

             2.     Fixed surface contamination will not come
                    off on a dry smear.

       B.    Units of Measurement

             Contamination is radioactive material, so it has to
             be measured in the same units as any other
             radioactive material,(i.e., dpm or curies).
             However, contamination is a very fine form of
             radioactive material, consequently, subunits tend
             to be used, (i.e., microcuries).

             The main concern with contamination is keeping
             it from entering our bodies. The routes of entry
             are the same as for any hazardous material:  open
             cut or wound, puncture, inhalation, or ingestion.
             Consequently, the important question when
             measuring contamination is the location of
             contamination, that is, is it on a solid surface, in
             the air, in water, or mixed in soil.  Depending on
             the medium containing the contamination, the
             following units are used:
             Surface
             Air
             Water
(dpm/100 cm2)= dpm per
100 square centimeters

(jtCi/mL) = microcuries per
milliliter of air

(/*Ci/mL)= microcuries per
milliliter of water
5/93

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CONTAMINATION CONTROL
                                             NOTES
             Soil
(MCiorpCi/g) =
microcuries or picocuries
per gram of soil
II.     Surface Contamination
       A.    Regulatory Guide 1.86 - Termination of Operating
             License for Nuclear Reactors

             1.     This document presents the guidance that
                    is normally referenced when contamination
                    limits are needed (Table 1).

             2.     The limits apply to contamination on
                    personnel, personal items, tools,
                    equipment, and large surfaces areas (i.e.,
                    floors).

       B.    Contamination Identification - Original Problem

             1.     Identification starts with the recognition of
                    a possible problem.

                    a.     spills

                    b.     leaks

                    c.     maintenance activities  that require
                          the opening of a contaminated
                          system

                    d.     sampling exercises

             2.      Identification continues  with surveying the
                    area:

                    a.     smear survey - for loose surface
                          contamination
5/93

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  CONTAMINATION CONTROL
NOTES
                              ACCEPTABLE SURFACE CONTAMINATION LEVELS
NUCLIDE3
U-n»t.U-235,U-23E.and
associated dec»y products
Transuumcs. Ra-226, Ri-228.
Th.230,Th-226.Pa-231.
Ac-227, 1-125, 1-129
Th-nat,Th-232.Sr-90,
Ra-223,Ri-22<.U-232,
1-126, 1-131, 1-133
Ben -gamma emitters (nuclides
with decay modes other than alpha
emission or spontaneous fission)
except Sr-90 »nd others noted above.
AVERACEb c
5. 000 dpm a/ 100 cm2
1 00 dpm/ 100 cm2
1000 dpm/ 100 cm2
5000 dpm P'r/ 100 cm2
MAXlMUMbd
15. 000 dpm a/ 100 cm2
300 dpm/ 1 00 cm2
3 000 dpm/ 100 cm2
IS.OOOdpmp-r/lOOcm2
REMOVABLE6 *
1.000 dpm a/ 100 cm2
20 dpm/ 100 cm2
200 dpm/ 100 cm2
1000 dpm 0-7/1 00 cm
'Where  surface  conuminslion by  both  alphi-  and bcu-famm»-emillin[ nuclidei exisli,  the limiu  established  lor alpht-  and
 bei>-tamm>-emiuin( nuclidei should apply independently.
bM uvcd in Ihu lablt, dpm (diunicpationi per minuie) means the rale of emiiuon by radioactive maicnaJ ai determined by correctint
 the counii per nunuic obiervtd by an appropriate delector for background, efficiency, and geometric faciort associated with the
 insim mentation.
cMeaturemems of average  contaminant should  not be averaged over more thin 1 tquart meici. For objecu of leu surface area, the
 average should be oenved (or each such object.
 The maximum contamination level applies to an area of nol more than 100 cm .
The amount of removable radioactive maleriaJ  per 100 cm^ of surface arei should be 6e\trmii>c<3 by wipm; that a/ei with dry filter or
 toft abiorbent papei, applying moderate pressure, and atseninf the amount of radioactive material or> the wipe with an appropriate
 instrument of known efficiency.  When removable contamination on objects of  less surface  uu  is determined, the pertineni leveU
 should be reduced proportionally and the enure surface should be wiped.
                                                TABLE 1
                             EXCERPT:  REGULATORY GUIDE 1.86
  5/93

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CONTAMINATION CONTROL
                                                                NOTES
      C.
             3.
       b.     direct survey with an instrument
             for both loose and fixed
             contamination; this is called
             "frisking"

       Once contamination is identified or
       suspected, the area is "posted"
       a.     ropes

       b.     signs

       c.     step-off pads

Contamination Identification - Problem Control

1.     Direct survey (frisking) - whole body,
       hands and feet, or equipment:

       a.     use a count rate meter instrument
             (beta-gamma or alpha)

       b.     hold  the probe close to the surface
             between one quarter to one half an
             inch  away

       c.     move the probe slowly,  about 1 to
             2 inches per second

       d.     observe the meter for any  increase
             in count rate

2.     Automatic Contamination Monitors

       a.     timed counter "gates"

       b.     walk-through monitors

       c.     computer counters
5/93

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CONTAMINATION CONTROL
NOTES
             3.     Removal of Material from a Contaminated
                    Area

                    a.      all material leaving a contaminated
                           area must be treated as
                           contaminated until proven otherwise

                    b.      bag the item and move it to the
                           place where it will be  dealt with,
                           (e.g.,  storage or decon)

                    c.      to release the item  for unrestricted
                           use, it must be surveyed for both
                           loose and fixed  surface
                           contamination

             4.     Documentation

                    a.      All surveys should be  properly
                           recorded

                    b.      All contamination events should be
                           recorded including an  investigation
                           to determine root cause. This is
                           the only way that recurrence can be
                           prevented.

       D.    Radioactive Waste

              1.     Most radioactive waste is  actually material
                    that is contaminated with radioactive
                    deposits.

                    Example:  A pipe wrench might be used in
                    a contaminated area. The teeth of the
                    wrench could get contamination embedded
                    in them. If that happens,  the entire
                    wrench is not radioactive.  The only thing
                    that is radioactive is the radioactive
                    material deposited on the teeth. But, the
                    entire wrench must be  handled as
                    radioactive until these deposits are
                    removed, (i.e., decontaminated).
 5/93

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CONTAMINATION CONTROL
NOTES
                   Consequently, if the wrench cannot be
                   decontaminated, and it is to be disposed
                   of, the entire wrench is considered
                   "radioactive waste."

             2.     Radwaste Reduction Techniques

                   a.     limit the entry of tools, equipment,
                         and packing material into
                         potentially contaminated areas

                   b.     keep contaminated and non-
                         contaminated materials separate

                   c.     reuse tools and equipment with
                         fixed contamination (make up a
                         contaminated tool crib)

                   d.     establish a sorting and segregation
                         practice that reduces the volume of
                         the radioactive waste (Figure  1)
                       FIGURE 1
                 CONTAMINATED RAG
5/93

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CONTAMINATION CONTROL
                                                         NOTES
4
             3.
Steps for sorting and segregating the cloth

       1.     Survey the article (A)

       2.     Locate the contamination
             (PartB)

       3.     Remove the contaminated
             section (cut on line C)

       4.     Dispose of noncontaminated
             section as regular waste
             (Part A)

       5.     Dispose of contaminated
             section as radioactive waste
             (PartB)

Liquids and wet material should be kept
segregated  from dry contaminated waste.

a.     all radioactive waste has to be
       totally dry for disposal

b.     liquids and wet materials have to be
       either solidified or dried
 III.    Airborne Contamination
       A.    Definitions
              1.     Airborne contamination - radioactive
                    material in the air that exceeds
                    the limits established in 10 CFR 20,
                    Appendix B

                    a.     paniculate

                    b.     gaseous
 5/93

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CONTAMINATION CONTROL
                                                          NOTES
             2.     Internal Exposure - exposure to radiation
                    from radioactive material which is
                    deposited inside of the body

       B.     Types of Airborne Contamination
             1.
             2.
Paniculate - radioactive material that exists
in a solid state at standard temperature and
pressure (STP). Ex: 60Co or 90Sr
                    a.
                    b.
       usually bound up in some form of a
       metal complex or salt (cobalt oxide
       or radium bromide)

       ability to go airborne is directly
       related to the physical medium it is
       in

       i.     dry and dusty

       ii.     water
                          m.
                                 oil
c.     can be removed from the air by
       high-efficiency filtration

Gaseous - any radioactive material that
exists in a gas state at STP   Ex:  131I, 133I

a.     Reactive gas (iodine)

       i.     can be removed from the air
             by filtering through an
             activated charcoal medium

b.     Noble gas (radon, krypton)

       i.     cannot be removed from the
             air in this form
5/93

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CONTAMINATION CONTROL
                                       NOTES
       C.     Regulations

                      1.      29 CFR 1910.96 (c) (Figure 2)
           1)  No employer shall  possess, use, or transport
           adioactive material in such a manner as to cause
          any  employee,  within  a  restricted area,  to be
          exposed  to  airborne radioactive  material  in an
          average  concentration  in  excess of the  limits
          specified in Table 1 of Appendix B to 10 CFR Part
          20.  The limits given in Table 1 are for exposure to
          the concentrations specified for  40 hours  in any
          workweek of 7 consecutive  days.  In any such
          period  where the number of hours of exposure is
           ess than 40, the limits specified in the table may be
           ncreased proportionately. In any such period where
          the number of hours of exposure is greater than 40,
          the limits specified in the table shall be decreased
          proportionately.

          (2)  No  employer  shall  possess, use,  or transfer
          radioactive material in such a manner as to cause
          any individual within a restricted area, who is under
          18  years of age,  to  be exposed to airborne
          radioactive material in an average concentration in
          excess of the limits specified in Table II of Appendix
          B  to 10  CFR  Part 20.   For purposes of this
          subparagraph, concentrations may be averaged over
          periods not greater than 1  week.

          (3)  "Exposed" as used in this paragraph means tha
          the   individual   is  present   in   an  airborne
          concentration. No allowance shall be made for the
          use of protective clothing or equipment, or particle
          size.
                           FIGURE 2
                EXCERPT: 29 CFR 1910.96 (c)
 5/93
10

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 CONTAMINATION CONTROL
                                   NOTES
                    2.     10 CFR 20, Appendix B
                          (Table 2)


           Port 20, App. B                        10 CFR Ch. I (1-1-91 .Edition)

            APPENDIX B—CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND—Continued
                              (S« loolnolM *! KXJ ol Apptndu B)
EIWMfll (llomic numtwf)
* 1
C*kiornwm (98) 	 . 	
Cftrtxxi (6) 	 „ 	 „
C«rium (58) ....
CvtXxn (5S) 	 	 „ 	 _ 	 —
Chlofift* (17) 	 _ 	
CtvoTMum (24) 	
Cototll (27) 	 	 _ 	 	
CoppM (29) 	
Cooom(9«) 	 	
liolcxx '

Ct 249 	
Cr 250
Cl 251
Cl 252 .. .
Cl 253
Cl 254 	 _..
C 14 	
(CO,)
C* 141 	
Ce 143 	
Ce 144 	
C» 131 	 	
Cs 134m. 	 	
Cs 134 	
C* 135., 	
Cs 136 	
Cs 137
Cl 36 	
Cl 36 	
Cr 51 	
Co 57_ 	
Co Mm 	 	
Co 58... 	
Co 60 	 _ 	
Cu S< 	
Cm 242 	
Cm 243 	
Cm 244 	
Cm 245 	
Cm 246
Cm 247 	
Cm 248 	
Cm 249 .. .

1
S
1
S
1
S
1
s
1
s
1
s
1
s
Suo
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
1
s
T«ol«l
Col 1-A«
did/ml)
2
2
1
5
1
2
1
6
3
g
s
5
5
4
5
4
2
3
2
1
6
1
3
4
4
1
9
4
2
6
1
4
2
3
2
1
2
3
2
9
5
9
1
1
2
C
1
9
1
5
1
S
1
5
t
6
1
1
1
ID'1
!0"-'
10"-
10'"
10"-
10'"
to- ••
10-"
10'"
10"-
!0"-
10'"
10"'
to-
10-
10-
10
10
10
10
10-
to-
to-
to-
10"
10"
to-
to-
10
to-
10
to-
10-
10
10-
10
10
to-
10"
10"
10"
10"
10"
10 '
10"
10"
10"
10"
10 •
10 '•
10 '•
10 "
to '•
10 "
10"-
10"'
10"-
10 "
10"-
10 "
10 '-
10 "
10 "
10 •
10 '
Cd. 2-
W>l«
I^D/ml)
1
t
7
4
7
2
2
4
4
4
4
2
3
3
1
1
3
7
3
2
3
1
2
1
1
3
1
6
7
7
2
1
1
e
i
i
4
6
6
10'
10"
10'
10"
10'
10'
to-
to-
10"
«T
10-
to-
io-
to-
10
10
to
10-
10-
10-
10-
to-
10-
10-
10-
1C'
10-
10
to-
to-
to-
10
10
10-
to-
10
to-
10
to-
10"
10
10
10
10-
10'
10
to-
10
10
10
10
10
to
10
10-
10
10
to
to
10
10
10
10
10
T«bl« II
Col 1-A»
(lid/ml)
6- 10"
5 -lO"'
3 10"'
J to"'
3 10"'
6 10"'
3 f. 10- "
2>'10"'
1 ^10-"
3. -10"'
3 'MO"1
2.-10"1
2- 10'"
1 . 1C'1
1 . 10"
2. 10"
5. 10 •
9. 10"
7 . 10"
3. 10- '•
2- 10"-
4 . to-'
1 . 10-'
1 . 10"
2. 10-'
1 . 10"
4. 10"-
2. 10"
3. 10"
1 . 10"
6. 10-'
2. 10"
5- 10"-
1 . 10"
g • 10 ••
9. 10"
7 . 10"
4 . 10"
g. 10"
1 . 10"
6. 10"
6. 10"
3. 10-'
3. 10"
2. 10-
1 • 10-
3. 10"
7. 10-
4 . 10
4 . 10 '
6. 10 '
2. 10' '
3. 10- '
3 . 10- •
3. 10 '
2. 10 •
4 . 10' •
2. 10 •
4 . 10 '
2 . 10' '
4 . 10' '
2. 10 '
4 . 10 '
4 . 10 '
4 . 10 '
Col. 2-
W»lef
(^Ci/ml)
3 .- 10-
4> 10-
2- 10-
1 / 10'
3>'10'
4f 10"
3x10"
7x10-
7x10"
t mo-
1X10"
Iv 10-
1x1.0'
g •: to-
bbbbbbbbobbbbbbbbbbbbbbbbbbobbboooobooooDooooooooo
5/93
                                    TABLE 2
                      EXCERPT: 10 CFR 20, APPENDIX B
11

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CONTAMINATION CONTROL
                                                                    NOTES
        D.     Indicators of an unposted airborne radioactive
               materials area
               a.
leaks
               b.      dusty  conditions
IV.    Soil Contamination

        A.      NRC Policy and Guidance Directive FC 83-23:
                Termination of By-product, Source and Special
                Nuclear Material Licenses

                1.      This document provides guidance on
                       acceptable soil contamination level
                       (Table 3).
                       Acceptable Soil Contamination Levels
              Kind of Material
          1)  natural Uranium (U-23B +
             U-Z34) with daughters present
             and 1n. equilibrium

         11)  Depleted Uranium or Hatural
             Uranium that has been separateJ
             from Its daughter; Soluble o.
             Insoluble

        111}  Natural Thorium (Th-232 + TK-2ZB)
             with daughters present and 1n
             equilibrium

         1v)  Enriched Uranium Soluble or
             Insoluble

          v)  Plutonium (Y) or (W) compounds

         v1)  Am-241 (W) compounds

        v11)  'All  Byproduct Material
        v111) External Radiation
                                         SoD Concentration Level
                                          for unrestricted ares
                      10 (pC1/gi cf soil)
                      35 (pCI/gn of soil)
                      10 (pd/gm of soil)

                      30 (pCi/gm of soil)


                      25 (pC1/oro of soil)

                      30 (pd/gra of soil)

                      Soil  concentrations
                        shall be determined
                        .on a case by case
                        basis

                      10 m1croroentgens/hr
                        above background
                        measured at one
                        meter from the
                        ground surface
                              TABLE 3
        EXCERPT:  NRC POLICY AND GUIDANCE
                      DIRECTIVE FC 83-23
 5/93
                          12

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CONTAMINATION CONTROL
                                 NOTES
V.    Water Contamination Levels

      A.    Surface and groundwater contamination levels
            should be below EPA's National Primary
            Drinking Water Regulations (40 CFR 141).
5/93
13

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                       SECTION 10

       ANTICONTAMINATION CLOTHING AND
 RESPIRATORY PROTECTION DEVICE EQUIPMENT
         After completing this unit, participants will be able to:

         •    Explain the limitation of anticontamination protective
              clothing.

         •    Describe one example of a radiologically engineered
              contamination control device.

         •    List two respiratory protection devices and under
              what circumstances they would be worn.
5/93

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ANTICONTAMINATION CLOTHING AND
RESPIRATORY PROTECTION DEVICE EQUIPMENT
NOTES
I.     Introduction

      The purpose of wearing anticontamination clothing is to
      prevent an individual from getting contamination on
      them.

      A.     Types of Clothing

             1.     General - The general practice is to not
                   wear personal items with anti-
                   contamination clothing

             2.     Modesty Garments

             3.     Gloves

                         cotton liners"
                         surgical gloves
                         rubber gloves
                         work gloves

             4.     Shoe Covers

                         toe/heel covers
                         booties (plastic/cloth)
                         rubber shoe covers
                         boot (low tops or waders)

             5.     Coveralls (one or two piece, hooded or
                   nonhooded)

                         cotton (zipper/velcro/drawstring
                         paper
                         plastic (zipper/button/drawstring)

             6.     Laboratory Coats

             7.     Head Covers

                         hoods
                         skull caps
5/93

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ANTICONTAMESATION CLOTHING AND
RESPIRATORY PROTECTION DEVICE EQUIPMENT
NOTES
      B.     Selection Process

             Note:  When you dress out for this type of work,
                   you may be dressed out for very long
                   periods of time. The affects of heat stress
                   and comfort of the worker must be
                   considered.

             1.     Factors to consider

                   a.     Levels of contamination

                                low
                                medium
                                high

                   b.     Location of contamination

                                localized or widely
                                dispersed
                                floors
                                walls
                                overheads
                                equipment

                   c.     Body position relative to the
                         contamination location

                                walking
                                kneeling
                                crawling
                                under contaminated objects

                   d.     Work to be performed

                                light and easy
                                physically demanding

                   e.     Is wet or will it become wet

                                feet
                                body
                                hands
5/93

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ANTICONTAMEVATION CLOTHING AND
RESPIRATORY PROTECTION DEVICE EQUIPMENT


      C.    Donning and Doffing of Anti-C's

            1.    Donning Steps

                  a.    plastic shoe covers

                  b.    cotton liners

                  c.    coveralls

                  d.    plastic shoe covers

                  e.    rubber shoe covers

                  f.    rubber gloves

                  g.    tape

                  h.    hood

            2.    Doffing Steps

                  a.    tape

                  b.    rubber shoe covers

                  c.    rubber gloves

                  d.    hood

                  e.    plastic shoe covers

                  f.    coveralls

                  g.    plastic shoe covers

                  h.    cotton liners

       D.    General Practices Performed in Anti-C's

                   1.     Dosimetry should be kept on the
                         breast pocket of the anti-C's
 5/93
NOTES
          4

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ANTICONTAMCVATION CLOTHING AND
RESPIRATORY PROTECTION DEVICE EQUIPMENT
                                                        NOTES
                  2.
         Note:  If entering an area with
               significant radiation levels, a
               second pocket dosimeter
               should be worn on the
               outside for convenient
               monitoring.

         If at any time protective clothing is
         torn or an individual is injured by
         any type of penetration  (such as a
         minor puncture wound), the
         individual must exit the area
         immediately and have the area of
         the penetration checked.
II.    Respiratory Protection Devices

      A.    Regulations, Guidelines and Standards

            1.    OSHA 29 CFR 1910.134

            2.    ANSI Z88.2-1969 and ANSI Z88.2-1980

            3.    NIOSH

            4.    NRC 10 CFR 20 listed Assign Protection
                  Factors (APF)
                  a.
                  b.
                  c.
                  d.
         APR
         Airline
         Bubble hood
         SCBA
        APF =
Concentration outside mask
 Concentration inside mask
50

1,000

1,000

10,000
5/93

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ANTICONTAMINATION CLOTHING AND
RESPIRATORY PROTECTION DEVICE EQUIPMENT
NOTES
            2.     Concentrations Limiting Respirator Use

                   a.     Maximum use concentration
                         (MUC)- The highest concentration,
                         not exceeding IDLH
                         concentrations, of a specific
                         contaminant in which a given style
                         respirator can be worn.

                   b.     Maximum permissible
                         concentration (MFC)-  The
                         maximum concentration of
                         radioactive contaminant in air that
                         an individual can be exposed to for
                         a 13 week period without exceeding
                         their quarterly dose limit.

                   MUC  = APF x MFC
      B.    Display and Discuss Respirators

            1.     APR

                   a.     High-efficiency paniculate air filter
                         (99.97% efficient for 0.03-micron
                         particles)

            2.     Airline

                   a.     300 feet maximum hose length

                   b.     tangles very easily

                   c.     more comfortable than APR

            3.     Bubble Hood

                   a.     same hose problems as the airline

                   b.     the most comfortable with the best
                         vision
5/93

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ANTICONTAMINATION CLOTHING AND
RESPIRATORY PROTECTION DEVICE EQUIPMENT
                                                                NOTES
III.
            4.
            SCBA
                  a.
                  most reliable
            b.     heavy

            c.     limited air supply


Engineering Controls

A.    Containment Devices

      1.     Catch containers

      2.     Glove bags

            a.     simple

            b.     complex

      3.     Dog houses

            a.     small work

            b.     sorting tables

      4.     Tents

            a.     temporary

            b.     permanent enclosures

B.    Ventilation Systems

      1.     HEPA - 99.97% efficiency

      2.     Charcoal absorber - normally used for the
            removal of iodine gas
5/93

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                        SECTION 11

        DEMONSTRATION OF RADIOLOGICAL
                     CONTROL AREA
         After completing this unit, participants will be able to:

         •    Explain the purpose of a radiological control area.

         •    List the equipment that is used in a radiological
              control area.

         •    Describe the entry procedure for a radiological
              control area.
5/93

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 DEMONSTRATION OF RADIOLOGICAL CONTROL AREA



 I.     Preparation for Entry

       A.     Control Point Setup

              1.     ropes

              2.     signs

              3.     step-off pads

              4.     frisking station

              5.     waste containers

              6.     decon area with supplies

              7.     bags

       B.     Counting Area Setup

              1.     Equipment Setup

                    a.     selection of:
                          alpha frisker
                          beta-gamma frisker
                          alpha bench counter
                          beta-gamma bench counter

                    b.     instrument checks:
                          battery check
                          calibration sticker check
                          efficiency determination

             2.     Area Layout

                    a.     smear/air sample handling area

                    b.     clean area for maps and notes

                    c.     clean waste bags

                    d.     contaminated waste bags


5/93                                      2
NOTES
          4

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DEMONSTRATION OF RADIOLOGICAL CONTROL AREA
                                                             NOTES
      C.
       D.
Survey Preparation

1.     Map

      a.    already drawn, or

      b.    materials ready to draw map

2.     Contamination Survey Instruments

      a.    smears numbered

      b.    envelopes prepared

3,     Radiation Survey Instrument

      a.    instrument selected

            i.     battery check

            ii.     calibration

            iii.    response check

4.     Air Sampler

       a.    loaded for sample

       b.    envelope prepared
                   c.
 Anti-C's
             bag ready for removal of air
             sampler from the contaminated area
             1.    Selection

             2.    Donning Steps

                   a.     plastic shoe covers

                   b.     cotton gloves
 5/93

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DEMONSTRATION OF RADIOLOGICAL CONTROL AREA
                                                                            NOTES
                          coveralls

                          plastic shoe covers

                          heavy shoe covers

                          rubber gloves

                          tape

                          hood
                   c.

                   d.

                   e.

                   f.

                   g-

                   h.


II.    Entry

      A.     Air Sample
             1.     Set up air sampler on clean cloth/plastic
                    away from the floor

             2.     Start air sampler and record the time

             3.     Secure air sampler after 5 minutes and
                    record the time

             4.     Remove the air sample and place it in an
                    envelope

       B.    Radiation Survey

             1.     Dose rates are taken first as a worker
                    moves around the room

             2.     Record dose rates on survey map

                    a.     general area dose rates are taken 12
                          inches from any surface and at
                          waist level

                    b.     contact dose rates are marked with
                          an asterisk (*)
5/93

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DEMONSTRATION OF RADIOLOGICAL CONTROL AREA
NOTES
             3.     Place the instrument in the lay-down area
                   identified as "to be surveyed" if
                   contamination is not expected

      C.     Contamination Survey

             1.     Take —100 cm2 smears

             2.     Mark location on the map with
                   sequentially numbered circles


             3.     Place all smears in an envelope

      D.     Tool Decon

             1.     Simple decon


III.   Exiting From a Contaminated Area

      A.     Air Sampler Removal

             1.     Wipe down the air sampler if
                   contamination is expected

             2.     Either set the sampler in the area "to be
                   surveyed" or put it in a bag

      B.     Survey Equipment Removal

             1.     Place smear envelope into a bag

             2.     Place survey map into a bag

      C.     Remove Anti-C's

             1.     tape

             2.     rubber shoe covers

             3.     rubber gloves
5/93

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DEMONSTRATION OF RADIOLOGICAL CONTROL AREA



            4.     hood

            5.     coveralls

            6.     plastic shoe covers

            7.     cotton liners

      D.    Frisk


IV.   Count Area Demo

      A.    Frisk Smears
                                                          NOTES
      B.
            1.

            2.
      Alpha

      Beta
Count Air Sample

1.     Cut sample

2.     Count for both alpha and beta

3.     Calculate curie content

Survey the equipment that is to be removed from
the contaminated area
5/93

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                          SECTION 12

                   DECONTAMINATION
          After completing this unit, participants will be able to:

          •    Explain the need for radiological decontamination.

          •    Define the types of radiological contamination.

          •    Compare  different  decontamination  methods  and
               techniques.
5/93

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 DECONTAMINATION
                                                                       NOTES
 I.
Definition:
              Radioactive decontamination can be defined as the
              systematic removal of radioactive contamination.
II.     Define the purpose of radioactive decontamination  as  a
       process of contamination control.

       A.     Decontamination has three purposes:

              1.     To  prevent  any  uptake  of  radioactive
                    material into the human body.

              2.     To limit external radiation exposure.

              3.     To   prevent   the   further   spread   of
                    contamination.
III.    Categories of radioactive contamination:

       A.    Personnel

       B.    Equipment and Material

       C.    Air

       D.    Water

       E.    Soil
IV.    Normal decontamination efforts usually apply to personnel,
       equipment,  and material contamination.   There are two
       types   of  contamination  that  are  of  concern   for
       decontamination:

       A.    Fixed contamination.

              1.     Contamination   not  detected  by   smear
                    survey.
5/93

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DECONTAMINATION
NOTES
             2.     Contamination that cannot  be removed by
                    normal washing.

             3.     Removal   normally   requires   chemical
                    solvents or mechanical applications.

       B.     Loose surface contamination.

             1.     Most decontamination efforts will be in this
                    category.

             2.     Can be detected by smear survey.

             3.     Can usually be removed by normal washing.

             4.     Does not require mechanical applications.


V.     General steps for decontamination.

       A.     Evaluate project and needs.

             1.     Determine what is to be decontaminated.

                    a.     Decontamination may be required for
                          tools, equipment,  materials,  work
                          areas, clothing, and personnel.

             2.     Determine   why   decontamination   is
                    necessary.

                    a.     Will result  in a cleaner area.

                    b.     Reduction in the use of protective
                          clothing and respiratory equipment.

                    c.     Reduced  inventory of contaminated
                          tools and materials.

                    d.     Reduction in accumulated personnel
                          exposures.

                    e.     Removal of personnel contamination.
5/93

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DECONTAMINATION
                                                    NOTES
                    f.
Release to unrestricted areas.
                    The  very   nature   of  decontamination
                    generates radioactive waste.

                    a.     Minimize the use of water and
                           materials.

                    b.     Cleaning solutions, cloths, or other
                           materials  actually   used  in  the
                           decontamination  process  must  be
                           disposed of as radioactive waste.

                    c.     Consider  waste  generation  and
                           disposal   prior   to    undertaking
                           decontamination  operations.

                    Establish  decontamination boundaries  and
                    control access to the area.

                    a.     Boundaries should allow for adequate
                           work   space  and  prevent  other
                           personnel from entering
                           decontamination  work area.

                    b.     Access  to and from  the work area
                           can be  controlled by  setting  up an
                           entry/exit control point.

                    Investigate how contamination occurred and
                    review survey data.

                    a.     May aid in  knowing  what kind of
                           radioactive materials you are dealing
                           with.

                    b.     Help determine needed
                           instrumentation.
                    c.     Help determine  protective clothing
                           and equipment requirements.
5/93

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DECONTAMINATION
                                                    NOTES
                    d.
                    e.
Help   to
priorities.
set   decontamination
Help determine corrective actions to
prevent future recurrence.
             6.     Evaluate  specific  items  or areas  to  be
                    decontaminated.

                    a.     This   will  aid  in  selection  of
                          de ontamination   equipment   and
                          materials.

             7.     Assess   radiation   and   contamination
                    protection  requirements   for   personnel
                    involved in decontamination efforts.

                    a.     The main  concern is  to  protect
                          personnel from becoming
                          contaminated and to keep radiation
                          exposures  as  low  as  reasonably
                          achievable (ALARA).

                    b.     Personnel performing
                          decontamination  should  take  all
                          necessary  precautions  to  protect
                          themselves.

       B.     Obtain equipment and materials.

             1.     Use  the  information  gathered  in  project
                    evaluation to aid in equipment and material
                    needs.

             2.     Select appropriate radiation and
                    contamination detection instruments.

                    a.     Selection shall be based on the type
                          of contaminants involved.

                          i.      alpha probe for alpha
                                 contamination.
5/93

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DECONTAMINATION
                                                    NOTES
                           ii.     beta-gamma probe for beta-
                                 gamma contamination.

                    b.     If needed, use a dose rate instrument
                           to monitor external radiation
                           exposure.

                           i.     personnel   dosimeters   may
                                 also be desired.

                    Select  appropriate protective clothing and
                    equipment  to  perform   decontamination
                    efforts.  Selection should be based on:

                    a.     Type of contaminants.

                    b.     Physical condition of the
                           contaminated material (i.e., is it wet
                           or dry).

                    c.     Extent of contamination.

                    d.     Decontamination methods used.

                    e.     Any other  considerations needed to
                           help prevent personnel
                           contamination.

                    Select appropriate decontamination
                    method(s) and obtain needed  equipment.
                    a.
                    b.
                    c.
Based on what is to be
decontaminated.

Methods may be used individually or
in combination.

Guidance  for selection of methods
can  be  found in  Chapter  7  of
DARCOM  P385-1,   Table   7.4,
located  in the reference  section of
this manual.
       C.    Start mild.
5/93

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 DECONTAMINATION
NOTES
              1.     Select the least harsh or abrasive  method
                    suitable for what needs to be decontaminated
                    before employing the more drastic methods.

                    a.     Adhesive  tape or  mild soap  and
                           water are good  starting points  for
                           most loose surface decontamination
                           efforts.

              2.     When more than one method is employed,
                    start with the least harsh or abrasive method
                    first.

       D.     Work from outside to inside.

              1.     Decontamination  normally  begins  at   the
                    perimeter  of a  contaminated  area  and
                    progresses toward the center.

                    a.     Stroking motions toward the center,
                           rather than  side to side scrubbing
                           motions,   tend  to  help  prevent
                           spreading the contamination.

              2.     Care  should be exercised to ensure excess
                    water is not used.

                    a.     Water  has  a  tendency  to  run,
                           possibly spreading the contamination
                           to other areas of the surface.

              3.     Perimeters  should   be   surveyed   and
                    reestablished as the size of the contaminated
                    area is reduced.  Document results.

              4.     When  decontamination  involves  vertical
                    surfaces   (i.e.,   walls),   decontamination
                    should start at the top and progress toward
                    the bottom.

       E.     Clean highest contamination first.
5/93

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DECONTAMINATION
                                                           NOTES
              1.
The idea is to attack the greater hazard first,
reducing it to fairly consistent levels with
the rest of the area to be decontaminated.

a.     If there is an area with a significant
       (gross) amount of contamination, that
       area should be cleaned up first.
                           i.
             2.


             3.

             4.
             use  the  same  technique  of
             starting at the perimeter and
             working  toward the center.
Will help lessen the degree of hazard in the
event of cross contamination.

In some cases will help control exposure.

Depending on  the  evaluation of the high
contamination area, you may elect to use a
different   decontamination  method   or
combination of methods.
       F.    Minimize contaminated waste.
                    All waste materials used in decontamination
                    are treated as contaminated waste; therefore,
                    use only what is needed to perform the tasks
                    and do not use materials excessively.

                    a.     Although some areas may need to be
                           cleaned    several   times   before
                           decontamination is complete, using
                           one applicator (i.e., rag) at a time
                           instead  of  several  will  help  in
                           reducing contaminated waste.

                           i.     be careful not to use the same
                                 applicator surface more than
                                 once.

                    Again, be careful not to use too much water
                    or liquids in decontamination.
5/93

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DECONTAMINATION
                                                    NOTES
                    c.
Liquids tend to run (spreading con-
tamination).

Liquids will ultimately generate more
contaminated waste.

There  should  be  no  free-standing
liquids in contaminated waste.
             3.     There are many concerns regarding storage
                    and disposit. jn of radioactive or contaminat-
                    ed wastes industry-wide.  The general prac-
                    tice  is to  take  all measures  practical to
                    reduce the generation of these wastes.

                    a.      Do not create unnecessary waste.

       G.    Minimize airborne contamination.

             1.     Decontamination methods may generate or
                    stir up loose contamination  that could be-
                    come airborne.  Caution should be taken to
                    prevent this from occurring whenever possi-
                    ble,  because it  will  only tend to create
                    another hazard which would  have to be
                    taken into consideration.

             2.     Be  conscientious  of your decontamination
                    techniques.

                    a.      Be careful  how you  move around.
                           Move slowly and try not to  stir up
                           loose contamination.

                    b.      Lay plastic  sheeting down whenever
                           practical  to help prevent airborne
                           contamination.

                    c.      Be careful during cleaning not to stir
                           up loose contamination.  Wipe care-
                           fully!!
5/93

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DECONTAMINATION
                                    NOTES
             3      In cases  where  it may be impractical  to
                    avoid the possible creation of airborne (i.e.,
                    grinding),  other  means   of   controlling
                    airborne contamination can be used.

                    a.    Portable HEPA filtered exhausters.

                          i.      help filter contaminants being
                                 generated.

                    b.    Containment tents and glove boxes.

                          i.      help  contain  contaminants
                                 being generated.

             4.     Be conscientious of area ventilation and air
                    movements.   Some air  movements may be
                    strong    enough  to   create   airborne
                    contamination.   In some cases, decreasing
                    the air  flow  will help prevent  airborne
                    contamination.

       H.    Survey between  major steps and document results.

             1.     Surveys and documentation between  major
                    steps is  vital to assessing how  well  the
                    decontamination  efforts are working.  The
                    data can be used:

                    a.    To verify cleanliness.

                    b.    For reclassification  or  rezoning  of
                          areas.

                    c.    For isolation.

                    d.    To support release to unrestricted
                          status.

             2.     Surveys between steps  can also be used to
                    measure the effectiveness of decontamination
                    methods.  A helpful tool in measuring the
                    effectiveness is the DECONTAMINATION
5/93
10

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DECONTAMINATION
                                    NOTES
                    FACTOR (DF), which can be calculated as
                    follows:

                    DF = Surface contamination before decon
                          Residual contamination after decon
                                or
                    DF = dpm before decon
                          dpm after decon

                    The higher the decontamination factor (DF),
                    the more effective the method.

             3.     Where extensive decontamination work is to
                    be   performed,   several   methods    or
                    combinations of methods can be tested  on
                    different areas of the same  surface and the
                    results   can   be  compared   using   the
                    decontamination factor.

       I.     Document completion.

             1.     Document the completion of
                    decontamination.

                    a.     Include standard information such as
                          name, date, time of completion, and
                          survey results.

                    b.     The  information  provided  on  the
                          survey  data  sheet  will serve  as
                          verification of cleanliness.

                    c.     Survey  results  will  be  used  in
                          classifying the area after
                          decontamination.

                          i.     If decontamination efforts are
                                proved to be effective,  the
                                area will most probably  be
                                downgraded   to   a  lesser
                                hazard  status or  released  as
                                unrestricted.
5/93
11

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DECONTAMINATION
                                    NOTES
                          ii.     Accordingly,   protective
                                 clothing   and   equipment
                                 requirements  may be  down-
                                 graded.

             2.     For  personnel  decontamination,   results
                    would show 1) that the exposure hazard to
                    the person no longer exists or 2) the extent
                    of exposure (if it takes considerable time for
                    the  contamination  to   be  completely
                    removed).
VI.    Personnel decontamination.

       A.    Observe physical effects.

             1.     Evaluate the  physical  condition  of  the
                    person(s) needing decontamination  for any
                    health-related problems.

                    a.      Check for burns, bleeding,  shock,
                           irregular   breathing,   and   life-
                           threatening   circumstances.     Get
                           medical  assistance immediately  if
                           needed.

       B.    Assess injuries.

             1.     Assess the extent of any injuries.

                    a.      Medical  treatment of injuries takes
                           priority over decontamination.

                    b.      Consult  immediately  with  medical
                           personnel if injury warrants  medical
                           attention.

                    c.      If it is determined that the injury is
                           minor,   decontamination   may
                           continue  with   medical   consent.
                           Remain  cautious  of  the  injury  in
                           question.
5/93
12

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DECONTAMINATION
                                    NOTES
             2.     Immediately  flush  with water  any  skin
                    contamination involving caustic,  corrosive,
                    or organic solvent solutions.

                    a.      Skin contact with these solutions will
                           require  medical  attention.     In
                           addition,  the solutions  will  cause
                           skin breakage, possibly spreading the
                           contamination  deeper into  human
                           tissue.

       C.    Survey person, document results.

             1.     Determine  the - extent  and  magnitude  of
                    contamination using personnel survey
                    techniques, and document results.

                    a.      During   the   survey,   particular
                           attention should be paid to locating
                           any hot spots of contamination.

                    b.      Will provide information needed to
                           set   a   baseline   prior   to
                           decontamination.

                    c.      Will provide information regarding
                           where the contamination is located
                           on the body.

                    d.      Will   help   set   priorities   for
                           performing decontamination.

             2.     The  results and  assessment  of the survey
                    should  be   recorded   on   a   Personnel
                    Contamination Record form.  An example of
                    this  form can be  found in  Chapter 7 of
                    DARCOM P385-1, Figure  7.1,  located in
                    the reference  section of this manual.

       D.    Remove contaminated clothing.
5/93
13

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DECONTAMINATION
                                             NOTES
             1.     Remove any contaminated clothing, place in
                    a  plastic  bag,  and  hold   for  further
                    disposition.

                    a.      Clothing must be removed carefully
                           to   help  control   the  spread   of
                           contamination.

                    b.      The clothing can  later be  used  to
                           help determine  how the personnel
                           contamination occurred.

                    c.      Help determine whether the clothing
                           requirements for the particular tasks
                           need to be reassessed.

                    d.      Help   prevent   possible   cross
                           contamination  from  the  clothing
                           during decontamination efforts.

                    e.      Help prevent any exposure that may
                           be  caused  by  the  contaminated
                           clothing.

             2.     The  contaminated  clothing will eventually
                    end up as contaminated waste.

       E.    Perform decontamination.

             1.     Various   procedures   for    personnel
                    decontamination of skin,  hair and scalp,
                    general body, face, eyes, ears,  mouth, and
                    nose  can  be  found  in  Chapter  7  of
                    DARCOM P385-1, Appendix 1, located in
                    the reference section of this manual. These
                    procedures have been found to be acceptable
                    practices in decontamination.
                    Note:
Medical personnel should be
consulted for decontamination
of  eyes,   ears,   nose,  or
mouth; or, if chemicals other
5/93
         14

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DECONTAMINATION
                                     NOTES
5/93
                                  than  soap  and water will be
                                  used.

                    Personnel  should  be decontaminated as
                    quickly as possible using the least drastic
                    means necessary.

                    a.     Begin  with  mild  methods such as
                           adhesive tape or mild soap and warm
                           water.

                    Soaps and detergents  emulsify and dissolve
                    contamination and are frequently all that is
                    needed.

                    a.     Continue  as  long   as   they   are
                           effective,  and progress  to  harsher
                           methods only if necessary.

                    Note:         Caution shall  be exercised to
                                  prevent excessive  skin irrita-
                                  tion.   Stop decontamination
                                  efforts  if  evidence of  skin
                                  damage appears or if person
                                  complains   of   soreness;
                                  contact medical personnel for
                                  assistance.

                                  Water temperature should be
                                  maintained   lukewarm   to
                                  avoid causing pores to open
                                  or  close.   The opening and
                                  closing of pores  can cause
                                  contamination   to  become
                                  embedded  in the skin.

                    Frisking shall be  conducted  after  each
                    attempt  to   reduce  contamination  until
                    contamination levels are acceptable.

                    a.     The  progress  of decontamination
                           should be  closely  monitored by
15

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DECONTAMINATION
                                                          NOTES
                          surveying   between   successive
                          washings or techniques.

                    b.     A log of methods used  and survey
                          results should be maintained.

                          i.     An example of a typical log
                                sheet for personnel
                                decontamination can be found
                                in Chapter 7 of DARCOM
                                P385-1,  Figure 7.2, located
                                in the  reference section  of
                                this  manual.

       F.     Investigate how contamination occurred.

             1.     A   thorough   investigation   of   the
                    contamination incident should be performed
                    as soon as possible.

                    a.     The results  of  the investigation
                          should show a valid cause of the
                          contamination, along with corrective
                          actions to prevent the reoccurrence
                          of the incident.

                          i.     There  are  a  number  of
                                reasons  why contamination
                                events occur, but the  main
                                issue becomes assessment of
                                corrective  controls for future
                                prevention.

                    b.     Document results  of the
                          investigation.
VII.   Equipment and material decontamination.

       A.    Equipment  and  materials   may   need  to   be
             decontaminated for  a number of reasons.   Some
             are:
             1.
Release for unrestricted use.
5/93
                      16

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DECONTAMINATION
                                                                  NOTES
       B.
2.     Salvage of valuable equipment.

3.     Reduce radiation exposure to personnel.

4.     Reduce the volume of contaminated waste.

Evaluate  the  need  for decontamination  versus
disposal or limited use.  Three examples are given
below:
              1.
             3.
       A highly contaminated area  may  warrant
       decontamination  because  its  use  would
       require frequent occupancy.

       a.     Decontamination in this case would
             minimize the collective exposures of
             personnel  expected to work  in these
             areas.

       Some equipment and materials are of signifi-
       cant   value   but   cannot   easily   be
       decontaminated or are not cost-effective to
       decontaminate.

       a.     Should be considered for limited use
             in normally contaminated areas until
             its use is no longer desired and final
             disposition is determined.

       b.     Expensive specialized tools may fall
             in this category.  They can be stored
             in a contaminated tool locker.

       Some equipment and materials are not of
       significant value  (low  replacement  cost
       materials) and are not easy or cost- effective
       to decontaminate.

       a.     May require more time and effort to
             decontaminate than they are  worth.

       b.     These materials should be considered
             for disposal as contaminated waste or
5/93
                             17

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DECONTAMINATION
                                   NOTES
                         assigned   for  use  in  normally
                         contaminated areas.

                   c.     Materials  such  as wood, clothing,
                         scrap  metal,  cords,   hoses,  and
                         damaged equipment may fall in this
                         category.

      C.     Many methods and techniques have been developed
             for decontaminating equipment and materials.

             1.     Most are physical or chemical processes.

             2.     In  Chapter  7  of  DARCOM  P385-1,
                   Appendix B (located in  the reference section
                   of  this  manual)   are  some  suggested
                   contamination removal  methods (Table 7.2)
                   and decontamination methods for  various
                   surfaces (Table 7.4).

             3.     Document contamination levels before and
                   after   each   application,   and   post
                   decontamination results.
5/93
18

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                        SECTION 13

       RADIOACTIVE MATERIAL PACKAGING,
               LABELING, AND SHIPPING
          After completing this unit, participants will be able to:

          •    Identify the federal regulations and/or organizations that
              govern the transport of radioactive materials.

          •    Given reference materials, determine package limits for
              individual radionuclides, mixtures of radionuclides, and
              material with unknown quantities of radionuclides.

          •    Given  the package radiation readings, indicate  the
              category of radioactive label required.
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                          NOTES
I.      Federal  regulators  and/or  organizations  that govern the
       transport of radioactive materials.

       A.    Interstate Commerce Commission  (ICC)

             1.     Established first regulations governing the
                    shipment of radioactive materials

             2.     Still exercise jurisdiction over the economic
                    aspects  of radioactive  materials transport
                    through the issuance of operating authorities
                    to carriers
       B.    Nuclear   Regulatory
             10 CFR 71
Commission   (NRC)
              1.     Responsibility for safety in the possession,
                    use, and transfer (including transport) of by-
                    product,   source,   and  special   nuclear
                    materials (Atomic Energy Act of 1954)

              2.     Promulgated 10 CFR Part 71; requirements
                    for licensees to deliver licensed materials to
                    a carrier for transport if fissile materials  or
                    quantities exceeding Type A are involved

              3.     The  NRC  assists  and  advises  DOT, has
                    adopted  portions  of DOT regulations by
                    reference,  and  inspects  its licensees  for
                    compliance with DOT regulations

              4.     Agreement States have  entered into formal
                    agreements with  the NRC for regulatory
                    authority over by-product, source, and less
                    than critical  quantities  of  special  nuclear
                    material

              Department   of  Transportation    (DOT)
              49 CFR 100-177

              1.     Regulatory responsibility for  safety in the
                    transportation  of all hazardous  materials,
 5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
NOTES
                   including radioactive materials (Department
                   of Transportation Act of 1966)

             2.    Promulgated   49   CFR  Parts  100-199;
                   jurisdiction includes all modes of transport
                   in interstate or foreign commerce, except for
                   postal shipments

       D.    Postal Service  - U.S. Postal Service Regulations,
             Part 124

             1.    Jurisdiction  over  all postal  shipments of
                   radioactive materials

             2.    Domestic Mail Manual, U.S.  Postal Service
                   Regulations, Part 124

             3.    Mailable quantity limits are generally one-
                   tenth of the values listed in DOT regulations

       E.     International Air Transport Association (IATA)

             1.    A body of air carriers that  publishes the
                   regulations for air  transport  of radioactive
                   materials  through  the  Dangerous Goods
                   Regulations
II.    Radioactive materials are defined in 49 CFR 173.403 as:

      A.     Those materials which spontaneously emit ionizing
             radiation and have a specific activity in excess of
             0.002 uCi/g

             NOTE:  Materials  not defined as radioactive  by
                                the DOT may be subject to
                                NRC or EPA regulations
III.  Classification of Radioactive Materials

      A.     Identification Requirements
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                  NOTES
                                                4
                   Identity of the radionuclides being  shipped
                   (if known)  and proper  shipping name  for
                   radioactive material as specified by  the
                   Hazardous  Materials  Table  (49  CFR
                   172.101; see Appendix I), in addition to the
                   following information:

                   a.      identifies the class of material
                   b.
                   c.
specifies  or
requirements
references packaging
lists identification numbers
                   d.     specifies   labeling  and   other
                          requirements

                          Note:  A  grid  of  the six proper
                          shipping   names   and   their
                          corresponding quantity Type is also
                          shown in Appendix I

                   Form  (degree of prepackaging)  of  the
                   radioactive material

                   a.     Special Form (49 CFR 173.403 (z))

                                Encapsulations  can only be
                                opened  by  destroying  the
                                capsule

                                At least one external physical
                                dimension must exceed 5 mm

                                Quantity   restrictions    for
                                "special  form"  materials are
                                generally larger than quantity
                                restrictions  for   "normal
                                form"   materials   (see
                                Figure 1)
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                                   NOTES
High Inlegrily
Massive Encapsulation
Solid Metal as a Sealed Source
X.;^^^:^ High Inlogrlly Weld
^rJ^felr Tantalum Inner
Stainless Sleel 	 j --""'"' Capsulo
Ouler Capsule 1 _*J 	 Radloisotope
VJy^-^il-, 	 High Inlegrily Welds

                      FIGURE 1
    SPECIAL FORM ENCAPSULATION EXAMPLES

                  b.    Normal Form (49 CFR 173.403 (s))

                              Any radioactive materials that
                              do not  qualify  as  "special
                              form" (see Figure 2)
          Wasle Material in
          Plastic Bag
                              Liquid in BolUe Within
                              Metal Container
     Powder in Glass
'£jl  Or
(-,J.M-JI  Plastic Bollle
                                       Gas in Cylinder
                      FIGURE 2
             NORMAL FORM EXAMPLES

      B. Quantity Determination for Radioactive Materials

            1.     A, and A2 System (49 CFR 173.433) is used
                  to determine package  limits based on the
                  radionuclide activity/quantity of radioactive
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                        NOTES
                   material (see  Appendix II for A, and A2
                   chart)
a.
                            and A2 System

                                Points   of   reference   for
                                activity   limitation   per
                                package  based  on    the
                                radionuclides present

                                Maximum activity in curies
                                for radionuclides that can be
                                transported  in  a  Type  A
                                package

                                Based on the radionuclide's
                                radiotoxicity

                                A, values apply to special
                                forms of radioactive material

                                A2 values apply to normal
                                forms of radioactive material

                                The  "ratio rule" is  used to
                                calculate the  A!  and  A2
                                values   for  mixtures   of
                                radionuclides
                   b.
             where:

             R = radionuclide (amount in
             Ci)

             X = At or A2 value for that
             radionuclide

       Limited Quantity, Instruments, and
       Articles (49 CFR 173.421)
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
NOTES
                             Very  small  quantities  of
                             activity

                             Examples:

                             medical test kits,
                             diagnostic aids,  some low-
                             level waste shipments,
                             button sources

                             Excepted  from   most
                             packaging   and  labeling
                             requirements

                             Activity limits per package,
                             instrument  or  article are
                             shown in Figures 3 and 4
LIMITED QUANTITIES
(package limits in curies)
SOLID LIQUID GAS
Special Form
Normal Form
Tritium
10-JA, lO-'A,
lO-'A, lO-'Aj lO-'A,
In paint or adsorbed on carrier 20
                     FIGURES
   PACKAGE LIMITS FOR "LIMITED QUANTITIES"
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
NOTES
INSTRUMENTS AND ARTICLES
(limiu in curio)


Special Form
Normal Form
Tritiated Water
Tritium Gaa
SOLID LIQUID
Item Package Item Package
10-2A, A,
10"JAj A, 10~SA, 10"'A2
Bated on Ci/liter
In paint or adsorbed on a solid carrier
GAS
Item Package
lO-'A, 10->A,
10-A, IO-A,

20 200
                      FIGURE 4
     LIMITS FOR INSTRUMENTS AND ARTICLES

                              Calculation:

                              Determine   the   limited
                              quantity   (LQ)   maximum
                              amount   per  package  for
                              normal   form  material
                              containing 129I

                              Solution:

                              A2 = 2 Ci
                              LQ Package Limit = 10'3

                              2 Ci x ID'3 = 0.02 Ci

            2.     Low  Specific  Activity  (LSA) (49 CFR
                  173.403 (n))

                  a.     Present a relatively low hazard as a
                        result  of limited concentration of
                        radioactive material

                              Examples:

                              Uranium or thorium  ores and
                              physical  or   chemical
                              concentrations of those ores
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                                      NOTES
                                Unirradiated   natural   or
                                depleted   uranium   or
                                unirradiated natural thorium

                                Tritium  oxide in  aqueous
                                solution,   provided   the
                                concentration does not exceed
                                5.0 mCi/mL

                                Nonradioactive materials that
                                are externally  contaminated
                                with    nondispersible
                                radioactive  material,   and
                                whose surface contamination
                                when averaged over 1 square
                                meter   does   not   exceed
                                0.0001 mCi/cm2

                          Mixtures are subject to the "ratio
                          rule"

                          Quantity limits are shown in Figure
                          5
           A; Value

         .05 Ci or less

        More than .05 Ci
          up to 1.0 Ci
                    Estimated Average
               Concentration does not exceed

                     0.0001 mCi/gram
                     0.005 mCi/gram

  more'than 1 Ci         0.3 mCi/gram

Nonradioactive material externally contaminated

     A2 Value        Allowable Concentration
        Less than .05 Ci

        .05 Ci or greater
                 0.0001 mCi/cm2 ol surface

                  0.001 mCi/cm2 ol surface
                        FIGURES
     LOW SPECDJIC ACTIVITY QUANTITY LIMITS
5/93

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                   VOTES
                   Type A quantity (49 CFR 173.435)

                   a.     Relatively small quantities of activity
                         that  are  less  than or  equal to the
                         appropriate Al or A2 values

                                Examples:

                                radiopharmaceuticals

                                research isotopes

                                industrial sources

                                Calculation:

                                Determine  the  maximum
                                activity for a Type A quantity
                                per package of normal  form
                                material containing 12*I

                                Solution:

                                A2 =  2  Ci,  therefore the
                                maximum activity  would be
                                2Ci

                   Type B quantity (10 CFR 71)

                   a.     Large  quantity  of  activity   that
                         exceeds  the  Aj  or  A2 where
                         appropriate  and  is  less  than the
                         Highway  Route-Control   Quantity
                         designation

                                Examples:

                                radiography sources

                                spent fuel shipments

                                high-level waste shipments
5/93
10

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                  NOTES
                               Calculation:

                               Determine the activity limits
                               for  a  Type B quantity  per
                               package  of  normal  form
                               material containing 129I

                               Solution:

                               A2 = 2 Ci
                               Highway   Route-Control
                               Quantity = 2 Ci x 3,000 =
                               6,000   Ci.  Therefore,  the
                               activity would be greater than
                               2 Ci but less than 6,000 Ci

                   Highway   Route-Control   Quantity
                   (49 CFR 173.403(1))

                   a.     Large  quantity of activity  that is
                         either  3,000  times the  A, or A2
                         value, or 30,000 curies, whichever is
                         the least

                               Calculation:

                               Determine the activity for  a
                               highway   route-control
                               quantity  per   package  of
                               normal  form   material
                               containing 129I

                               Solution:

                               A2 = 2 Ci
                               Highway   Route-Control
                               Quantity (HRCQ)
                               = 2 Ci x 3,000 = 6,000 Ci
                               HRCQ= 6,000 Ci, which is
                               less than 30,000 Ci

                   Quantity determination for mixtures
5/93
11

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                       NOTES
a.
                         The "ratio rule" is used to calculate
                         the A, and A2 values for mixtures of
                         radionuclides
                                            tf
                         where:

                         R = radionuclide (amount in Ci)

                         X   =   AI or  A2 value  for  that
                         radionuclide

                               Calculation:

                               Determine  the  Type  A
                               quantity amount for normal
                               form material containing a
                               mixture of 0.5 Ci  of 137Cs,
                               0.5 Ci of '"Co, and 0.2 Ci of
                               Solution:
             137
                                     A2
                                 Cs  10 Ci
                                     7 Ci
                                     0.4 Ci
                                 0.5    0.5   0.2
                                 -  +  - + -
               10
                                            0.4
                               0.62  Ci  is  less  than  1;
                               therefore,  the  mixture is a
                               Type A quantity

             7.     Reportable Quantity (49 CFR 172.101)
5/93
                     12

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                   NOTES
                   a.     Material   listed  as  a  hazardous
                         substance under Section  101(14) of
                         CERCLA

                                Example:

                                1  pound of radionuclides

                   Fissile Material (49 CFR 173.403  and
                   173.455)

                   a.     Material  that contains one or more
                         fissile  radionuclides  (233U,  235U,
                         238Pu, 239Pu,  or 241Pu)

                   Determination of Aj and  A2  Values for
                   Unknowns (49 CFR 173.433)

                   a.     Radioactive  material containing one
                         single unknown radionuclide

                                the value  of A, is 2 Ci and
                                the value of A2 is 0.002 Ci

                                if the atomic number of the
                                radionuclide is less than
                                82,  the  value   of  At  is
                                10 curies and the value of A2
                                is 0.4 curies

                   b.     Radioactive  materials with  known
                         radionuclide identities but unknown
                         individual radionuclide activities

                                same as "a" above

                                all   radionuclides   whose
                                individual activities  are not
                                known  (but  whose  total
                                activity is known)  must be
                                classified  in a single  group,
                                and the most restrictive A, or
                                A2 applicable to any  one of
5/93
13

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RADIOACTIVE MATERIAL PACKAGING.
LABELING, AND SHIPPING
NOTES
                                them  shall be  used  as the
                                value of  A,  or A2  in the
                                denominator of the fraction

                   c.     Radioactive materials  with  known
                          radionuclides but individual activities
                          unknown

                                the most restrictive value of
                                A,  or  A2 applicable to any
                                one  of   the  radionuclides
                                present is the applicable value

                   d.     Radioactive   materials   containing
                          unknown radionuclides

                                the value  of  A, is 2 curies
                                and the value of A2 is 0.002
                                curies

                                if alpha emitters are known
                                to be absent, the value of A2
                                is 0.4 curies

                   e.     Packages  containing  samples  for
                          which a reasonable doubt exists as to
                          its class and  labeling  requirements
                          and for which  a sample must be
                          transported for  laboratory analysis
                          may be labeled  according  to the
                          shipper's tentative class assignment
                          based upon (49 CFR 172.402):

                                defining criteria in 49 CFR
                                Parts 100  to 177

                                the   hazard   precedence
                                prescribed in 49 CFR 173.2

                                the shipper's  knowledge of
                                the material

IV.    Radioactive Material Packaging Requirements

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                   NOTES
      A.     Limited  Quantities Packaging Requirements  (49
             CFR 173.421)

             1.     Strong, tight packages that will not leak any
                   of   the   radioactive   materials   during
                   conditions   normally   incident   to
                   transportation

             2.     Radiation level at any point on the external
                   surface  of  the  package  cannot  exceed
                   0.5 mR/hr

             3.     The external surface of the package must be
                   free of significant removable contamination
                   (limits contained  in 49  CFR 173.443)

             4.     For instruments and articles, the radiation
                   level  at 4  inches from any  point on the
                   surface  (unpacked)  may  not  exceed 10
                   mR/hr

             5.     A description of the contents is in,  on, or
                   forwarded with the package

      B.     Low  Specific  Activity  Quantities   Packaging
             Requirements (49 CFR 403 (n))

             1.     Nonexclusive use shipments  -  "essentially
                   Type A packages"

             2.     Exclusive use - "strong, tight packages"

      C.     Type A Packaging Requirements

             1.     Designated for the containment of Type A
                   quantities

             2.     Shipper must make   an  assessment  and
                   certification of the particular package design
                   against performance requirements
5/93
15

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                  NOTES
             3.     General packaging requirements contained in
                   49  CFR  Sections 173.24,  173.411,  and
                   173.412

             4.     Typically performance-based DOT Specified
                   7A packaging (49 CFR 178.350)

             5.     Package must withstand normal conditions
                   of  transport without loss or dispersal of
                   radioactive contents (49 CFR 173.465)

             6.     Shipper of each DOT Specified 7A package
                   must maintain on  file  for at least 1 year
                   after  the  latest  shipment  a  complete
                   certification and supporting safety analysis
                   that:

                   i)     construction   methods,   packing
                          design, and materials of construction
                          are in compliance  with  49  CFR
                          173.461

                   ii)     all  the  requirements of  Sections
                          173.24,  173.463, and 173.465  are
                          met

       D.    Type B  Packaging Requirements

             1.    Designated for the containment of Type B
                   and High Route-Control Quantities

             2.    Must meet all of the performance standards
                   for Type A packages

             3.    Must  withstand  certain serious  accident
                   damage  test conditions  (10 CFR  71.73),
                   including:

                    a.     30-foot free drop onto an unyielding
                          surface
 5/93
16

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                  NOTES
                   b.
                   c.
                   d.
a puncture test, which is a free drop
(over 40  inches)  onto  a 6-inch
diameter steel pin

thermal exposure at 1,475 degrees
Fahrenheit for 30 minutes

water immersion for 8 hours
             Highway  Route-Control  Quantity   Packaging
             Requirements

             1.     Must meet all the performance standards for
                   Type B packages

             General Packaging Requirements (49 CFR 173.411)

             1.     The smallest  outside dimension  of any
                   radioactive materials package (other than
                   excepted quantities)

             2.     Package must have a tamper seal

             3.     Packages  must be  designed  to maintain
                   shielding efficiency and leak tightness under
                   conditions    normally    incident   to
                   transportation.     Internal  bracing   or
                   cushioning must be adequate.

             4.     External package surface  must be  easily
                   decontaminated

             5.     Characteristics  of  a  package  used  for
                   radioactive materials  in  a liquid  form
                   (173.412 (n)):

                   a.     Leak-resistant inner container

                   b.     Package must be adequate to prevent
                          loss or dispersal of the contents
5/93
               17

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                  NOTES
                   c.     Enough absorbent to absorb at least
                         twice  the  volume  of radioactive
                         liquid present

V.    Radiological Limits and Control for Radioactive Material
      Packages

      A.     Radiation limits for radioactive material packages
             (49 CFR 173.441)

             1.     Radiation level cannot exceed  200 mrem/hr
                   at any point on the external surface of the
                   package, and  the transport  index must not
                   exceed 10 mrem/hr  (dose rate at 1  meter
                   from surface of package)

             2.     Packages  exceeding the limit  must comply
                   with additional requirements

             3.     Contamination Control (49 CFR 173.443)

                   a.     The  level of nonfixed (removable)
                         radioactive  contamination on  the
                         external  surfaces of  the package
                         cannot exceed the limits as shown in
                         Figure 6
TABLE 10— REMOVABLE EXTERNAL
RADIOACTIVE CONTAMINATION— WIPE LIMITS
Conlaminanl
Beta-gamma emitting radionuclides; all
radionuclides with halt-lives less than
ten days; natural uranium; natural
thorium; uranium-235; uraniurn-236;
thorium-232; thorium-228 and Ihori-
um-230 when contained in ores or
physical concentrates 	
All other alpha emitting radionuclides 	
Maximum
permissible limits
uCI/cm'
10-'
10-'
dpm/
cm'
22
2.2

                        FIGURE 6
                EXCERPT: 49 CFR 173.443
5/93
18

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                          NOTES
             B.
b.     Radiological survey procedure

             Take a  swipe of a  300 cm2
             area on the exterior surfaces
             of the package

             Measure activity on  swipe
             sample and record in
             disintegrations per minute per
             square centimeter

             Perform radiation dose rate at
             package  surface  and  at  1
             meter from the surface of the
             package

Radiation  Limits for Highway Transport
Vehicles  of  Radioactive   Materials  (see
Appendix III)
VI.    Hazard Communications and Shipping Papers Requirements
       for Radioactive Material Packages

       A.    Marking and labeling requirements for radioactive
             material packages

             1.     Three types of radioactive labels (49 CFR
                    172.403  (c)) are selected  based on the
                    external dose rate of the package and/or
                    transport index (TI)

                    a.     Radioactive - White I

                    b.     Radioactive - Yellow II

                    c.     Radioactive - Yellow III

             2.     Required Label Information

                    a.     Identity of radionuclide(s)

                    b.     Activity in Ci
5/93
                      19

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                        NOTES
                   c.
      Dose rate and/or TI
             4.
TI (49 CFR 173.403 (bb))

a.     A number marked on the radioactive
      label that represents the maximum
      radiation level in mrem per hour at
      1 meter (3.3 feet) from the external
      surface of the package

b.     A number marked on the radioactive
      label of a package containing fissile
      materials that designates the degree
      of control exercised by the carrier to
      reduce the risk of a criticality

c.     The number of packages at any one
      location must be limited so  that the
      total TI  number (cumulative) does
      not exceed 50

A  chart  of labeling  requirements relative to
TI and fissile class is shown in Figure 7
5/93
                     20

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RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                        NOTES
     LABEL CATEGORY
     WHITE I

     YELLOW II



     YELLOW III
          While I
  MAXIMUM
PKG. SURFACE
 (MREM/HR)
TRANSPORT
  INDEX
(MREM/HR)   FISSILE DATA
    05

    50
                        200
                       Yellow II
                                   N.A.
                                    10
         CLASS I ONLY

         FISSILE I OR II.
        Tl OF 1 OR LESS
         NO FISSILE III

            FISSILE II,
            Tl OVER 1
         ALL FISSILE III
                                      Yellow
                        FIGURE?
              LABELING REQUIREMENTS

      B.     Shipping Paper Information (See Appendix IV for
             example of a typical shipping paper form)

             1.     Package marking requirements
                   (49CFR 172.300- 172.310)

                   a.     Proper shipping name

                   b.     Identification   number   (Section
                          172.101)

                   c.     The appropriate specification number
                          (Section 173.24(c)(I)(i))

                   d.     The gross weight of any packages in
                          excess of 50 kilograms

                   e.     "Type A" or "Type B" in letters 1/2
                          inch high as appropriate
5/93
                    21

-------
RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                  NOTES
             2.     Required   shipping   paper   information
                   (49 CFR 172.202(a)(I))

                   a.     Proper shipping name from 172.101

                   b.     Hazard   class   from   Section
                         172.202(a)(2))

                   c.     Identification number  from  Section
                         172.202(a)(3)

                   d.     Net quantity of material by weight or
                         volume from Section 172.202(a)(4)

                   e.     Radionuclide(s)  contained  in   the
                         package which comprise 1 % or more
                         of the total activity

                   f.     The physical form of the material or
                         a statement that  the  material  is
                         "special form"

                   g.     The activity in curies

                   h.     Category    of   "RADIOACTIVE"
                         labels applied to package

                   i.     "Transport   Index"   if   labeled
                         RADIOACTIVE-Yellow   II   or
                         RADIOACTIVE-Yellow III

                   j.     Identification markings shown on the
                         package

                   k.     Other information as required by the
                         mode of transportation or subsidiary
                         hazard of the material

                         Note:   See Appendix 4  for an
                         example of a typical shipping paper
5/93
22

-------
RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                                                       NOTES
       C.    Conditions for which the placarding of a transport
             vehicle  with  radioactive materials  is  required
             (49 CFR 172.504) (see Figure 8)

             1.    Highway Route-Control

             2.    LSA, exclusive use
                  i
             3.    Radioactive Yellow III
D.
                                   'RADIOACTIVE
           Vehicle Warning
              Placard
                        Special Placard for Highway
                         Route-Controlled Quantity
             4.
                  FIGURE 8
            VEHICLE PLACARDS

             Placards are to be affixed on each side and
             each end of a transport vehicle
             Shipper's Certification (49 CFR 172.402 (a))

             1.     Required for all shipments, other than air

                    "This is to  certify  that the above-named
                    materials are properly classified, described,
                    packaged, marked, and labeled, and are in
                    proper condition for transportation according
                    to   the   applicable  regulations  of  the
                    Department of Transportation."

             2.     For air transportation
5/93
                                   23

-------
RADIOACTIVE MATERIAL PACKAGING,
LABELING, AND SHIPPING
                                  NOTES
                   "I hereby certify  that the contents of this
                   consignment  are  fully  and  accurately
                   described above  by  the  proper  shipping
                   name and are classified, packed, marked,
                   labeled, and in proper condition for carriage
                   by air according to the applicable national
                   governmental regulations."
5/93
24

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      RADIOACTIVE MATERIAL PACKAGING, LABELING, AND SHIPPING
                                 APPENDIX I

                         f 17? 101 HAZARDOUS MATERIALS TAINT- Gonimtipd
to
ib-
o
5/93
Sym
bols
(i)



	

	
	
	
Hazardous matcuals descriptions and
proper sfi'pp*og names
(2)
Py/ophonc liquids rt o s
Pyiophonc melals, n,o s , of Py/o-
phone alloys, n o s
Pyrophonc so'ids. n o s 	
Pyiosulfuiyf cWO"do 	
PyiOxyhn solution or solvent, set Ni-
trocellulose

OwckKme. see Calcium oxide 	
ft }U. seo CKhlcxoielraliUoro* Inane .

R 124. see CMorotekaflt/o/oelhane .. ,.
Ft )33a, see CWwotrilluoroethane 	
ft $00, see Dtchtofodifluorome thane
and ditluorelrune. etc..
ft 502. see Chlorodtfluofomelhan*
and chkwopeolafluwoelhane mixture,
e/c. '
fi £03, «e Chloroui(luwomeih»ne
and Uif'uoroowtrwoa. »/c.
/? r7tf/. t+* ChlorodinuofOtxomo-
melrun«,
A ;J. see CMwolrilluofomethane 	 „
fl 1381, see Bromotrilluoromelhane 	


RadtoacUve material, excepled pack-
»ge.»rticlex nuwfacluced (rom n»lu-
ra) 
-------
     RADIOACTIVE MATERIAL PACKAGING, LABELING, AND SHIPPING

                            APPENDIX I (cont.)
                           Limited  Low Specific
        Proper Shipping Name   Quantity    Activity   Type A  Type B  HRCQ
1. Radioactive Material,
Limited Quantity,
N.O.S. (UN 2910)
2. Radioactive Material,
Instruments and
Articles (UN 2911)
3. Radioactive Material,
Low Specific Activity,
N.O.S. (UN- 291 2)
4. Radioactive Material,
Fissile, N.O.S.
(UN 2918)
5. Radioactive Material,
Special Form, N.O.S.
(UN 2974)
6. Radioactive Material,
N.O.S. (UN 2982)
*
*


*



*






*

»



*
»
*




*

5/93
26

-------
     RADIOACTIVE MATERIAL PACKAGING, LABELING, AND SHIPPING
       O

       Q
       •O
                                 APPENDIX II
              2£  S8
                                                 8 "8
                                            888 SH?1"  §8 §§l8
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                                                        1:
                                                            f-  i
                                                    I?'.l£
                   :FF'
          BI
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; i  UfssIS  §S  1 =
«~88
                                                    82
                                             ir*^ «c»n«rtn  r» «
                                             . 01 ^ ^i^OtOtM  
-------
    RADIOACTIVE MATERIAL PACKAGING, LABELING, AND SHIPPING
          ~ a a «> c S o e.c 8
          Sj*8|<=S5o
                         APPENDIX II (cont.)
 M i«, " ** O **    Tl

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                      Q >-,
                t|I * 23s§
                °«? =t:».,,
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                rf.E  Ki.&oTcisiitSojjSg-'"!
                S g  3cg2>J5^wc.og"».c.
                                                                 !>.£
     I  I

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 fegS-s^au,
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                         >c5 =
                         1.1-5 ss
                         " 3 o o Jtt ^
                         0^4 c a5 •

                             8|"R"  8288S28 88?§| 88
                                            JQV09 :  O O O O O O- fi r t «^ **! 1
                                             l"l   K R*"- Stf goodrfj
5/93
                                                                      o
                                                                      09


-------
            RADIOACTIVE MATERIAL PACKAGING, LABELING, AND SHIPPING

                                          APPENDIX III
<59 CFR 173.441(b)-  10 mr/hr at a point
2m from the vertical plane projected
by the outer lateral surface of the
vehicle; 200 mr/hr  at any point
•on the external surface of
the vehicle; 1000 mr/hr
at the external surfac_,v.   —     , ,.
of the package;    ^°         ^-^
2'mr/hr in cab.
                                    NOTE:
                                    LIMIT AUOAJf LIE!
                                    AT UNDERSIDE OF
                                    VEHICLE.
    EXCLUSIVE-USE CLOSED TRANSPORT
                                                        49 CFR 173.441(b)- 10 mr/hr at any point 2m
                                                        from the vertical planes  orojected from the
                                                        outer edges of vehicle;  2 mr/hr in cab;
                                                        200 mr/hr at the external surface of the
                                                        package.
                                                             EXCLU-SIVE-USE OPEN TRANSPORT
    CFR 173.441(a)- 200 mr/hr at any point
  on the external, surface of the package;
  10 mr/hr at 1m from any accessibj
  external surface of the
  package (i.e.,  T.I.i 10)
NOTE:
7OTAI.TJ IS HOT
10 FXCEEO.-SO P.£R
A» CFR 177.««(,].
                            NOTE;
                            STOWAGE OF rACKAGEX
                            WITIIW VEHICLE MUST
                            COMPLY WITH -(3 CFI1 177.W3(bl,
RADIATION
    LIMITS

   49 CFR  173.441
       (6/87)
   NON-EXCLUSIVE-USE OPEN OR CLOSED
                 TRANSPORT
      5/93
                                                29

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      RADIOACTIVE MATERIAL PACKAGING, LABELING, AND SHIPPING

                                       APPENDIX IV
  STRAIGHT BILL Of- LADING — SHORT FORM — ORIGINAL — NOT NEGOTIABLE.
  RECEIVED, jubjocl to lh. cloijllicollonj ond lorlfli in .li.cl on Ih. dol. ol lh« Inu. of Ihli Bill ol
                                                 Control Numb.
  FROM.
  Ai
  CONSIGNED TO
 (uxx. ot man wxxui Of CO«W&H«( _. KX n«rouj Of notv>c/.rxx omr.1

	Dol«	.	
  DESTIMAT1ON
                                                STATE
                                                                 COUNTY
HO.
PKGS.










IU
cU,
Ul
o.
HM
,









KIND OF PAOCAGE/DESCRIPTION bf ARTICLES, LINER NUM&ERS.
SfK3Al AAARKS 1 EXCEPTIONS











TTY=V^ fXU.

SCHEDULED DME T'WE
TO A.M.
ARRIVE f-W.
DATE TIME
AJ«IVED M
fc »t
^-,^



COMPLETED
LOADING
LEFT
SHIPPER
•W£ti?HT










^*<*4v»p*f1y d*-ccrlt>*-d obov* In poo of- correct to* r*<:6r*i U«~*pVSg t>rxJ bittrng
5/93
             30

-------
                          SECTION 14

     RADIOACTIVE SOIL AND WATER SAMPLING
          After completing this unit, participants will be able to:

          •    Explain the purpose of radioactive soil and water
               sampling.

          •    Given a site with radioactive soil and/or water
               contamination, list the minimum protective clothing
               requirements.

          •    List the three reference documents regarding
               unrestricted release.

          •    List the three primary radiation monitoring
               instruments used during soil sampling.

          •    Explain proper area setup for soil and water
               sampling.

          •    Explain proper protocol for radioactive soil and water
               sampling.

          •    Given a soil and/or water sample,  determine whether
               the sample is radioactively contaminated.

          •    Describe the procedure for handling a contaminated
               sample.
5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
                                                                       NOTES
I.
II.
Purpose

A.
       B.
To determine whether the soil or water is
radioactively contaminated

To characterize the radioactive components by
laboratory isotopic analysis
C.     To effectively control the radiological hazard


Sampling Preparation

A.     Reference documents for unrestricted release

       1.     10 CFR 20, Appendix B, Table II for
             water

       2.     Regulatory Guide 1.86, Table I for surface
             contamination

       3.     Policy and Guidance Directive FC 83-23
             for soil

B.     Protective clothing

       1.     Minimum protective clothing requirements
             shall include waterproof shoe covers
             (preferably boots) and waterproof gloves
             (preferably mid-forearm length)

       2.     Waterproof apron

C.     Monitoring and survey equipment

       1.     Portable alpha probe instrument

       2.     Portable beta-gamma probe instrument

       3.     Portable dose rate instrument

       4.     Smear paper (swipes,  wipes)
 5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
NOTES
             5.     Small plastic bags

             6.     Pen and data sheet for documentation

       D.    Sampling equipment and supplies

             1.     Sampling apparatus.  Optional depending
                    on type and depth of desired sample (i.e.,
                    auger, bailer)

             2.     Sample container(s) (i.e., 1 liter
                    polyethylene bottle with cap)

             3.     Metal tray

             4.     Medium mesh screen  (sieve - about 1 ft.
                    by 1  ft.)

             5.     Trowel or spoon

             6.     Plastic bags (medium  and large)

             7.     Plastic sheeting

             8.     Absorbent materials (i.e., cotton rags or
                    diapers)

             9.     Adhesive tape

             10.    Mild soap or detergent

             11.    Bucket of water

             12.    Barrier rope or tape, and stanchions

             13.    Appropriate work site posting (i.e., EPA
                    work area, do not enter without approval)

             14.    Scissors

             15.    Yellow and  magenta radiation rope
5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
                                                                        NOTES
III.
              16.    Radiological Control Area signs (e.g.,
                    Radiologically Contaminated Water,
                    Controlled Surface Contamination Area)
Simple Soil and Water Sampling Procedure

A.     Don protective clothing

       1.
       B.
                    Put on minimum shoe covers, gloves, and
                    apron
       Set up work area

       1.
             2.

             3.
Take area and surface readings to ensure
no surface contamination or exposure dose
rates are significant (Ref: U.S. EPA
SOSGs).

Establish perimeter work area.

Install barriers and posting (EPA work
area).
              4.     Spread plastic sheeting near sampling
                    location.

              5.     Set up small cleaning area on plastic
                    sheeting (about 2 ft. by 2 ft.).  Surround
                    with absorbent material.

              6.     Place bucket of mild soapy water, large
                    plastic bag, and some  absorbent cleaning
                    material near cleaning area, but not inside
                    cleaning area.

              7.     Set up staging area near sampling location
                    and place remainder of equipment.
5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
NOTES
       C.    Prepare to obtain sample

             1.      Ensure monitoring instruments are near
                    sampling location and operating

             2.      Have open sample bottle and cap handy

             3.      Have dry and dampened absorbent material
                    handy

       D.    Obtain sample

             1.      Draw  sample and monitor with portable
                    contamination instruments while sample is
                    being  obtained (Primarily beta-gamma,
                    spot check with alpha)

             2.      If background levels increase, it is a good
                    indication the sample is radioactive.  Use
                    dose rate instrument to periodically
                    monitor and  continue to draw sample

             3.      Soil sample - use dampened absorbent to
                    wipe container while still positioned over
                    sample area. Water sample - use dry
                    absorbent to wipe container while still
                    positioned  over sample area.  Dispose of
                    absorbents as waste.

             4.      Cap container

                    Note 1:  If soil sample is too chunky,
                    break up with trowel or use medium mesh
                    screen to sift sample onto metal tray.
                    Spread into thin layer, field monitor, and
                    transfer to sample container.

                    Note 2:  If exposure rate increases to 3-5
                    times  above  background, work can
                    continue, but a health physicist should be
                    consulted.
5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
NOTES
                    Note 3:  If exposure rate reaches 1 mR/hr
                    or above, stop work and consult with a
                    health physicist.  If not, continue as
                    normal.

             Survey sample(s) for surface contamination

             1.     Survey outer surfaces of sample container

             2.     If no contamination is detected by dry
                    smears, you can assume container is free
                    of loose surface contamination. Continue
                    as normal.

             3.     If any contamination at all is detected on
                    outer surfaces of container, dispose of
                    smears and used adsorbents in large plastic
                    bag as contaminated waste.  A direct
                    reading increase with beta-gamma probe
                    may indicate sample itself is contaminated.
                    Document results.

             Treatment of contaminated sample

             1.     If sample container is found to have loose
                    contamination on the outer surfaces, it
                    must be decontaminated. Transfer sample
                    to cleaning area,  change gloves, and
                    prepare to decontaminate.

             2.     Use bucket of mild soapy water and dip
                    some absorbent cleaning material in it.
                    Hand wring over bucket, then wipe outer
                    surfaces of container inside
                    decontamination area.  Dry and resurvey
                    for contamination.  Dispose of smears and
                    adsorbents as contaminated. Repeat as
                    necessary.

             3.     After outer surfaces are found to be
                    contamination free using dry smears,
                    decontamination is complete. Change
                    gloves. Place sample in medium plastic
5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
                                                           NOTES
             4.
bag and seal bag using adhesive tape.
Document results.

Note: 4 If outer surfaces remain
contaminated after decontamination, place
sample in medium plastic bag and seal.
Repeat using second medium plastic bag
and seal. Document results on data sheets
and containment labels.

Ensure outer surfaces  of plastic bag
containment is contamination free and
prepare sample for shipment to analytical
laboratory using prescribed radioactive
materials labeling and shipping methods.
IV.    Termination of Sampling Work

       A.    Monitor unused materials to verify they are
             contamination free, and remove from work area.
             Document results.

       B.    Monitor remaining materials, and decontaminate
             as necessary.  Those that can be verified as
             contamination free shall be removed from work
             area. Those that remain contaminated shall be
             disposed of in large plastic waste bag, or sealed
             in large plastic bag for subsequent use.
             Document results.

       C.    Blot any  wet areas on plastic sheeting using dry
             absorbent material. Carefully roll or fold up
             plastic sheeting and dispose of in large plastic
             waste bag.

       D.    Perform surface contamination survey of entire
             work area to ensure no residual contamination is
             left on the ground.  Document results.

       E.    If any freestanding water is noticeable in
             contaminated waste bag, place some additional
             dry absorbent material in bag to absorb water.
5/93

-------
RADIOACTIVE SOIL AND WATER SAMPLING
                                   NOTES
V.     Securing the Work Area/Setting up Radiological Area

       A.    Remove the work site perimeter barriers and
             postings.  Monitor to ensure they are
             contamination free.

       B.    Replace perimeter barriers with yellow and
             magenta radiation rope.

       C.    Classify area  based on survey results by posting
             appropriate radiological control area signs.
             Postings should be about waist high, and seen
             from all accessible approaches.

       D.    Secure contaminated waste inside posted area or
             properly transport for disposal as contaminated
             waste.
VI.    Final Monitoring

       A.    Monitor protective clothing; remove and properly
             dispose

       B.    Perform whole-body personnel monitoring prior
             to leaving work site
VII.   Considerations

       A.    Be conscientious of cross contamination at all
             times

       B.    Change gloves regularly, especially when
             performing different tasks.  Consider wearing two
             pair of gloves.

       C.    Work carefully to prevent the spread of
             contamination and be conscious of where
             potentially contaminated areas may be.

       D.    Monitor yourself whenever  you suspect you may
             have become contaminated,  and between steps
5/93
8

-------
RADIOACTIVE SOIL AND WATER SAMPLING
NOTES
      E.     Bring extra protective clothing

      F.     Dispose of all used smear paper as contaminated
             waste or place in small plastic bags to go to
             laboratory for further analysis

      G.     Ensure no freestanding water is noticeable in
             contaminated waste bag(s)

      H.     Remember double containment
5/93

-------
                       SECTION 15

 REGULATIONS AND GUIDANCE ON RADIOACTIVE
                    WASTE DISPOSAL
         After completing this unit, participants will be able to:

         •   Explain  the technical basis for  radioactive waste
             regulation development.

         •   Identify a source and category of radioactive waste.

         •   Identify the principle agencies governing radioactive
             waste disposal activities.
5/93

-------
REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
NOTES
I.      Exposure pathways associated with radioactive waste

       A.    inhalation of radionuclides

       B.    gamma radiation

       C.    ingestion of radionuclides
II.     Technical   basis   for   radioactive   waste   regulation
       development

       A.    ALARA

             As  low  as reasonably achievable,  taking  into
             account the state of technology the economics of
             improvements in relation to 1) the benefits to the
             public  health and  safety and  other societal and
             socioeconomic considerations  and 2) the use of
             atomic energy in the public interest

       B.    Biological Risk

             The probability of adverse health effects to the
             public due to radiation exposure

       C.    Comparable Risk

             The probability of risks of adverse health effects
             associated with different and/or alternative disposal
             methods
III.   Regulated  waste   management  activities   relating   to
      radioactive waste disposal

      A.    Accumulating

      B.    Processing

      C.    Handling

      D.    Packaging

5/93                                      2

-------
REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
NOTES
       E.     Transporting

       R     Storing

       G.     Disposing


IV.    Radioactive waste sources and types

       A.     Radioactive waste definition

             1.    Equipment or materials that are radioactive
                   or  have radioactive contamination and are
                   required  pursuant  to any  governing  laws,
                   regulations, or licenses to be disposed of as
                   radioactive waste.

             2.    Waste that  has  a  specific  activity  (the
                   activity of radionuclides  per  unit mass)
                   greater than 0.002 ^Ci per gram.

             3.    Waste that meets  the criteria in  10 CFR
                   61.55

       B.     Sources of radioactive waste

             1.    Fuel cycle facilities

             2.    Institutions

             3.    Commercial business

             4.    Government agencies

       C.     Categories of radioactive waste

             1.    Mill  tailings  -  rock  and  soil  that are
                   naturally radioactive and are by-products of
                   uranium mining and milling operations

             2.    Low  level  - waste  that contains  small
                   amounts of radioactive material dispersed in
5/93

-------
REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                                                     NOTES
V.
                   large  volumes and  poses little  potential
                   hazard

             3.     Transuranic - waste that contains more than
                   100 nanocuries of alpha-emitting  isotopes,
                   with half-lives greater  than 20 years per
                   gram of waste

             4.     High level - waste that is highly radioactive
                   and decays very rapidly

             5.     Mixed -  waste  that  contains hazardous
                   properties, as defined by 40 CFR  261, and
                   radioactive properties

             6.     HARM - naturally occurring and accelerator
                   radioactive waste
Federal Regulations and Guidance Documents

A.    Regulators/Responsible Parties

      1.     U.S.  Environmental  Protection  Agency
             (EPA)

             a.     Office of Radiation Programs (ORP)

      2.     Nuclear Regulatory Commission (NRC)

      3.     U.S. Department of Energy (DOE)

      4.     U.S. Department of Transportation (DOT)

      5.     Agreement States

      6.     International Atomic Energy Agency (IAEA)

B.    Radioactive waste regulation chronology

      1.     Atomic Energy Act of 1946
5/93

-------
REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
NOTES
                  a.     Established   the  Atomic   Energy
                        Commission (AEC)

                              Chartered to conduct research
                              and development on peaceful
                              application   of   nuclear
                              materials

                              Regulatory authority

            2.    Atomic Energy Act of 1954

                  a.     Emphasized  domestic  and
                        international   uses   of   atomic
                        applications

                  b.     Provided for the control of source,
                        by-product,  and  special   nuclear
                        materials

            3.    National Environmental Policy Act of 1969

                  a.     Established   the  Council   On
                        Environmental Quality

                  b.     Established the EPA in the executive
                        branch

                  c.     Requires the  assessment of every
                        federal  action that  may   have  a
                        significant   impact   on   the
                        environment

            4.    Energy Reorganization Act of 1974

                  a.     AEC divided into NRC and Energy
                        Research   and  Development
                        Administration (ERDA)

                  b.     ERDA   role   was   research,
                        development, and production
5/93

-------
REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
NOTES
                               National  defense  uses   of
                               radioactive materials

                   c.     NRC role was regulatory

                               Nondefense-related uses  of
                               radioactive materials

                   Energy Organization Act of 1977

                   a.     Replaces ERDA with DOE

                   b.     DOE National Security and Military
                         Applications of Nuclear
                         Energy  Authorization  Act,  Public
                         Law 96-164 (1979)

                   c.     Authorized  the DOE  to site and
                         construct the Waste Isolation Pilot
                         Project

                   Nuclear Waste  Policy  Act of 1982 - Public
                   Law 97-425

                   a.     Title 1,  Subtitle A - Repositories for
                         Disposal of High Level and Special
                         Nuclear Fuel

                               Mandates  the following:

                               EPA   must   promulgate
                               generally applicable standards
                               for  the protection of  the
                               general   environment  from
                               radioactive   material   in
                               repositories

                               NRC   must   promulgate
                               technical  requirements and
                               criteria  that  it will apply in
                               approving  or  disapproving
                               repository   construction
                               authorizations,   operating
5/93

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                                        NOTES
             7.
             licenses,  and  closure  and
             decommissioning
             authorizations

             DOE  must  issue   general
             guidelines for the
             recommendation of  sites for
             repositories

b.    Title 2, Research and Demonstration
      Regarding Disposal of High Level
      and Special Nuclear Fuel

             Authorizes DOE to provide
             for overall  management for
             construction, operation, and
             maintenance   of   a  deep
             geological test and evaluation
             facility

c.    Title 3, Other Provisions Relating  to
      Radioactive Waste

Low-Level   Radioactive   Waste   Policy
Amendments Act of 1985

a.    Endorses   the  below  regulatory
      concern (BRC) concept

b.    NRC role is  to establish procedures
      for acting on petitions to exempt
      specific radioactive wastestreams

c.    EPA   role is to  define  radiation
      exposure  allowable  in  such BRC
      deregulation

d.    States  are responsible for  low-level
      waste   generated   within  their
      boundaries,   with the exception  of
      DOE wastes  and U.S. Navy wastes
 5/93

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                   NOTES
                                 States may  form  compacts
                                 that shall not take effect until
                                 Congress has consented  by
                                 law to the compact

                                 Compacts  may   restrict  the
                                 use of the regional disposal
                                 facilities to  the  disposal of
                                 waste generated  within  its
                                 compact region after January
                                 1,  1986

                                 Extension  of deadline from
                                 1986 to 1992 for states to
                                 take  responsibility   for
                                 internally  generated  waste
                                 with the following  interim
                                 milestones:

                                       July  1986  -  ratify
                                       compact   of   state
                                       intent   to   site   a
                                       disposal facility

                                       January 1988  - select
                                       host state, or site  and
                                       developer, or certify
                                       that  storage will be
                                       provided

                                       January 1992 - license
                                       application

                                 Established disposal  facility
                                 volume  caps  and increasing
                                 surcharges  during  transition
                                 period

                                       1986 to  1987-$10.00
                                       per cubic foot

                                       1988 to  1989-$20.00
                                       per cubic foot
5/93
8

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                                        NOTES
             8.
                   1990 to 1992440.00
                   per cubic foot

Nuclear Waste Policy Amendments Act of
1987
                   a.     Placed a  moratorium on  all site-
                         specific activity at all candidate sites
                         except Yucca Mountain in the state
                         of Nevada

                   b.     Suspended any site-specific activities,
                         regarding a second repository

                   c.     Authorized the  construction of  a
                         Monitored  Retrievable   Storage
                         (MRS) facility

                   d.     Authorized payments to Nevada of
                         $10 million per  year during siting
                         and  construction  of  the  Yucca
                         Mountain repository, and $20 million
                         per year during operation, and  half
                         of  those amounts to the state of
                         Indian tribe where an MRS is located

      C.     Regulatory requirements applicable to radioactive
             waste disposal

                   1.     Title 40 - Protection of Environment,
                         Parts 190-192:   EPA's  Radiation
                         Protection Programs

                         a.     40  CFR 191 - Environmental
                                Standards for the
                                Management and Disposal of
                                Spent  Nuclear Fuel, High
                                Level   and   Transuranic
                                Radioactive Wastes

                         b.     40  CFR  192 - Health  and
                                Environmental   Protection
5/93

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                  NOTES
                               Standards  for Uranium  Mill
                               Tailings

                   2.     Title 10 - Energy, Parts 0-199: NRC

                         a.    10 CFR 20 -  Standards for
                               Protection Against Radiation

                                      Establishes  standards
                                      for protection against
                                      radiation   hazards
                                      arising   out   of
                                      activities (possession,
                                      use,  and  transfer  of
                                      radioactive materials)
                                      under licenses by the
                                      NRC, pursuant to the
                                      Atomic Energy Act of
                                      1954,  as  amended,
                                      and   the  Energy
                                      Reorganization Act of
                                      1974

                         b.    10 CFR 51  - Environmental
                               Protection  Regulations  for
                               Domestic    Licensing   and
                               Related Regulatory Functions

                                      Establishes  the NRC
                                      policy and procedure
                                      for   the  preparation
                                      and  processing   of
                                      environmental impact
                                      statements and related
                                      documents for actions
                                      significantly affecting
                                      the   quality  of  the
                                      human  environment
                                10 CFR  60
                                High-Level
                                Wastes   in
                                Repositories
      Disposal  of
     Radioactive
      Geological
 5/93
10

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                  NOTES
                                     Prescribes rules
                                     governing the
                                     licensing of DOE to
                                     receive  and  possess
                                     source,   special
                                     nuclear,   and   by-
                                     product  material at a
                                     geologic   repository
                                     operations area sited,
                                     constructed, or
                                     operated in
                                     accordance with the
                                     Nuclear Waste Policy
                                     Act of 1982.

                               10  CFR  61  -  Licensing
                               Requirements   for  Land
                               Disposal  of   Radioactive
                               Waste

                                     Establishes,  for land
                                     disposal of radioactive
                                     waste, the procedures,
                                     criteria,  terms,  and
                                     conditions upon which
                                     the   NRC    issues
                                     licenses   for   the
                                     disposal of radioactive
                                     wastes containing by-
                                     product,  source,  and
                                     special   nuclear
                                     material

                               10 CFR 71 - Packaging and
                               Transportation of Radioactive
                               Material

                                     Sets requirements for
                                     packaging,
                                     preparation for
                                     shipment, and
                                     transportation of
                                     licensed   radioactive
5/93
11

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                  NOTES
5/93
                                     materials,   and
                                     procedures   and
                                     standards  for  NRC
                                     approval of packaging
                                     and   shipping
                                     procedures for fissile
                                     material  and  for  a
                                     quantity   of  other
                                     licensed  material  in
                                     excess of a  Type A
                                     quantity

                         f.     10  CFR  73  -   Physical
                               Protection  of  Plants  and
                               Materials

                                     Prescribes
                                     requirements for the
                                     establishment  and
                                     maintenance    of   a
                                     physical   protection
                                     system   for   the
                                     protection of special
                                     nuclear  material  at
                                     fixed   sites  and  in
                                     transit, and of plants
                                     in   which   special
                                     nuclear  material  is
                                     used

                   3.     Title 49 - Transportation, Parts 171-
                         179

                         a.     49  CFR  173  -  Shippers-
                               General  Requirements  for
                               Shipments and Packaging

                                     Sets forth
                                     requirements  for the
                                     transportation   of
                                     radioactive materials
                                     by    carriers  and
                                     shippers
12

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REGULATIONS AND GUIDANCE ON RADIOACTIVE
WASTE DISPOSAL
                                 NOTES
                  4.     IAEA Transport Regulations

                        a.     Sets forth  minimum  safety
                              requirements   based   on
                              performance standards that
                              would   be  universally
                              applicable and could serve as
                              a  basis  for  national and
                              international  regulations
                              applicable to the transporta-
                              tion radioactive materials
5/93
13

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                         SECTION 16

          REMEDIAL AND DISPOSAL OPTIONS
          After completing this unit, participants will be able to:

          •    Identify two characteristics of CERCLA remediation
               activities.

          •    List the two kinds  of disposal options and  give an
               example of an associated disposal technique for each.

          •    Discuss one remedial treatment technology.
5/93

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REMEDIAL AND DISPOSAL OPTIONS
I.   Technological Approaches to the Cleanup of Radiologically
    Contaminated Superfiind Sites,  EPA/540/2-88/002,  August
    1988

    A.  Comprehensive Environmental Response, Compensation
        and Liability Act mandates:

        1.  Protective actions

        2.  Permanent solutions

        3,  Use of alternative treatment technologies or resource
            recovery options to the maximum extent practicable

        4.  Cost efficiency

        5.  Consideration of  any  applicable or  relevant  and
            appropriate requirements (ARARs)

    B.  Radiation Risk Assessment Procedures

        1.  Developed by the ICRP in 1979, adopted by EPA in
            Federal Guidance Report 11,  1988

        2.  Guidance  on potential  ARARs is available in  the
            CERCLA Compliance with Other Laws manual (US
            EPA 1988c)

    C.  Technological    Approaches   to   the    Cleanup   of
        Radiologically Contaminated Superfund Sites

        1.  Focus  -  treatment  and disposal  of radiation
            contaminated soils and radon  control

        2.  Principle contaminants

            a.  by-products of uranium, thorium, and radium
                processing

            b.  contaminated buildings and equipment

            c.  stream sediments

    D.  DOE Remediation Sites
NOTES
5/93

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 REMEDIAL AND DISPOSAL OPTIONS


         1.  Formerly  Utilized Sites  Remedial Action  Project
             (FUSRAP) -
             29 sites (3 on NPL)

         2.  Uranium Mill  Tailings  Remedial Action  Project
             (UMTRAP) - 24 sites

         3.  Grand Junction  Remedial  Action Project (GJRAP) -
             1 site

         4.  Surplus Facilities Management Program (SFMP) -17
             sites



                 thousands of DOE vicinity properties

                 single largest problem is the disposal of waste

     E.  Remedial and Disposal Options

         1.   Categories of Disposal

             a.   Onsite  - waste  is  disposed of at the  site  of
                 generation

             b.   Off site - waste is disposed of at a site other than
                 the site of generation

         2.   Onsite Disposal  Methods

             a.   Capping - covering the radioactive material with
                 a barrier sufficiently thick and impermeable  to
                 minimize diffusion  of radon gas and attenuate
                penetrating  radiation, as well as prevent  or
                minimize water  infiltration without enhancing
                erosion.

                i.   Advantages

                        ease of application

                        high reliability if properly maintained
NOTES
5/93

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REMEDIAL AND DISPOSAL OpriONS
NOTES
                         soil characteristics not  as  critical as
                         other treatment technologies

                ii.   Disadvantages

                         no reduction in radioactivity

                         long-term maintenance

                         site security

            b.  Vertical Barriers - installation of a wall of low-
                permeability material around the  outside of
                contaminated area to limit lateral migration

                i.   Grout materials:  Portland cement,  alkali
                     silicate, and organic polymers

                ii.   Advantages

                         simple to install

                         act as a treatment vessel

                iii.  Disadvantages

                         difficult to  obtain low permeability

                         chemical waste incompatibility

                         requires detailed physical and chemical
                         data on soil characteristics

        3.  Off site Disposal Methods

            a.  Land Encapsulation - the physical isolation of
                waste in a structure  that is  entombed in  the
                ground

                i.   Advantages

                         mature technology
5/93

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REMEDIAL AND DISPOSAL OPTIONS
NOTES
                         complete  removal  of contamination
                         from site

                         simple prerequisite informational needs

                ii.   Disadvantages

                         site selection

                         acceptance of waste by existing facility

                         handling and transportation

                         concentration of wastes may result in
                         unacceptably high radioactivity

            b.  Land Spreading - excavating the contaminated
                material,  transporting it to a suitable site, and
                spreading it on unused land, assuming that the
                radioactivity levels will be the  same as natural
                background

                i.   Advantage

                         suitable for dry, granular tailings and
                         soils

                ii.   Disadvantage

                         not appropriate for mixed waste

            c.  Underground Mine Disposal - the placement of
                waste in subterranean mines

                i.   Advantages

                         suitable for highly concentrated waste

                         suitable for untreated wastes

                ii.   Disadvantages

                         high cost
5/93

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REMEDIAL AND DISPOSAL OPTIONS
NOTES
                        additional research is needed on costs
                        and   risks   associated   with
                        transportation of material

                        groundwater  contamination must be
                        considered and prevented

            d.  Ocean Disposal - the disposal of low-level waste
                in the sea

                i.   Advantage

                        extreme isolation of low-level waste

                ii.  Disadvantage

                        high   degree   of   risk   due   to
                        transportation safety issues

        4.  Remedial  Treatment Technologies

            a.  General

                i.   Radioactive contaminants are not altered or
                    destroyed by treatment technologies.  The
                    volume of contaminated material may be
                    reduced, with a corresponding elevation in
                    concentration, but containment and/or burial
                    is the only remedy.

                ii.  Chemical extraction and physical separation
                    technologies  have  been  used   in  a site
                    remediation situation. The same holds true
                    for stabilization and solidification.   Only
                    excavation and land encapsulation have been
                    used    at   radiologically  contaminated
                    Superfund  sites.

            b.  Onsite Treatment

                i.   Stabilization  or   solidification
                    immobilization or restraint of radionuclides
                    by trapping them in an impervious  matrix
5/93

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REMEDIAL AND DISPOSAL OPTIONS
NOTES
                    (i.e.,  Portland  cement,  silica  grout,  or
                    chemical grout)

                    (a)  Advantages

                            can be accomplished in situ or by
                            excavation,    mixing,   and
                            replacement

                            can be combined with capping

                    (b)  Disadvantage

                            interference   by   hazardous
                            chemicals

                ii.  Vitrification- immobilization of radioactive
                    waste by melting waste material between
                    two  or more electrodes, resulting in  a
                    glassy mass after cooling

                    (a)  Advantage

                            effective immobilization

                    (b)  Disadvantages

                            energy intensive

                            volatilization of waste substances

                iii.  Chemical   extraction   -  concentrate
                    radioactive  contaminants  by   chemical
                    extraction, thereby  reducing the volume of
                    waste for final disposal

                    (a)  Water - contaminated soils  or tailings
                        are mixed  with  large  quantities  of
                        water.     The  water,  with  soluble
                        radionuclides  fraction is removed  by
                        physical separation
5/93

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REMEDIAL AND DISPOSAL OPTIONS
                                   NOTES
                        (i)  Advantages
5/93
                                 inexpensive,   simple
                                 technology

                        (ii) Disadvantages

                                 consumes large  quantities of
                                 water

                                 low efficiency of extraction

                    (b) Inorganic Salts  - theory  of operation
                        similar to water extraction

                        (i)  Considerations

                                 nature   of   the   tailings;
                                 geochemistry  particle  size,
                                 and chemical   composition;
                                 concentration of salt solution

                            -    pH

                                 solid-to-liquid ratio

                                 process time

                                 temperature

                                 method of extraction

                    (c) Mineral Acids - soils are ground and
                        mixed with water to form a slurry.
                        Contaminants are leached from the
                        slurry   using   inorganic   acids.
                        Radionuclides  are  removed   from
                        leachate  by  ion  exchange,   solvent
                        extraction, or precipitation

                        (i)  Application

                            extraction of radium, thorium, and
                            uranium as well as other metals
8

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REMEDIAL AND DISPOSAL OPTIONS
                              NOTES
                         (ii)  Advantages

                                 high extraction efficiency

                                 other metals are removed

                                 relatively small liquid-to-solid
                                 ratio compared to extraction
                                 with water or inorganic salts

                                 costs  can be reduced  by
                                 recycling acids

                         (iii)     Disadvantages

                                 increased operating costs due
                                 to expensive reagents, higher
                                 operating  temperatures,  and
                                 stainless steel reaction vessels
                                 necessary   for   strong
                                 corrosives

                                 multistage process

                                 chemically leached material
                                 containing   nitrates   or
                                 chlorides  that  may be more
                                 harmful  than  the  original
                                 contaminated soils
                    (d) Complexing   Agents
                        agents in the slurry

                        (i)  Application
complexing
                                 removal  of  radium
                                 contaminated  soils

                        (ii) Liquid Treatment

                                 precipitation and
                                 coprecipitation
      from
5/93

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REMEDIAL AND DISPOSAL OPTIONS
                                    NOTES
                                 solvent extraction

                                 ion exchange

                        (iii)      Advantages

                                 high extraction efficiency for
                                 radium

                                 low   volume  of   reagent
                                 needed

                        (iv) Disadvantages

                                 reagents are very expensive

                                 only effective for radium


        5.  Superfund Innovative Technology Evaluation (SITE)
            Program

            a.  General

                i.   Administered  by EPA's Office of Research
                    and Development

                ii.  Encourages the development of innovative
                    treatment technologies

                iii.  Evaluates new treatment technologies

                iv.  Disseminates   technical   information  on
                    treatment technologies

            b.  SITE Program Description

                i.   Emerging technology

                ii.  Demonstration program
5/93
10

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REMEDIAL AND DISPOSAL OPTIONS
                                   NOTES
                iii.  Monitoring and measurement technologies
                    program

                iv.  Technical transfer program

            c.   SITE Information Resources

                i.   Alternative   Treatment   Technology
                    Information   Center   (ATTIC),  System
                    Operator: 1-301-670-6294

                ii.   Vendor Information System for  Innovative
                    Treatment Technologies (VISITT)

                    Hotline: 1-800-245-4505

                iii.  SITE Clearinghouse Hotline:

                    1-800-424-9346 or 1-202-382-3000

                iv.  Center   for   Environmental   Research
                    Information (CERI):  1-513-569-7562

                v.   ORD Publications, 26 Martin Luther King
                    Dr. (G72), Cincinnati, OH  45268
5/93
11

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                                   GLOSSARY
ABSORPTION:
ACCELERATOR. PARTICLE;
ACTIVATION;
ACTIVITY;
ACUTE EXPOSURE;
AGREEMENT STATE;
The process by which radiation imparts some or
all of its energy to any material through which it
passes.

A device for imparting large quantities of kinetic
energy to electrically charged particles such as
electrons, protons, and helium ions.

The process of inducing radioactivity by
irradiation.

The number of nuclear transformations occurring
in a given quantity of material per unit time. The
unit of measure is the curie (Ci).

Exposure occurring over a short, up to a few
days, period  of time.

Any state in  the United States with which the
NRC has made an effective agreement under
Subsection 274(b) of the Atomic Energy Act of
1954, as amended, relative to the licensing and
control of radioactive material used or produced
within that state.

The term applied to radioactive contamination
loose in air,  filtered from the air, or deposited
from the air, as contrasted with contamination
spread by splashing, dripping, etc...
AIR-WALL IONIZATION CHAMBER; An ionization chamber in which the materials of
                                      the wall and electrodes are so selected as to
                                      produce ionization essentially equivalent to that in
                                      a free-air ionization chamber. This is possible
                                      only over limited photon energy ranges. Such a
                                      chamber is normally called an "air-equivalent
                                      ionization chamber".
AIRBORNE CONTAMINATION;
ALARA;
An acronym for "As Low As Reasonably
Achievable"; refers to the operating philosophy in
which occupational exposures are maintained as
far below the specified limits as is reasonable to
achieve.

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ALPHA PARTICLE;
AMPLIFTCATIQN;
ANALYZER. PULSE HEIGHT;
ANGULAR DEPENDENCE;
ANODE;
ARTIFICIAL RADIOACTIVITY;
ATOMIC NUMBER
ATTENUATION;
AUTHORIZED MATERIAL;
AVALANCHE:
A charged particle that is emitted from the
nucleus of an atom, and that has a mass and
charge equal in magnitude to those of the helium
nucleus, i.e., two protons and two neutrons.

As  related to radiation detection instruments, the
process (gas, electronic or both) by which
ionization effects are magnified to a degree
suitable for their measurement.

An electronic circuit that sorts and records pulses
according to their height.

The varying ability of an instrument to detect
radiation, depending on its orientation with
respect to the radiation field.

A positive electrode; the electrode to which
negative ions are attracted.

Manmade radioactivity produced by bombardment
or electromagnetic irradiation.

The number of protons in the nucleus of an atom.

The process by which a beam of radiation is
reduced in intensity or energy when passing
through some material.

Radioactive material not requiring a specific NRC
license. The receipt, possession, use, or transfer
of radioactive material requires authorization by a
specific agency or organization.

The multiplicative process by which a single
charged particle accelerated by a strong electric
field produces additional charged particles through
collision with neutral gas molecules. This
cumulative increase in ions  is also known as
"Townsend Avalanche" or "Townsend
Ionization".

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BACKGROUND RADIATION;
BEAM;


BECOUEREL;


BETA PARTICLE;



BIOASSAY;


BREMSSTRAHLUN6;



BYPRODUCT MATERIAL
CALIBRATION;



CATHODE;


CELL. BIOLOGICAL;


CHARACTERISTIC RADIATION;
Radiation arising from radioactive material other
than the one directly under consideration.
Background radiation due to cosmic rays and
natural radiation is always present. There may
also be background radiation due to the presence
of radioactive substances in other parts of a
building.

A unidirectional or approximately unidirectional
flow of electromagnetic radiation or of particles.

The SI unit of activity equal to a nuclear
disintegration rate of 1 disintegration per second.

A charged particle emitted from the nucleus of an
atom,  with a mass and charge equal in magnitude
to that of an  electron.

An evaluation of the amount of radioactivity taken
into the body.

Secondary photon radiation produced by the
deceleration of charged particles passing through
matter.

Any material (except special nuclear material)
made  radioactive by either exposure to radiation,
or the process of producing or using special
nuclear material.

The determination of a measuring instrument's
variation from the standard, to ascertain necessary
correction factors.

A negative electrode. The electrode to which the
positive charged ions are attracted.

The fundamental unit of structure and function in
organisms.

Radiation originating from an atom resulting from
removal of an electron or excitation of the
nucleus. The wavelength of the emitted radiation
is specific, depending only on the nuclide and the
particular energy levels involved.

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CHRONIC EXPOSURE:
COLLECTIVE DOSE EQUIVALENT;
COLLISION;
COMPTON EFFECT;
CONDENSER R-METER;
CONTAMINATION. RADIOACTIVE;
CONDENSER R-METER:
CONTAMINATION. RADIOACTIVE:
CRITICAL ORGAN;
Radiation exposure occurring over a long but not
necessarily continuous period of time.

The sum of dose equivalents received by a given
population or group of workers, expressed in
units of person-rem.

An encounter between two subatomic particles
that changes the initial momentum and energy
conditions. The products of the collision need not
be the same as the initial system.

An attenuation process observed for X or gamma
radiation in which the incident photon interacts
with an orbital electron of an atom to produce a
recoil electron and the scattered photon with an
energy  less than that of the incident photon.

An instrument consisting of an air-wall ionization
chamber together with the auxiliary equipment for
charging and measuring voltage. It is used as an
integrating instrument for measuring the exposure
of X or gamma radiation in Roentgens (R).

The deposition of radioactive material in any
place where it is not desired, and particularly in
any place where its presence may be harmful.

An instrument consisting of an air-wall ionization
chamber together with auxiliary equipment for
charging and measuring its voltage. It is used as
an integrating instrument for measuring the
exposure of x or gamma radiation in roentgens
(R).

The deposition of radioactive material in any
place where it is not desired, and particularly in
any place where its presence might be harmful.

The organ of the  body receiving a specified
radioisotope that results in the greatest
physiological damage to the body. For exposure
to ionizing radiation from external sources, the
critical organs are the skin, blood-forming organs,
gonads, and eyes.
                                                                                     4

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I
           CROSS-CONTAMINATION;
           CUMULATIVE DOSE;
           CURIE;
           DAUGHTER;
           DECAY CONSTANT;
           DECAY. RADIOACTIVE;
           DECONTAMINATION;
           DETECTOR. INTEGRATING;
           DETECTOR. RADIATION;
           DETECTOR. SCINTILLATION;
Contamination not from an original source, but
acquired from another contaminated object.

The total dose resulting from repeated exposure
over some period of time.

The unit of activity (abbreviated Ci). One curie
equals exactly 3.7 X  10 (10) nuclear
disintegrations per second.

Synonym for radioactive decay product.

The fraction of the number of atoms of a
radioactive nuclide that decay per unit time.

The disintegration of the nucleus of an unstable
nuclide by the spontaneous emission of charged
particles or electromagnetic  waves.

The reduction or the removal of radioactive
contamination from any given surface.

A detector that measures the total accumulated
radiation quantity (such as exposure or dose)
rather than the rate of the accumulation of the
radiation. Devices that accumulate and hold
charges (e.g. electrometers) and that indicate
measures proportional to the total dose are this
type. Examples of integrating detectors are
electrometers, film badges, pocket dosimeters,
and neutron activation detectors.

Any device for converting radiant energy to a
form more suitable for observation. An
instrument used to determine the presence, and
sometimes the amount, of radiation.

A radiation detector whose response is a light
signal generated by the incident radiation and a
scintillating medium. The light signal is
transformed into an electronic signal through an
adjacent, optically coupled,  photo-sensitive
device.

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DETECTOR. SOLID STATE;
DETECTOR. TRACK ETCH;
DISINTEGRATION. NUCLEAR;
DOSE;
DOSE. ABSORBED;
DOSE. WHOLE-BODY;
DOSE. EQUIVALENT;
DOSIMETER;
A detector that uses a semiconductor such as
selenium or germanium that responds to radiation
with an electronically measurable pulse.

A device that records the paths of heavy charged
particles in a transparent solid. The tracks may be
directly visible, or may be enhanced by etching
with an appropriate reagent (such as potassium
hydroxide  for etching cellulose acetate).

A spontaneous  nuclear transformation
(radioactivity) characterized by the emission of
energy and/or mass from the nucleus.

A term denoting the quantity of radiation energy
absorbed. The  term must be qualified. If
unqualified, it  refers to absorbed dose.

The amount of energy deposited in matter per unit
mass of material by ionizing radiation. The
common unit of absorbed dose is the rad, which
is equal to 100 ergs of absorbed energy per gram
of material ( or 0.01 J/Kg). The SI unit of
absorbed dose  is the gray, which is equal to 1
J/Kg of material.

The average uniform absorbed dose or dose
equivalent received by a person whose whole
body is exposed to ionizing radiation from a
source.

The product of the absorbed dose, quality factor,
and other modifying factors necessary to evaluate
the effects of irradiation received by exposed
persons, this unit of measure takes into account
the particular characteristics of the exposure.  The
common unit of dose equivalent is the rem. The
SI unit is the sievert. Absorbed doses of different
types of radiation are not additive, but dose
equivalents are, because they express on  a
common scale  the amount of damage incurred.

An instrument  to detect and measure accumulated
radiation exposure. In common usage, a pencil
sized ionization chamber with a self-reading
electrometer, used for personnel monitoring.

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DOSIMETER. FILM:
DOSIMETER. PERSONNEL;
DOSIMETER. POCKET;
DOSIMETER.
THERMOLUMEVESCENT;
EFFICIENCY. COUNTING;
ELASTIC COLLISION;
ELECTRON;
ELECTRON VOLT;
EMULSION. NUCLEAR;
An integrating detector that uses photographic
film and density measuring instruments to
determine the absorbed dose.

A dosimeter of small size carried by a person to
determine the exposure, absorbed dose, and/or
dose equivalent received during the wearing
period.

An ion chamber type dosimeter the shape and size
of a pencil with a clip, to be worn in the pocket
like a fountain pen.

An integrating detector that utilizes a phosphor
sensitive to ionizing radiation. The phosphor
stores the energy of the ionizing radiation within
itself and releases it as low-energy photons (light)
when  heated. The total amount of light emitted is
proportional to the total absorbed dose.

A measure of the probability that a count will be
recorded when radiation is incident on the
detector. Uses of this term vary considerably, so
it is well to ascertain which factors ( window
transmission, sensitive volume, energy
dependence, etc.) are included in a given case.

A collision in which there is no change either in
the internal energy of each participating system or
in  the sum of their kinetic energies.

A stable elementary particle that has an electric
charge equal to (+/-)  1.60210 X 10-19 coulomb
and a rest mass equal to 9.81091 X 10-30 Kg.

A unit of energy equivalent to the energy gained
by an electron in passing through a potential
difference of 1 volt. Larger multiple units of the
electron volt are frequently used: KeV, MeV. 1
eV = 1.6X 10-12 ergs.

A photographic emulsion specially designed to
permit observation of the individual tracks of
ionizing radiation.

-------
ENRICHED MATERIAL;
EXCITED STATE;
EXPOSURE;
EXPOSURE RATE:
EXTERNAL RADIATION;
FINGER DOSIMETER:
FISSILE;
The characteristic response of a radiation detector
to a given range of radiation energies or
wavelengths, compared with the response of a
standard free-air  chamber.

(1) Material in which the relative amount of one
or more isotopes of a constituent has been
increased.

(2) Uranium in which the abundance of the U-235
isotope is increased above natural levels.

An unstable condition of the nucleus of a atom
after the entrance of a nuclear particle or gamma
ray photon.

(1) The incidence of radiation upon inanimate or
living matter by intent or accident.

(2) For X or gamma radiation, the sum of the
electrical charges of all ions of one sign produced
in air when all electrons liberated by photons in a
suitable small volume of air are completely
stopped in air, divided by the mass of air in the
volume.

The unit of exposure is the roentgen (R).

(1) The exposure divided by the time over which
it was accumulated.

(2) The increment of exposure during a suitably
small interval of time, divided by that interval of
time.

The typical unit for exposure rate is roentgens per
hour (R/hr).

Radiation from a source outside of the body.

A dosimeter in the form  of a ring to be worn by
personnel to determine radiation doses to the
hands.

A nuclide capable of undergoing fission by
interaction with thermal neutrons.

-------
FISSILE MATERIAL:
FISSION. NUCLEAR;
FISSIONABLE;
FISSION PRODUCTS;
FLUORESCENCE;
GAS AMPLIFICATION;
GEIGER-MUELLER COUNTER;
GEOMETRY. GOOD;
GEOMETRY. POOR;
GEOMETRY. RADIATION;
Plutonium-238, 239, 241, uranium-233, 235, or
any material containing any of the foregoing [49
CFR 173.389(a) and 173.398(a)].

A nuclear transformation characterized by
splitting of a nucleus into at least two other nuclei
and the release of a relatively large amount of
energy.

Pertaining to a nuclide that is capable of
undergoing  fission by any process.

Elements or compounds resulting from fission.

The emission of radiation of particular
wavelengths by a substance as a result of the
absorption of radiation of shorter wavelengths.
This emission occurs essentially only during
irradiation.

As applied to gas-ionization instruments for
detecting radiation, the ration of the charge
collected to the charge produced by the initial
ionizing event.

A highly sensitive, gas-filled radiation measuring
device. It operates at voltages high enough to
produce avalanche ionization.

In nuclear physics measurements, an arrangement
of source and detecting instrument that introduces
little error when a finite source size and finite
detector aperture are used.

In a nuclear experiment, an arrangement in which
the angular  aperture between the source and the
detector is large, introducing a comparatively
large uncertainty for which a correction would be
necessary.

A nuclear physics term referring to the physical
relationship and symmetry of the parts of a
radiation detection instrument. Counting
efficiency is closely related to geometry.

-------
GLOW CURVE;
GLOW PEAK;
GRAY:
GROUND STATE;
HALF-LIFE. BIOLOGICAL;
HALF-LIFE. EFFECTIVE;
HALF-LIFE. RADIOACTIVE;
HALF-VALUE LAYER (HVL);
HEALTH PHYSICS;
 INDUCED RADIOACTIVITY;
 INELASTIC COLLISION;
In thermoluminescent dosimetry, a graph of the
released luminescence photon fluence as a
function of temperature or time of heating.

In thermoluminescent dosimetry, the time or
temperature during the heating of a
thermoluminescent phosphor at which  the release
rate of the luminescent photons is at its
maximum.

The SI unit for absorbed dose, equal to the
absorbed energy from ionizing radiation of 1
joule/kg, and equal to 100 rads.

The state of a nucleus, atom, or molecule at its
lowest energy. All other states are "excited"

The time required for the body to eliminate one
half of an  administered dosage of any substance
by process of elimination.

The time required for a radioactive element in the
body to be diminished by one half as a result of
the combined biological elimination and
radioactive decay.

The time required for a radioactive substance to
lose one half of its activity due to radioactive
decay. Each radionuclide has a specific half-life.

The thickness of a specified substance that, when
introduced into the path of a given beam of
radiation,  reduces the exposure rate by one half.

A science  and profession devoted to protecting
man and environment against unnecessary
radiation exposure.

Radioactivity produced in a substance after
bombardment with neutrons or other particles.

A collision in which there are changes both in the
internal energy of one or more of the colliding
systems and in the sums of the kinetic energies of
translation before and after the collision.

-------
INTENSITY:
INTENSITY. SOURCE;
INTERNAL RADIATION:
IN-VIVO COUNTING;
ION;


IONIZATION;


IONIZATION CHAMBER;
ION PAIR;
ISOMERS:
ISOTOPES;
The amount of energy per unit time passing
through a unit area perpendicular to the line of
propagation at the point in question.

A general term for the magnitude of the source
emission rate. Source intensity is usually
expressed in units of curies or bequerels.

Radiation from a source within the body as a
result of deposition of radionuclides in body
tissue.

Measurements of internal radiation made at the
surface of the body and based on the fact that
radioisotopes emit radiation that can traverse the
tissues and be measured outside the subject. In-
vivo counting is synonymous with whole-body
counting.

An atomic particle or atom bearing an electric
charge, either positive or negative.

The process by which a neutral atom or molecule
acquires a positive or negative charge.

An instrument designed to measure a quantity of
ionizing radiation in terms of the charge of
electricity associated with ions produced within a
defined volume.

Two particles of opposite charge, usually
referring to the electron and the positive atomic
or molecular residue resulting from interaction
with ionizing radiation.

Nuclides with the same number of neutrons and
protons but capable of existing, for a  measurable
period of time, in different quantum states with
different energies and radioactive properties.

Nuclides that have the same number of protons,
hence the same atomic number, but differ in
neutron number and therefore differ in atomic
mass.
JOULE;
The unit for work and energy, equal to 10-7 ergs.

-------
LATENT PERIOD:
LICENSE. SPECIFIC;
LICENSE-EXEMPT MATERIAL;
LICENSED MATERIAL;
LINEAR ENERGY TRANSFER;
MAN-REM;
MAXIMUM CREDIBLE ACCIDENT;
MONITORING:
MONTE CARLO METHOD:
NEUTRINO;
The interval of seeming inactivity between the
time of irradiation and  the appearance of an
effect.

A document issued by NRC under 10 CFR that
gives the bearer the right to procure, receive,
store, transfer, use, export, and import specified
radioactive items under specific terms.

Radioactive material not subject to NRC or
exempt from NRC licensing under 10 CFR.

Source,  special nuclear, or byproduct material
received, stored, possessed, used, or transferred
under a  general or specific license issued by the
NRC or an Agreement State.

The linear rate of loss of energy per unit distance
transited by an ionizing particle moving through a
medium.

A unit of collective dose  equivalent. Man-rem is
the product of the population times their average
dose equivalent.

The worst accident in a reactor or nuclear energy
installation that, by agreement, need be taken into
account  in devising protective measures.

Periodic or continuous determination of the
amount  of ionizing radiation or radioactive
contamination present in  an occupied region.

A method of permitting the computer solution of
physics  problems, such as those of neutron
transport, by determining the history of a large
number  of elementary events by the application of
the mathematical theory of random variables.

A neutral particle of very small rest mass
originally postulated to account for the continuous
distribution of energy among particles in the beta-
decay process.

-------
NEUTRON;
NUCLEON;
NUCLEUS;
NUCLIDE;
PAIR PRODUCTION;
PARENT;
PERSONNEL MONITOR;
PHANTOM;
PHOSPHORESCENCE;
An elementary particle with neutral charge and a
mass of 1.008 mass units, and is found in the
nucleus of the atom.

The common name for a constituent particle of
the nucleus, usually applied to protons and
neutrons.

That part of the atom in which the total positive
charge and most of the mass are concentrated.

A species of atom characterized by the
constitution of its nucleus. To be a nuclide, the
atom must exist for a measurable period of time.
Thus nuclear isomers are considered separate
nuclides.

An absorption process for x or gamma radiation
in which the photon is annihilated in the vicinity
of a nucleus of an atom with the production of an
electron and positron. Minimum energy for this
reaction to occur is 1.022 MeV.

A radionuclide which upon disintegration, yields a
specified nuclide, either directly or as a later
member of a radioactive decay series.

An instrument that measures a radiation quantity
proportional to dose equivalent, for use by an
individual working in a radiation area.

A volume of material approximating as closely as
possible the density and effective atomic number
of body tissue. Ideally, a phantom should  absorb
radiation in the same way tissue does. Radiation
dose measurements made with a phantom  provide
a means of determining  the radiation dose within
the body or tissue under similar exposure
conditions.

The emission of radiation by a substance as a
result of the previous absorption of radiation of
shorter wavelength. In contrast to fluorescent
emissions, phosphorescent emissions may
continue for a considerable time after cessation of
the existing irradiation.

-------
PHOTOELECTRIC EFFECT;
PHOTON;
PIG;


PRIMARY IONIZATION;
PROPORTIONAL COUNTER;
PROTECTIVE CLOTHING;
PROTECTIVE EQUIPMENT;
PROTON;
 QUALITY FACTOR (O);
 QUENCHING;
The process by which a photon ejects an electron
from an atom. All energy of the photon is
absorbed in ejecting the electron and imparting
kinetic energy to it.

A quantity of electromagnetic energy (E) whose
value in joules is the product of its frequency (v)
in hertz and Planck's constant  (h). The equation is
E=hv.

A container, usually made of lead, used to ship or
store radioactive materials.

(1) In collision theory: the ionization produced by
primary particles, as contrasted with total
ionization, which includes secondary ionization.

(2) In counter theory: the total ionization
produced by incident radiation without gas
amplification.

A gas filled detector that operates in that range of
applied voltage  in which gas amplification  occurs,
such that the total charge collected is proportional
to the charge liberated by the original ionizing
event.

The clothing worn by radiation workers to
prevent radioactive contamination of the body or
personal clothing.

Safety devices such as goggles or clothing  used to
do a job safely.

An elementary nuclear particle with a positive
electric charge equal numerically to the charge of
the electron and a mass of 1.007277 mass  units.

The factor dependent on linear energy transfer by
which  absorbed doses are multiplied to obtain a
quantity that expresses the effect of the absorbed
dose on a common scale for all ionizing radiation.

The process of  inhibiting continuous or multiple
discharge in a counter tube that uses gas
amplification.
                                                                                              4

-------
RAD:
RADIATION;
RADIATION AREA:
RADIATION. PRIMARY;
RADIATION. SCATTERED;
RADIATION. SECONDARY;
RADIOACTIVE MATERIAL;
RADIOACTIVE WASTE;
RADIOACTIVITY;
The unit of absorbed dose equal to 0.01 J/kg in
any medium.

Energy travelling through space in the form of
electromagnetic waves or energetic particles.

Any accessible area where a major portion of the
body can receive an exposure of 5 mrem in any
one hour, or 100 mrem in any five consecutive
days.

Radiation emitted by the primary nuclear reaction
/ transformation (as opposed to the subsequent
atomic or nuclear interactions as a result of the
primary radiation)

Radiation reaching a given location after having
undergone at least one scattering.

Radiation emitted by some nuclear or atomic
process as a result of previous nuclear or atomic
interactions by the primary radiation  source.

Any material which spontaneously emits ionizing
radiation, or any item contaminated with material
which emits ionizing radiation.

Waste materials that include the following:

a.    property  contaminated to the extent that
      economical decontamination is not
      feasible.
b.    surplus radioactive material whose sale,
      transfer,  or donation is prohibited.
c.    surplus radioactive material that it is
      determined to be unwanted after being
      advertised as surplus.
d.    waste that is radioactive due to production,
      possession, or use of radioactive material.

A natural and spontaneous process by which the
unstable atoms of an element emit or radiate
excess energy from their nuclei as particles or
photons and thus change (decay) to atoms of a
different element or to a lower state of the same
element.

-------
RADIOBIOLOGY;
RADIOCHEMISTRY;
RADIOSENSITIVITY;
REM;
ROENTGEN;
 SATURATION. ION CHAMBER;
 SCATTERING;


 SEALED SOURCE;
 SECONDARY IONIZATION;
The branch of biology that deals with the effects
of ionizing radiation on biological systems.

The aspects of chemistry connected with
radionuclides and their properties, with the
behavior of minute quantities of radioactive
materials,  and with the use of radionuclides in the
study of chemical processes.

The relative susceptibility of cells, tissues,
organs,  organisms, or any living substance to the
injurious effects of radiation.

A special unit of dose equivalent. The dose
equivalent in rems is numerically equal to the
absorbed dose in rads times the quality factor and
any other necessary modifying factors.

A unit of exposure equal to the charge liberated
by x or  gamma radiation of 2.58 X 10-4
coulombs  per kilogram of dry air. It is equal to
the absorbed energy in air of 87.7 ergs/g or in
tissue of 96.5 ergs/g.

The condition in an ionization chamber in which
the applied voltage is sufficient to collect all of
the primary ion pairs, but insufficient to cause
secondary ionizations.

Change of direction of subatomic particles or
photons as a result of interaction or collision.

Any radioactive material that is permanently
bonded  or fixed in a capsule or matrix designed to
prevent the release or dispersal of the material
under the  most severe conditions encountered in
normal  use or handling.

(1) In collision theory: Any ionizing particle that
results from the interaction of primary radiation
as it passes through a medium.

(2) In detector theory: Any ionizing radiation that
is a result of the gas amplification of the
ionizations caused by the incident ionizing
radiation.
4

-------
SELF-ABSORPTION;
SERIES. RADIOACTIVE;
SHIELD;
SIEVERT;
SOURCE GEOMETRY;
SOURCE MATERIAL;
SPECIAL NUCLEAR MATERIAL;
SPECIFIC ACTIVITY;
SPECIFIC IONIZATION;
SPECTROMETER;
The absorption of radiation (emitted by
radioactive atoms) by the material in which the
atoms are located; in particular, the absorption of
radiation within a sample being assayed.

A succession of nuclides, each of which
transforms by radioactive disintegration into the
next until a  stable nuclide results. The first
member is called the "parent," the  intermediate
members are called  "daughters," and the final
stable member is called the "end product".

A body of material used to reduce  the intensity of
ionizing radiation at a  given point by placing the
material between the source of radiation and the
respective point.

The SI unit  for dose equivalent equal to the
absorbed dose in grays multiplied by the quality
factor and other necessary modifying factors.

The shape, size, and configuration  of a radiation
source, taken as a whole.

Uranium or thorium or a combination of both, in
any physical form, or  ores that contain one-
twentieth or more by weight of uranium or
thorium or any combination. Source material does
not include special nuclear material.

Plutonium or uranium enriched in the isotope 233
or 235, and any other  material the  NRC
determines to be special nuclear material.

The total activity of a  given nuclide per gram of a
compound, element, or radioactive nuclide.

The number of ion pairs produced  per unit path
length of ionizing radiation in a medium.

A device or instrument,  usually electronic,
capable of measuring the energy distribution  of
nuclear transformations.
STABLE ISOTOPE;
A nonradioactive isotope of an element.

-------
THIMBLE IONIZATION CHAMBER;



THRESHOLD DOSE;


TISSUE DOSE;


TISSUE EOUTV. ION CHAMBER;
TISSUE EOUIV. MATERIAL;
TRACK;
VOLUME. SENSITIVE;
X-RAYS;
A small cylindrical or spherical ionization
chamber, usually with walls of tissue equivalent
material.

The minimum absorbed dose that produces a
detectable effect.

The absorbed dose received by tissue in a region
of interest, expressed in rads.

An ionization chamber in which the materials of
the walls, electrodes, and gas are so selected as to
produce a response to radiation similar to the
response  of tissue.

A liquid or solid whose absorbing and scattering
properties for a given radiation simulate as closely
as possible those of a given biological material,
such as fat, bone, or muscle.

The visual manifestation of the path of an ionizing
particle in a chamber or photographic emulsion.

The portion of a detector that responds to a
specific radiation.

Penetrating electromagnetic radiation whose
wavelengths are shorter than those of visible light
and originate external to the nucleus. These rays
are sometimes called roentgen rays after their
discoverer,  W.K. Roentgen.

-------
RADIOLOGICAL
         REVISED EDITION-
         JANUARY 197 O
        U.S. DEPARTMENT OF
     HEALTH. EDUCATION. AND WELFARE
        Public Health Service

-------
36
                                                                              He
                                                      (MTW)
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                                            oOi«r« (M01AS4)
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-------
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-------
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-------
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-------
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                                                                   . Anlcli6».
.. Cr.vCSZ.  l
   , W.fM}l.
                                                                                      .  Do)Vi3.
                                                                                                           . OVriC40b. MolSJf.
                                                                                                                      . U.tRJO,
                                                                                                      . OrabCSS.

-------
IftMOpe
7 A •
«'•" i
27-"
Cc"
C*SS
Oc"


c.57


Co"
Co*""
fisl'

c."

C.k0m




H.II-Wc
4.0 m (ltraJt.4,
Meet SI |
ouer> (RtccE55)
0 194 i (FreeJti)
other. (U.rtiWii,
i*iO5t. TyrHM)
1.5 m (SutDSS)
It. 2 >> (D«rB37)
17.9 ri |RudCS2)
lt.0 h |ljv.>4l)
77.3 d (WrlH57)
77 d (burfW54)
olher> |CookCS42,
LivJ41)


270 d (JJvJ4»
267 d (CorkJSS)


71.3 ct (ScnumH5o)
71.0 ct (GorUSS)
724 (Uvj4i, Hoirosi,
Pr.IltO)
9.2 k (Chri.DIO)
9.0 h (PrtUbO)
8.1 h (SlraK50)


5-ttl f |GorhSb3)
£.14 y (&.IKW57)
5.20 T (U,tE5t)
5.11 j fKaiJSU)
5.27 y (ToaJSi. TokJSl)
•Uwra (U-cESJ,
UvJ41, BrowCJO.
Sli>WSl>
10.41 m (hirtWlSJb)
IC.Jm (SehmWt»
IC.Srn IPrcUtO)
1C. 7 m (JUO41)




Type ol Occa* ' ^*
9b apwnOKocc. M»Aft cxccai
(/i.M-A). MfV (C"'0;
Thermal ncuiroc »
crou tecixxt (7). b»if>i J
1
If p" (RkccCS5, R>ceE57)
t -5» (MTW)
¥ P4 (M«rtiWS2)
^ -47.99 (MTW)
V p4 (SulD59)
1
¥ p4 81%, EC 19%
(MUKA58)
f," -fcO%, EC -40%
(c»lc from I>euM49)
i -M.O) (MTW)
V EC 80%. S420V.
(CookCS5t)
fi -5t.OJ (MTW)
i

¥ JEC. no p4, llm 0.002%
i (CruBSS]
t i-59.339 (MTW)


V EC 85%, p4 15%
(OooW46, CookC556)
A -59.84 (MTW)
rf 2500 (Co)dmDTo4)
V IT. nc p4 (Str*K50)
ft. -59.81 imp. MTW)
r£ 1.4 x 1C5 (Co:dmDT64)
% 100 (MiU41)
A -62.U3 »rB51)
cK«rr>. croak bomb, fenet
(UvJ4l)
p*rem F«5i (UvJ41)
A ch«m, •xcll, crosi bomb
(UvJ«);i
diufhler Ni* (ShelRSi.
\VorW52)


A ch«m, txcll, croe* bomb
(UvJ41)
diujht.r NiS1 (rrleC52)


A chem, tKCil, CTO»* bomb
(UvJ41|
A ch«m, excu (Str»KSO)


A n-e»pl (Sw»M)»)
ck*m, util. tro«i bomb
(UvJ41)

A n-capt (Herr37»)
ch*m. K»cii, cross porno
(LWJ4I)
cUurhler F7. 0.024
i i
i
1 i
:
il
I i
1 J
f } 1.41 max (0.12%), 0.114 ma*
, | f»»«%)
Y (1.173 (100%), 1.1)1 (100%)
i
j
p ! 1.55 max
t~i 0.051, 0.05J
Y Co X-ray.. 0.059(1.1%), 1.3J
(C.15%)


1
1
f-fifW«pal meant |
ol Droowctioc ^
Jr*4)!...) (RictC57j
•It*4!*, ap) |RiccCi7)
FeM(p,n) (Fr«tjbi.
MarliWS2|
FtM(p,n) IFr««J65)
FeM(d.n) (D»rB57.
LivJAl, D«uM49)
F«M|p, Y) U.IVJ4I)
Feifc((i,2n) (MukA5«)
Fe"(p,n) IKiePS?.
CrabloOl, &>kM>4)
Mr,55(a,3n) (CncnL52t)
daufhttr Ni1' |ShclR52,
VorW^2)
FeSfc(d. in) (UvJ4l,
JeniA4J. PlcE42,
CllIL43a)
NIH(d,»l (UvJ41,
CookCi4i, ElHMll)
N!"|Y.p);
Ft^ld.n) (LlvJlli, ,
P.rrCJE, BarrC39,
UvJ4l) ^
*•«**•». M (UvJ41)
MnSi(a. 2n) (Ch«nt,52a)

MrSS(a,n) (UvJMa,
UvJ41)
Mt.55(»,n) (Str.KSO)


Oo*'l», t) (RUJ37,
LivJ3U. IJ»J41,
fc.rUTb, YatUl)

»
C«"(i>.t) |H«rD1a.
UvJ}7». 1J»J41.
ScrLX7b)





-------
266
(iKNOpc
M
^f_





ill1


i,Ml"





ST



.,•'





1
fl Sr8*"
Sr»°





Sr"






Sr"
Sr

Sr*


Half-life










Z.»3 k IBormMfcS)
2.10 k IM.nnLSl,
HydESl)
2.et k |GravGS2j
otkcrt {H«rrmGSf>,
Dubl»4 0)






52. 7  (KnlJDS?)


4^ eDuAOanct. Maat rxcoi
Tnernul nrwiroo
crau KCuan (D, bum
e. t.t* (Ni.rAXb)
«.I7 lAldLSJ)
A (-»*.4»» (MTW)
, 1.3 (U.Sr'1rn)
(GeldmDTM)

1
% 7.02 IKKrAJIb.
AldL5)j
A -»V»t5 (MTW)
¥ IT »«,%, EC(K) O.t%
(SunA60)
A -P4.477 (LHP. MTW)




f. t2.5fc (N.irASSb,
AJdLSJ)
ii -87. H (MTW)
r( 0.006 (OolOrnDTM)
V p" (Sl«»D37)
A -«t.22 (MTW)
»t 0.4 (Go)dmDTMI






f p' (NotRSJ)
6 -«i.« (MTW, LHP)
»c 1 IGoldmDTM)





•t p'-(Go»H41)
£k -Bl.bE (MTW)






V P* (GOIH41)
A §l.» (MTW)
* " (UeC3»)
A 7».4 (MTW, Sl.lnEtS)

V " (Hak043b. HahO43)
6 7».»'(MTW)


Oencuc rclattoAjhipi


i
1
i
i

• *^
i (HaKOJI, MattaJ37)

A f chcm, axclt (SlewD37)
ckem, exctt, croti bomb,
'• f«n«l (Dub 1^40)
i DubL4C, MannLSC,
' ManrtLSl Jumd*tM5t»4
I HydCSl) '
•
!
i


A : ckem, eKclt (St*vO37)
ldau|hterRb (Gla»G40.
! HahO40, HaliO40b, HahO4),
! GrumW4t)
i parent Y 0.009% (&alAb2);
: 0.02% (LyoWiSb); C.01%
! (HtrrmCSt);  (HakCMS,
! HahO43b. KnUOSt)
ei
}dau|hler Rb (FrltKtO)
Aj chem. (enet (HahOOti,
: HahO43)
iparent YW (HahO43.
• HakO43b, KnUD5»)
;
j
i
>pproxim>u nxitici (McV)
and uuauiiia
:









Y Sr X-rayi, 0.311 (10%)
e" 0.372, 0.3«fc








p" 1.443 max
Y O.«l (O.OOT%, »iOi Y**1")






p~ O.S46 max
Y tie Y
daughter radiationi from Y





.
p 2.47 max
Y 0.445 (15%). 0.746 (27%).'o.«
(>%). 1.02!. (JO*), t.413 (S%)
daufKter radiationi irom y"m.
v°l




p' l.Smax (10%). O.SStMK
Y B.U (3%). 0X4 (4%). 1.37 (««%)
daufhter radlatlaix Itwn Y12
p" 4.» max 1 (weak). 2.1 max
Y 0.40. C.l. 1.2. »tker» between
0.2 and 3.0
daughter radialien* Irom Y

p" 2. 1 max
Y 1.42 (100%)
daughter radiatieni Irom Y


PfinCipal meani
oi pradunton










daufhlir Y" (DubL39,
M»n«L5C, MannL5l|
Sret(n, Y) (S»«wD37.
DobL3S, HedH40.
Ka4H40a)
07
Rb (p.n) IDubJ.3.)





Sr*B(d. p) (SI*wO)7.
Sr**|n.Y) (i«rL4Tb.
St«wO37, St*wO3*|






<»iion (ODC51. DUC5U,
KofOSlb. GrumVV44
GrumW4e)





(mior IGotml, HahO43.
FwBSl. KatcS4E.
FlnBSlc)
it |n, a) (i««W4jb)





fieelon (Ha»O4C.
HahCHJ. HahO43b,
KatcSiU. BradEil,
KatcS4l)
If" In. a) (VaUlDtl,
BakHtS)
fioion (UeOT. HahO4Z.
HahO43, KnUDSt)

(»akon IHahO43.
HahOOb. DDCS1.
KniJOSV FritK4l,
HovDH)



-------
IKKOBC
Z A
*( ,,
Tt
43



Tc'7



TC"-"


Tc"




Tc"







Tc"™


Tc'00



Tc101


Tc>"



102
Te* '



Tc'W



Te1"



Tc105




HtlMile
Urn (M.AXSC. t..H53)




2.4 x 1C y yield
(KateSie.)
other. (BoydGMj


91 d (BoydCM,
HelmhA4U)
•0 d (MetE48b.
CufP47, CacB37)
17 c |UnlJS9)
9i 4 (E«wJ47)

l.S X 1C y «p act
(OK.lC4»b|
other. (KalcSSS)


2,12 x 10S y .p act
IFrieSSl)
2.1>x 10S y (pact
(BoydCbO)




».049fc (CleCM)
b.OOh (ByiDit)
other. IGleLilc,
BallCSJ, PorlRtO,
GreT»S|

15.8 t (BoydCSZn)
17. i > (HouRiJ)
17 • (C.iCtS)


J4.0 m (OKelCS7,
MauW41, HabO41b)
H.3 m (WUeDRM)
14. 5 m (Perln>M48)
16.5m (MaeD«S)
4.Sm (Fl.JM, F1.JS7)




* i (FleJMl
other. (HahO41a)


SO • (KiePtU.
VBacAtS)
72 • (F2eJS7)


It m (FUJSki, KiePoZ)



7.7 m (KitPtU)
7.8 m (VBaeAtS)
>0 m fenet (HeJi&a,
ricJibi)



1 ypc 01 *tCT v ( *•» ; :
% Bouodancc, MaAt excess |
l£tM-A) MrV (T'-O), |
Toenul OCHUOX
crou MCuan (0), to. mi
!
y jrr (M.W»OI
p4 -0 01% (Ea.HSJ!
A -»5.» (UiP. WTW)


y EC (BoydCiAj
ii -87 (MTV)


V IT  -«7.33 (MTW)

r \H (OolojTtDTM)




y IT (SeaC39)
A -87. l» (LHP. MTW)

V p" (Hc.uRS2)
A -85.9 (MTW)


'.- p" (SafR40)
£, -8t.ll (MTW,

v p* (nust»)
A -8S (MTW)



t f~ (HanCHU)
A -8S (MTW)
j
1
y P* (KltPUb)
A -M.9 (MTW)


V p" (D.JSb*. KI.P62)
A -12.2 (MTW)


y p" (BornH4Jo)
6 -tt.t (MTW)



i
Oats, loeftvficviion ,
Genetic rctaitoAthip*
B Ick.m, .xcil (MadHSO)
jeh.m, ««CU, >ep i.otop.1
i (Ma«H»2)
i
!

A tfantt (BoydCila)
|ch«m (KitcSiki)
iliaufhl.r Tc"m) (BoydCSU)
(daufhter Ru (994%)
(KatcS&ta)
A chant (P«rrC37, CacB37)
eh.tn, f.net (MOIE47)
cxcU, »ap itotopcl (Mot£4ftb)
dau|ht«r Huf1 (0.04%)
(K.lcSibt!

A eh.m, ma., .p.ct (BoydC55)




A chtm (UncO4t, SchumR4tl
ch.m, ma., .pect (ln|M47cl
9%p*\
daufhter Tc"™ (S.aGSf,
HahO41a)
99
de.cendant Me (MO1L47.)


A chem, f.nct (&eaG)9)
daughter Me** (Se»C39.
SafK40a, MtdH49, ClcLild.
MihJSl)
CB
parent Tc (S..O39,
HahCMla)
A aen iaaup*. (HouRSZ)
• ep i.otopc.. n-capt
(BoydG52j)


A chem, penet (SafR40)
ihi
daufhter Me ' (BotW41,
HahO41a, HahCXlb,
MauW41, Sa^R40)
:
Bichem, fenct energy level.
i |r>eJSka. rieJST)



C eh.m, »«n«l (HahO41a.
ricJM)
IB)
idaurhter Wc'ui (HahO4U,
HahO41b. rjeJM)
B excit (naJ57)
)ehen>. f~>.t fKi.P.le)
• 1BV
i [pa rent Ru' ^ (Kl.PtJa)
j daufhter Mo>0> (KUPti»)
Bichem (FleJSU)
jchem, fenet energy level.
j (KitPl2)
1CM
jdaufhtcrMe (KiePb2)

i
Bichem. r«"cl (BornH<3b)
: i0£ tn
jvarentRti , dau|thier Xto
| (BornM4)ti, ricJSSa,
{ KicPbU)
] 1D^
(ancour RJ-. (KiePila)
j
i
i
M»«?f f»«ua',«om '
• (>proximaic rnc'pitn (MeV) j
• ltd imtfMUKl
1
v Tc X-«r« 1
0.01J. O.OiZ
96
cUufhicr rAdiAtioni (rom Tc
I

V Mo X-r*y» |
j
1

Y Tc X-r»yl
0.075, O.tm
i
1
1
I
p' O.JO max
Y O.bk (100%), 0 7b | 100%)


!
(S~ O.l^i m*x
X we Y

'




Y TcX-rayi, 0.140(90%)
e" 0.001, 0.115

p~ 1.36 max
Y C.&40 (.tronj), 0.60 (.trongl,
0.71. O.H. 0.8S. 1.01. 1.J1.
M*. J.t

p~ 1.12 max
Y 0.13 (3%. complex). 0.107
(Y 91L), O.MJ (Y 8%)

p~ 2 max
Y 0.«7



P 4.4 max



P~ 2,cnuuc
Y io.ISi (T 17). 0.21 (T 10), 0.)5
i
j
i
i
p" (S.tmax) («>«ak), 4.tmat
Y 0.36, 0.53. 0.89, 1.15, 1.25.
1.37, l.t (complex), 1.9. 2.2
2,7, J.I. 3.4. 1.7. 4.0, 4.4. 4.7

p" >.4 max
Y 0.110
cUufhler radiation, (rom Ru


i
F*(incip»l mrani ^
ol production •
r4k*5|..i>) (E*.H51)



96 97
Ru (n. Y|RU"(P |
(K.tcSibi)
Me (d. 2/i) (BoydCM)


We*fc(d.n) (CacBll,
PerrC37, CacB39)
klc"(p.n) (t*wJ47)
Mt*7|d,2«i) IMoIt4ab)
D/ 47 .
Ku In, YIRu (p )
(Kaic5!.lla)
McM|p,n) (BoydGSS)
Ki,"ln. YIRu'^p")
97
Tc (r.,Y) (K.lcSSS,
K.lcSSb.)
ft. .ion rin|M47(,
1.WCO51. SchumR&l)
Ob Ob „
Me"(n, Y)Mo"(p )
|MotE47a)




daufhter Mo*' (ScaG>°
S«jR40», McdH49,
ClcJ,i)d, WlhJil)

Tc"(n. Y) (BoydGSi*..
OKtlCSej
100
Me (p.n) (MouRS2)
Rn103(t>, o) (C.lCtS)
1OO 10 1 o.
Me (iv. Y)Me (p )
(S«rR4C, S.tR40b.
MauW4I)

fKu'C2(iv, p) |rieJS7)
(...ion (F1O5U)
1

102
Idaufhler Me *
1 |HahO4)l. HahCKlb,
1rn« i *.A\
t **•* «n/
I
1
n..ion (KI.P4U.
Kl«Pt3b, VB*.A4$)
IftJ
Ru""(B.»p) (FJeJ57)

(...Ion (ritjibi.
K.cPb2)
R<.lt>4(n.p) (r>eJS7l

M
d.iion (BornX4)t>, V
rieJ5i«. rieJSe*. ^
K.ePbU, VB.tAbS)




-------
! iKKODC
L '
_,c.m
1!


1 JSm
Ci


C.1U




c."7









c,1M



C.1"


C."8

' c."1



c.'«

c.">

c."*


56B*

_
Half-life
).0 x IP y ap act
<2.alH4»)
1. 1 x 1C y yield
(»tlfaN4Va)

Hallcl»4)


13.1 d (ClelXV)
12. « d (Ol.Ji*.!
IS. 5 d (WUleRkO)




30.0 y (~«l|h(«d
averaft by FlyKtf
29.1 y (CorbS63)
30.4 y met* ipcct
19. 1 y mat* apect
(RldthtS)
>0.0 y ap act, maa«
apect (BrowFSS)
other* (FlyKtS,
FUDbii, WlieDMSia,
QatVI6l, WUeDR.53,
CltLSlJ)
JJ.2 m (BarthRSfc)
32..) m (SunkMSb)
other* (Gla*O4C,
Wlll.RtC, EvaHBSl.
AteA)9, HahO39a,
ClcLSlk, OekDb2,

S.& tn (Suf aNSO,
ZheCtl)
othcri (AteA.3?.
HeyFJS, OekDb2,
HahCMO)

it i (Su/aNSO)
tj > (ZheCfcJ)
24 a (FritKtia)
2i a (WaaAtZ)


Z.3 i (FrltKoU)
"lESHl?***-
2.0 > (FritKtU)

•Kort (OUCS1, DllCSla)


2.0m (PreU62)


T rpc of »m v ' *f* ;
% aounaancc. Mui cxcui
Thermal ntuiron
mu aection (C). barm
i .
A; -87.8 (MTW)
t 1 (GoldmDT*4)
c

C -«1.2 (MTW, l«HP)


V p" (CleJ-Slt)
i> -8t.b (LHP. MTW)




•t p" (MelW-141)
A -tk.1 (MTW)
r£ 0.1) (OoldmDTM)







V p" (HahOJVc)
^ -81.1 (NDS, MTW)


•-* p" (HahOSSc)
A -81.1 (MTW)

V p" (HahCMO)
£ -77 (MTW)
" IP") (BradESl)



* IP") (FrltKbU)

f IP") (BradCSl)

v [p") tmicsi)


•.' IP*. EC) (PreUfcZ)


-==^

charr.. maa* apect (lnfM4*j
Aaufhtar X« (»ttfaH4w)


bomt. crtl abi (WarhH42)
chirr., ma*. «pect (Halltlt^)


A ch«m (CUMt, CUL.SIX)
chem. exelt |C)«L.4«)
chero. maa* apett (O)tJS4t)




A cherr.. fa»tt (M«lhW4l)
chem, ma*» apcci
daufhler X.'51 (TurASl.
f-\ t \ I > W \
ull «U3 *KJ
parent B.131m (To«nJ4«|






A cham (HahO3»c, HeyFJ»)
daufhttr Xe J>* (HahO)fc,
HahO40, CU.C40.

A chem. fenet (H»hO3»c,
daufhter Xt1" (HahCHfc,
HeyFlf, H»hO40i,
HahCMO)
HahCMO, Su»»N50)
A cham (HanO40)
chem, t*"" (SufaNSOJ
A! cham. fanat (WaXA.t,
j FritKkUl
! 141
Mceetor Ce>4> (FrltK»i»)
i
S j cham, fanet (FrivKbta)
ianceator Lal" (OUCSl)
{do cendant Xt . anceaior
Ct'44) (DUCSI)
B chem, croaa bomb, tenet
(PrcUt2)
parent C« (Prctlb2}
Ma»or radiation!
aporoximau enertict 41 ma*
e~ O.IU, O.Ufc, O.lil, O.>02
Y Ba X-ray*. 0.0t>7 (11%). C.OBt
(6T.). C'.)fc|)fc% complex}.
0.27) '11%). 0.340 15)%].
O.eiC (100%), l.Oi (82%), 1.25
daughter radialionk IrOfrt Ba TT}
included tn above liatinf,
p" l.mmaje (7%). O.il4 max
t* O.»l4, O.tSt
Y BaX-rayi, 0.it2(B5%)
Included in above Uilxnf






.
p* J.40 max
Y 0.46) (23%), O.S5 f8%), 1.01
(257.). 1.4*4 (73%). 1,11 (18%),
2.6) (f%)


Y D.50. 0.6S, 0.80, 1.28 |atron{).
1.65 (complex), 1.90, 2. Of
daughter radiauoni irom Bfc

Y 0.5$. 0.86. l.K. 1.62. 1.8S. l.Ok.
2.J2, i.72. 1.15














Pnncipal mcani
*...«., X.'Ji
b£y«*£\Mi}


134
132
X« (a. p| (»arhHt2|
. I3i
B* ln,p) (ViarhH62)
proton* or, Bi (HUlell>4)
L* In. a) (CamM44.
CUiXf, Strn.HfcJ)
B»'J'ld.a) (GirRSV
OrablbOb)




Itiaion (HaydR4t.
ln|M4f. GULMJ,
GrumW4e, TlnBilc)








fiaaiox (HahO3»c..
HahCMOa., H.yDS.
HahCMC, BunkM»6)
»»'J*|n,p) fWUleRtO,
«>»)

luiion (HahO)Sc,
HtyFJ9. HahO40a.
Alt A3?. &u|aKSC,
HahCMO*. Hah04C,
AkiV62, rheCbl,
OCX.D62)

(tiiien (HahCMO.
iufaNSC. ZhtEtJ)
[
-------
351
1
IMMOCC
Z A

T])»ir.
*'


T,199





T,200



TJ201




TJ«2

T,*"


n1 (KrUUOb)



12.0 d (HameH57)
ether. (MirtiHC52,
WllkCSOb, r*JK4i.)



3.«1 y (LeuH»2)
3.10 y (rUrbC63)
3.78 y (FinH.59)
3.91 y (W«JiAS9,
NUR&2)
3 »B y (FlyK»S»)
other. (EdwJie.
MerW57, TobJSic,
Wya£61, HorrDM)
SpeaH64)



4.19m (S«rgB53)
4.»m (F»jK40)
other. (FvuA5V,
AlbuD5U. P001M37,
H«yFS7)




4.79m (SatfBSJ)
4.76m (CurlMJl,
SargB39»)
other. (F.JK4C,
BretE40, BaldG46)


1.3 • (EccDbS)

* uxiaduoc: MmunccM
(£iM-A), M«V49)
me.i ipect (MicMM)
dvifhter Pb"° IN.umHSO.)
de.cendent Po2°° (BnmC6i.)


A chem m».. .peet. ..net
(Iob>£59, rlerrlCbO)
chem, excil, cro.i bomb
(N.umHSO.)
deufhter Pb (NeumHSOa,
JOB.B59, H«rrlC60)
(BrunCate)
A cbem, exclt (KrUUOb,
FAJK4U)
diufhter Pb202 (HuUM)



A chem. Ti-c»pt (Fe)K40)
m... epect (MicMM)






A n-c.pt (PrelP35)
ehem. fenet (BrodE47)
•xcit, >ep Uotopxi
(NeumHSO)
id*a(bt*r B1Z>0 (KaX)
{ (Bi*dE47)
dn>|bter BS2l0m (NeumHSO,
|d.ufhter H|*°' (NurMtl.
j K»uP62, WolfCKM)
A ! ch.m, j.net (CuriM3J)
d»ufhtcr BIZ (AcC)

i
i
i
i
E jexcit (EccD65)
i
M%), 0.579
|)0%), 0.129 («%), 1.21 (35%,
complex), 1.364 (4%), 1.410
(1.6%), 1.517 (4%), oth.r.
p* 1.44 m.« (0.06%), 1.07 max
(0.3%)
.' O.U5, 0.354
Y HI X-r.y. , 0.135(2%), 0.167
.' 0.016, 0.052, O.OM



Y H» X-r.y., 0.439(95%), 0.522
(0.1%), 0.961 (0.07%)
." O.J56



p~ 0.766 max
Y Hf X-ray.






p' 1.S2BU
Y so Y





p' 1.44
Y 0.<77 (0.16%)




Y 0.35, 1.00
i

porpn«»

Au'^fe. >») (n.cP5e,
MicMM, BrlnOOST) '


A\j"7(«, 2n) fVVIJR63)
H( (6, in) (KriK40t>)




deuteron. on Hf
(KrlR40b, VMooBti,
GupKtC*)
Au"7(e,B) (OrtD49l
(S*kM65)

dAufhter Pb
|NeumH50a)
deuteroni on Hg
(KrUUOb, UnfdtM)



Hi*°Z(
-------
367
ItOtOfX
Z A
«CTI>




ThZZt





ThZZ7
(RdAc)





TkZZ«
(RdTh)


»«





Th""
(1*1






Tb"1
(Wf)


it!!!





»%*
TbW>



Half-life
1. Om (MelWSl)




JO. 9m (StuM4B)





IB. 2 d (Haf«G54)
other. (PeteS49b,
CurlM31)





1.110 y (KirH>6)
other. (CuriMJl)


7340 y fenat (HafeFSO)
other* (EnflA47)





B.O x 1C4 y ip act
(Hy1CZI y (neGSB)
Other* (PocASS,
SefES2)

22.12m (JenkCSS)
22.4 m (DroB57)
22.3 m (BunkMSOa)
22.5m (SeaG47)
other* (RUIW52.
Gre**A41)


% aouadmcc; Mat* cxccu
(A«M-A),M«V (C""0),
Thermal neutron
crau aecuoo (O), banu
*ia .90%, EC -10%
(MelWSl)
£ 22.30 (MTW)



t o IStuM4B)
ft *table (con. enerfy)
(ForB58)
£ 21.19 (MTW)



V a,
ft .table (conk enerfy)
(ForBSS)
L. 25. B2 (MTW)
,, -1500 (GoldmDT64)



Y a,
A (table, (con* anarfy)
(ForfiSB)
A 26.77 (MTW)
»c 123 (O»ldmDT64)
r( <0.3 (ColdmDT64)

*ia;
If aUMl (con* enerfy)
(ForfeSi)
A 29.61 (MTW)
t, 32 (OoldmDTM)
I

* «;
B (table (con* enerfy)
(ForB5ft)
& 30.S7 (MTW)
rc 23 (ColdmDTM)
,. S0.001 (GoldmDTM)


* p":
& 33.13 (MTW)


V v
A etable (ooa* etnarfy)
(ForBM)
% tOO (A*tF35. DempA36)
A 35.47 (MTW)
re 7.4 |C«ldn>DTo4)
r, C0.0002 (GoldmDT64)
I

* »" * (A«V) ,
param Pa"1


Aj Cham, fanat (CurtMJl)
! parant RiZZ* (M.Tkj)





A chem. n-capt (MHL3B)
parant P»233 (MelL38,
CroaaA41, SeaC4U,
HahO41, SaaG47)


apfwoaimau encrf w> (McV)
and Hicentiuef
a t.ao (8%), 6.75 (6%), t. 50 (12%),
4.48 (39%!, 644 (13%)
Y (Ac X-ray.). R, X-ray*. 6.24t
(5%), 0.322 (27%), 0.362 (5%),
0.45 (1%), 049 (1%)
22 1
daufhter radiations Irom Ra ,
• LC

o b.34 (79%), t.22 (19%)
\ RaX-ray», 0.111 (3.4%). 0.131
(0.34%), 0.20 (0.4%, complex),
0.242 (1.2%)
e" O.OW, 0,107
222
daughter radiation* from Ra ,
RnZ1B, .u
a 6.04 (23%), 5 9B (24%), 5 76
(21%), 5.72 (14%. doublet)
Y R* X-ray. , 0.050(8%). 0,237
(15%, complex), 0.31 («%.
complex)

e~ 0.013, 0.026, 0.044, other.
223
daufhter radiation, from R* ,
RnZ", PoZ)5, etc
a 5.43 (71%), 5.34 (28%)
Y RaLX-rayi, 0.084 (1.6%), 0.132
(0.2%), 0.167 (0.1%), 0.214
(0.3%)
e~ 0.067, 0.080
daufhter radiation, from Re ,
Rn , Po , etc.
a 5.05 (7%), 4.97 (complex, 10%),
4.90 (11%), 4.84 (58%), 4.81
(11%)
Y Ra X-ray* , 0.137 ("3%.
complex), 0.20 (>10%, doublet)
e" 0.006-0.090
daufhter radiation* /rom Ra ,
Ac"6, etc.
a 4.68 (76%), 4.62 (24%)
Y Ra i. X-ray. , 0.068 (0.6%), 0.142
(0.07%), 0.184 (0.014%), 0.253
(0.017%)
e~ 0.051, 0.064
daufhter radiation, from Ra ,
222

p" 0.30 max
e~ 0.040, 0.054, 0.061
t Pa 1. X-ray., 0.026 (2%). 0.084
(10%, complex)

a 4.01 (76%). 3.95 (24%)
Y (Ka L X-ray* ]
e~ 0.042, 0.055
daufhter radiation* from Ra ,
AC228 Th22B ^224 >t(.




p~ 1.23 max
e" 0.009, 0.024, 0.036, 0.0 5J-. 0.067,
0.082
Y PaX-raf*. 0.029(2.1%), 0.087
(2.7%). 0.171 (0.7%), 0.195
(0.3%). 0.453 (1%), 0.67
(0.25%), 0.895 (0.14%)
T£=
daufhter UZ2' (MetW49,
MelWSl)




daufhter f250 (HydEo4)





daufhter Ac , from
natural vource or trom
(HydE64)




natural *oarce
daufhter U
RaZZt(n,Y|RaZZ7(p")
227 226 -
At"'(n,Y)Ac "(B |
(HydE64)

daufhter U233 |HydE64)





natural eource (MydE64)







Th (n,1f) (BaranS60,
Ho>taM66)
daufhter U23S


natural ecniree (HydE*4)





>\ >
Thii4 (»,•») (MelLIt,
SaaG47, SeaC41a,
CroaaA4.1).



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-------
REGULATIONS AND PROCEDURES
                                                      OCCUPATIONAL SAFETY AND HEALTH
                    1910.96—IONlZING  RADIATION  INTERP
(a) Definitions applicable to this section.

  (1) "R.adiation" includes alpha  rays, beta
  rays, gamma rays, X-rays, neutrons, high-
  speed electrons, high-speed protons,  and
  other atomic particles; but such term does
  not include sound or radio waves, or visible
  light, or infrared or ultraviol . light.

  (2) "Radioactive  material" means any
  material which emits, by spontaneous nu-
  clear disintegration, corpuscular or elec-
  tromagnetic emanations.

  (3) "Restricted area" means any area access
  to which is controlled by the employer for
  purposes of protection of individuals from
  i-.xposure to radiation or radioactive mate-
  rials.
 (4)  "Unrestricted area"  means any area
 access to which is  not controlled by the
 employer for purposes  of  protection  of
 individuals from exposure to radiation  or
 radioactive materials.

 (5)  "Dose" means the quantity  of ionizing
 radiation absorbed, per unit of mass, by the
 body or by any portion of the body. When
 the provisions in this section specify a dose
 during: a period of time, the dose is the total
 quantity of radiation absorbed,  per unit of
 mass, by the body or by any portion of the
 body during such period of time. Several dif-
 ferent units  of dose  are in  current  use.
        Definitions of units used in this section are
        set forth in subparagraphs (6) and (7) of this
        paragraph.

        (6) "Rad" means a measure of the  dose of
        any ionizing  radiation to body tissues in
        terms of the  energy  absorbed per  unit of
        mass of the tissue. One rad is the dose cor-
        responding to the absorption of 100 ergs per
        gram of tissue (1 milhrad (mrad)  =  0.001
        rad).

        (7) "Rem"  means a measure  of the  ciose of
        any ionizing radiation  to body tissue in
        terms of its estimated biological effect rela-
        tive to a dose of 1 roentgen  (r) of X-rays
        (1 millirem (mrem) =0.001 rem). The  rela-
        tion of the  rem to other dose units depends
        upon the biological effect under considera-
        tion and  upon the conditions for irradiation.
        Each of the following is considered to  be
        equivalent to a dose of 1 rem:

          (i) A dose  of 1 (r) due to X-  or  gamma
          radiation;

          (ii) A dose of 1 rad due to X-, gamma, or
          beta radiation;

          (iii) A dose of 0.1 rad due to neutrons or
          high energy protons;

         (iv) A dose  of 0.05 rad due to particles
         heavier than protons and with sufficient
         energy to reach the lens of the eye;
Change 31
144.15

-------
OCCUPATIONAL SAFETY AND HEALTH
                                                                            1910.«M»X7XT)
                                                        STAhTDARDS AND  INTERPRETATIONS
     (v) If it is more  convenient  to measure
     the neutron flux, or equivalent, than to
     Determine tne neutron dose in  ratis, as
     provided in subdivision (iii)  of this suo-
     paragraph, 1 rem of neutron radiation
     may, for purposes of the provisions in this
     section  be  assumed to  be equivalent to
     14 million neutrons per square centime-
     ter incident upon  the body; or, if there
     is sufficient information to estimate with
     reasonable accuracy the approximate dis-
     tfibutioi in energy of the neutrons, the
     incident number of neutrons per square
     centimeter equivalent to 1 rem may be
     estimated from Table G-17:

   TABLE G-l~—NEUTRON' FLUX DOSE EQUIVALENTS
Neutron
energy
(million
electron
volts (Mev))
Thermal 	
0.0001 	
0.005 	 	
002 	
01 . 	
05 . . .
10 - ...
2,5 	
6,0 . -.
75 	
10 	
10 to 30 	

Number of
neutrons per
square ci-nUmcler
equivalent to &
oosr. of 1 rem
(neuirons/cm>)
070 X ICX
7-.20 X 1C*
820 X 10»
400 X IP
120 X 10s
43 X 10s
26 X 1C4
•_*) X 10=
26 X 10s
24 X 10s
24 X JO
14 X ID*

Aver.Tffr. flux
to deliver
100 inllllrpm
in 4U hours
(neutrons/cm1
per sec.)
670
500
570
280
80
30
18
20
18
17
17
10

   (8) For determining  exposures to X- or
   g-amma rays up to 3 Mev., the dose limits
   specified in this section may be assumed to
   be equivalent to the "air dose." For the pur-
   pose of this section  "air dose"  means that
   the  dose is measured  by a properly cali-
   brated appropriate instrument in air at or
   near-the body  surface in the region of-the
   highest dosage rate.

 (b) Exposure of individuals to radiation in
 restricted areas.

   (1) Except as provided  in subparagraph (2)
   of this  paragraph,  no employer shall
   possess, use, or transfer sources of ionizing
   radiation in such a manner as to cause any
   individual in a restricted area to receive in
   any  period of  one calendar quarter from
   sources in the employer's possession or con-
 tro! a dose in excess of the limits specified
 in Table G-18:

               TABLE G-18
                                  Rems
                                    per
                                 calendar
                                  quarter
Whole  body: Head  and trunk; active
  blood-forming organs;  lens of eyes;
  or  goneds ------------------------   IVi
Hands  and  forearms; feet and ankles. 18%
Skin of whole body ___________ ......  7>/2


  (2) An employer may permit an individual
  in a restricted area to receive doses  to the
  whole body greater than those permitted
  under subparagraph (1) of this paragraph,
  so long as:

   (i) During any calendar quarter the dose
   to the whole body shall not exceed 3 rems;
   and

   (ii)  The  dose to the  whole body, when
   added  to the accumulated  occupational
   dose to the whole body, shall not exceed
   5 (N-18)  rems, where  "N" equals the
   individual's age in years at his last birth-
   day; and

   {iii) The employer maintains  adequate
   past and current exposure records  which
   show that the addition of such a dose will
   not  cause the individual to exceed the
   a,mount authorized in this subparagraph.
   As  used in this subparagraph, "dose to
   the whole  body" shall  be deemed to
   include any dose to the whole body, gonad,
   active bloodforming organs, head  and
   trunk, or lens of the eye.

  (3) No employer shall permit any employee
  who is under  18 years of age to receive in
  any period of one calendar quarter a dose
  in excess of 10 percent of the limits specified
  in Table G-18.

  (4) "Calendar quarter" means any 3-month
  period determined as follows:

   (!) The first period of any year may  begin
   on any date in January: Provided, That
   the second, third, and fourth periods
   accordingly begin on the same date in
                                           145
                               1910.96(bX
-------
 1910.96(bK4Xl)
                                                       OCCUPATIONAL SAFETY  AND HEALTH
 STANDARDS AND INTERPRETATIONS
    April, July, and October, respectively, and
    that the  fourth period extends into
    January of the succeeding year, if neces-
    sary to complete a 3-month quarter. Dur-
    ing the first year of use  of this method
    of determination, the first period for that
    year shall also  include any  additional
    days in January preceding; the starting
    date for the first period; or

    (ii) The first period in a calendar year of
    13 complete, consecutive calendar weeks;
    the second period in a  calendar  year of
    13 complete consecutive weeks; the third
    period in a calendar year of 13 complete,
    consecutive calendar weeks;  the  fourth
    period in a calendar year of 13 complete,
    consecutive calendar weeks. If at the end
    of a calendar year there are any days not
    fallinp within a  complete calendar week
    of that year, such days shall be included
    within the last complete calendar week
    of that year.  If  at the beginning of any
    calendar year there  are days not falling
    within a complete calendar week of that
    year, such days  shall be included within
    the last  complete calendar week of  the
    previous year; or

    (iii) The four periods in  a calendar year
    may consist of the first 14 complete, con-
    secutive calendar weeks; the next 12 com-
    plete, consecutive calendar weeks,  the
    next 14 complete,  consecutive calendar
    weeks, and the last 12 complete, consecu-
    tive calendar weeks. If at the end  of a
    calendar year there are any days not fall-
    inp within a complete calendar week of
    that year, such days shall be included (for
    purposes of this section)  within the last
    complete calendar week of the year. If at
    the beginning of any calendar year there
    are days  not falling within a complete
    calendar week  of that year, such days
    shall be included (for purposes of this sec-
    tion) within the last complete week of the
    previous year.
 (c) Exposure to airborne radioactive material.
   as to cause any  employee, within a  re-
   stricted area, to  be exposed to airborne
   radioactive material  in  an  average
   concentration in excess of  the limits
   specified in Table 1 of Appendix B to 10 CFR
   Part 20. The limits piven in Table 1 are  for
   exposure to the concentrations specified  for
   40 hours in any workweek of 7  consecutive
   days. In any such period where  the number
   of hours of exposure is less than  40, the
   limits specified in the table may  be
   increased proportionately. In any such
   period where the number of hours  of
   exposure  is greater than 40, the limits
   specified in the table shall  be decreased
   proportionately.

   (2) No employer shall possess,  use, or.
   transfer  radioactive material in such a
   manner as to cause any individual within
   a restricted  area, who is under 18 years of
   age,  to be exposed to airborne radioactive
   material  in  an average  concentration in
   excess of the limits specified in Table II of
   Appendix B to 10 CFR Part 20. For purposes
   of this subparagraph, concentrations may
   be averaged over  periods not greater than
   1  week.

   (3) "Exposed"  as  used  in this paragraph
   means that the individual is present in  an
   airborne concentration. No allowance shall
   be made for the use of protective clothing
   or equipment, or particle size.


(d) Precautionary procedures and  personal
monitoring.

  (l) Every employer shall make such-surveys
  as may be necessary for him to comply with
  the provisions in  this section. "Survey"
  means an  evaluation  of the radiation
  hazards incident to the production,  use,
  release, disposal, or presence bf radioactive
  materials  or other sources  of radiation
  under a specific set of conditions. When
•  appropriate, such evaluation  includes a
  physical survey of the location of materials
  and equipment, and measurements of levels
  of radiation or concentrations of radioactive
  material present.
  (1) No employer shall possess, use or trans-     (2) Every employer shall supply appropriate
  port radioactive material in such a manner     personnel  monitoring equipment, such as

19)0.96(dK2)                                 146                             '

-------
OCCUPATIONAL SAFETY  AND HEALTH
                                                                               l»10.»«dXt)
   film badges, pocket chambers,  pocket
   dosimeters,  or  film ring-s, to, and  shall
   require tne use of such equipment  by:

     (i) Each employee who enters arestricted
     area  under such  circumstances  that he
     receives, or is likely to receive, a dose in
     any calendar quarter in excess of 25 per-
     cent of the applicable value specified in
     paragraph (b)(l) of this section; and

     (ii) Each employee under 18 years  of age
     who enters a  restricted area under such
     circumstances that  he receives, or is
     likely to receive, a dose in any calendar
     quarter in excess of 5 percent of the
     applicable value specified  in paragraph
     (b)(l) of this section; and

     (iii)  Each  employee who enters a high
     radiation area.

    (3)  As used in this section:

     (i) "Personnel monitoring equipment"
     means devices designed to be worn  or .car-
     ried  by an individual for the purpose of
     measuring the dose  received (e.g., film
     badges, pocket chambers, pocket dosime-
     ters, film rings, etc.);

      (ii) "Radiation area" means any area,
      accessible to  personnel, in which there
      exists radiation at such  levels that a
      major portion of the  body could receive
      in any 1  hour a  dose in excess  of 5 mil-
      Hrem, or in any 5 consecutive days a dose
      in excess of 100 millirem; and

      (Hi) "High radiation area"  means  any
      area, accessible  to personnel, in  which
      there exists radiation at such levels that
      a major portion of the body could receive
      in any one hour a dose in excess  of 100
      millirem.

  (e) Caution signs, labels, and  signals.

    (1) General.

      (i) Symbols prescribed by this paragraph
      shall use the conventional radiation cau-
      tion colors (magenta  or purple on  yellow
           STANDARDS AND INTERPRETATIONS


    background). The symbol prescribed  by
    this paragraph is the conventional three-
    biaded design:
           RADIATION SYMBOL
  1. Cross-hatched area Is to be  magenta
or purple.
  2. Background Is to be yellow.
                -60°
              FIGURE G-10



       (ii) Deleted

   [43 F.R. 49746, October 24, 1978]
   (2) Rodiotion area. Each radiation area shall
   be conspicuously posted with a sign or signs
   bearing the  radiation  caution symbol
   described in subparapraph (1) of this para-
   graph and the words:

                  CAUTION
              RADIATION AREA
I
  Chanre 38
                                            147
                                  1910.96(e)(2)

-------
                                                       OCCUPATIONAL SArETY AND  HEALTH
 STANDARDS AND INTERPRETATIONS
  (3) High radiation area.

    (i) Each hig-h radiation area shall be con-
    spicuously  posted with  o  sign  or signs
    bearing the radiation caution symbol and
    the words:

                CAUTION
          HIGH RADIATION AREA

    (ii) Each high radiation area shall  be
    equipped with  a control  device which
    shall either cause the level of radiation
    to  be  reduced below that at which  an
    individual might receive a dose of 100 mil-
    lirems in 1 hour upon entry into  the area
    or shall energize a conspicuous visible or
    audible alarm  signal in  such a  manner
    that the individual entering  and the
    employer or a supervisor of the activity
    are made aware of the entry. In  the case
    of a high radiation  area established for
    a period of 30 days  or less, such control
    device is not required.

  (4} Airborne radioactivity  area.

    (!) As used in the provisions of this section,
    "airborne radioactivity area"  means:

     (a) Any room,  enclosure, or operating
     area  in which airborne radioactive
     materials, composed wholly or partly of
     radioactive material, exist in concentra-
     tions  in excess of the amounts specified
     in column 1 of Table 1 of Appendix B
     to 10  CFR Part 20  or

     (b) Any room, enclosure, or operating
     area  in which airborne radioactive
     materials exist in concentrations which,
     averaged over the  number of hours  in
     any week during which individuals are
     in the area, exceed 25 percent of the
     amounts specified in column 1  of Table
     1 of Appendix B to 10 CFR Part 20.

    (H) Each airborne radioactivity area shall
    be  conspicuously posted  with a sign  or
    signs bearing the radiation caution sym-
    bol described in subparagraph (1) of this
    paragraph and the words:

                CAUTION
     AIRBORNE RADIOACTIVITY AREA
      (5) Additional requirements.

        (i) Each area or room in which radioactive
        material is used or stored and which con-
        tains any  radioactive material (other
        than natural uranium or thorium) in any
        amount exceeding 10 times the quantity
        of such material specified in Appendix C
        to 10 CFR Part 20 shall be conspicuously
        posted  with a sign  or signs bearing the
        radiation caution  symbol described  in
        subparagraph (1) of this paragraph and
        the  words:
                    CAUTION
            RADIOACTIVE  MATERIALS

        (ii) Each area or room in which  natural
        uranium or thorium is used or stored in
        an amount exceeding 100 times the quan-
        tity  of such  material specified in 10 CFR
        Part 20 shall be conspicuously posted with
        a sign or signs bearing the radiation cau-
        tion  symbol described in subparagraph (1)
        of this paragraph and the words:

                    CAUTION-
            RADIOACTIVE MATERIALS
      (6) Contoiners.

        (i) Each container in which is transported,
        stored,  or used a quantity of any radioac-
        tive material (othev than natural
        uranium or thorium) greater than the
        quantity of 'such material specified  in
        Appendix C to 10 CFR Part 20 shall bear
        a durable, clearly visible label bearing the
        radiation caution  symbol described  in
        subparagraph (1) of this paragraph and
        the  words:
                    CAUTIO.V
            RADIOACTIVE  MATERIALS

        (ii) Each container in which natural
        uranium or thorium is transported.
        stored,  or used in a quantity greater than
        10  times  the  quantity  specified  in
        Appendix C to 10 CFR Part 20 shall bear
        a durable, clearly visible label bearingthe
        radiation caution symbol  described  in
        subparagraph (1) of this  paragraph and
        the words:

                    CAUTION-
            RADIOACTIVE  MATERIALS
1910.96(O(6)(ii)
148

-------
OCCUPATIONAL SArETY AND  HEALTH
                                                                           l»10.M
-------
                                                        OCCUPATIONAL SATTTY AND HEALTH
STANDARDS  AND INTERPRETATIONS


    shall be low enough to minimize personal
    injuries or excessive property damage
    that might result from such evacuation.

  (3) letting.

    (i) Initial tests, inspections, and checks of
    the signal-generating system shall be
    made to verify  that  the fabrication and
    installation were made in accordance
    with design plans and specifications and
    to develop a thorough knowledge of the
    performance of the system and all compo-
    nents  under  normal  and  hostile condi-
    tions.

    (ii) Once the system  has been placed in
    service, periodic  tests, inspections, and
    checks shall be made to minimize the pos-
    sibility of malfunction.

    (Hi) Following significant  alterations or
    revisions to the system, tests and checks
    similar to the  initial installation tests
    shall be made.

    (iv) Tests  shall  be designed to minimize
    hazards while conducting the tests.

    (v) Prior to normal operation the signal-
    generating system shall be checked phys-
    ically and functionally to assure reliabil-
    ity and to demonstrate accuracy and per-
    formance. Specific tests shall include:

      (a) All power sources.

      (b) Calibration and calibration stability.

      (c) Trip levels and stability.

      (d) Continuity of function with loss and
      return of required  services such as AC
      or DC power, air pressure, etc.

      (e) All indicators.

      (f) Trouble indicator  circuits and
      signals, where used.

      (g) Air pressure (if used).
         graph (l)(ii)  of this paragraph at all
         points that require  immediate evacua-
         tion.

        (vi) In addition to the initial startup and
        operating tests, periodic scheduled  per-
        formance tests and status checks must
        be made to insure that the system is at
        all times operating within design limits
        and capable of the required response.
        Specific periodic tests or checks  or both
        shall include:

         (a)  Adequacy of  signal  activation
         device.

         (b)  All power sources.

         (c)  Function of all alarm circuits and
         trouble indicator circuits including trip
         levels.

         (d)  -Air pressure (if used).

         (e) Function  of entire system including
         operation  without  power  where  re-
         quired.

         (f)  Complete operational tests includ-
         ing sounding of the signal and determi-
         nation that sound levels are adequate.

        (vi!)  Periodic tests shall be scheduled on
        the  basis of need, experience, difficulty,
        and  disruption of operations. The entire
        system should  be operationally tested at
        least quarterly.

        (viii) All employees whose  work may
        necessitate their  presence in an area.
        covered by the  signal shall be made famil-
        iar with the actual sound of the signal—
        preferably as it sounds at their work
        location. Before placing the system into
        operation, all employees normally work-
        ing in the area shall be made acquainted
        with the signal by-actual demonstration
        at their work locations.

   (g) Exceptions from posting requirements.
      (h) Determine that sound level of the     Notwithstanding  the  provisions  of  para-
      signal is  within  the  limit of subpara-   graph (e) of this section:
1910.96(s)
150

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OCCUPATIONAL SArrTT AND KEAI.TE
                                                          STAHDAADS  AND IKTERPJUTTATJOKi
(1)
        loom  01 are:i  is not leouireiJ to be
         with .•; r;mtion  sijrn because  of trie
   prosence of ?. se:iled  source, provided tne
   radiation  level \'l inches from the surface
   oflhe source container or housing rioes no;
   exceed I milhrem per houi.

   (2) Rooms o; olher areas in onsite meriicni
   facilities are not required to be posted \vith
   caution Mtrns because of the presence of
   patients containing  radioactive  material,
   proviuec  tr.ai  there are  personnel in
   attenoance who shall take the precautions
   necessary  to prevent the exposure of any
   inriividunl lo radiation or  radioactive
   material in excess of the limits established
   in the provisions of this section.


  (3) Caution  sijrns are not required  to be
  posted at areas or rooms containing' radioac-
  tive materials for periods  of less than  8
  hours: Proridfff, That

    (!) The materials are constantly attended
    during such periods by  an individual who
    shall  take the precautions necessary to
    prevent the exposure  of" any individual
    to  radiation or radioactive materials in
    excess of the limits established in the pro-
    vision?, of this section;  and

    (ii) Such  are?,  or room is subject to the
    employer's control.

(h) Exemptions for roo'iooctive mderiois
pockoged for shipment.

   Radioactive materials packajred and
labeled in accordance with  regulations of the
Department of Transportation published in -J9
CFR Chapter I, are exempt from the labeling
and posting requirements of this subpart dur-
ing shipment, provided that the inside con-
tainers are labeled in accordance with the pro-
visions of paragraph (e) of this section.

(i) Instruction of personnel, posting.

   (1) Employers regulated by the Atomic
   Energy Commission  shall be governed by
   10 CFR Part 20  standards. Employers in  a
   State named in paragraph (pX3) of this sec-
   tion shall be poverned by the  requirements
   of the  laws and  regulations of that State.
All  other  employers shall be regulated b-
the
                                              (2) All individuals working in or freouentinr.
                                              2 n ^  portion of t. -radiation area shall bt
                                              in formed of the occurrence of  radioactive
                                              materials  or of radiation in such portions
                                              of the radiation area; shall be instructed
                                              in the safer;1  problems associated  with
                                              exposure to such materials or radiation enc
                                              in precautions or devices tc  minimize
                                              exposure;  shall be  instructed in the appli-
                                              cable provisions of this section for the pro-
                                              tection of  employees  from  exposure to ra-
                                              diation or radioactive materials;  and shall
                                              be advised of reports of radiation expusuiv
                                              which employees may request pursuant  !••
                                              the  regulations in  this section.

                                              (3) Each employer to -whom  this  sec.tioi.
                                              applies shall post a current copy of its provi
                                              sions and a copy of the ope rating procedures
                                              applicable to the work conspicuously in such
                                              locations as to insure that employees work-
                                              ing  in or  frequenting radiation areas will
                                              observe these documents on the way to and
                                              from  their place of employment, or shall
                                              keep such documents available for examina-
                                              tion of employees  upon request.

                                            (j) Storage of radioactive materials.

                                              Radioactive materials stored in a nonrsulifc-
                                            tion  area shall  be- secured  against  unau-
                                            thorised removal from the place  of storage.
                                            (k) Waste disposal.

                                              No  employer, shall dispose of radioactive
                                            material except, by transfer to an authorized
                                            recipient, or in 2 manner  approved by  the
                                            Atomic Energr Commission or z State named
                                            in paragraph (p)(3) of this section.

                                            (I) Notification of incidents.

                                              (1) Immediate notification. Each employer shall
                                              immediately notify the Assistant Secretary
                                              of  Labor or his duly authorized  represent-
                                              ative, for employees not protected by the
                                              Atomic Energy Commission by means of 10
                                              C"R  Part 20;  parajrraph (oX2) of  this sec-
                                              tiun,  or  the  requirements  of the  laws and
                                                                                 191C.96(l)tl)

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                                                       OCCUPATIONAL SAFETY AND HEALTH
STANDARDS AND  INTERPRETATIONS
  regulations of Slates named in paragraph
  (p)(3) of this section, by telephone  or tele-
  graph of any incident  involving radiation
  which  mR\_have caused or threatens to
  cause:

    (i)  Exposure of the  whole  body of any
    individual to 25 rems or more of radiation;
    exposure of the skin of the whole booy
    of any individual to 150 rems or  more of
    radiation; or exposure of the feet, ankles,
    hands, or forearms of any  individual to
    376 rems or more of radiation; or

    (H) The release of radioactive material in
    concentrations which, if averaged over a
    period of 24 hours, would  exceed 5,000
    times the limit specified for such  materi-
    als in Table 11 of Appendix B to  10 CFR
    Part 20.
  (2) Twenty-four hour notification. Each employer
  shall within 24 hours following its occur-
  rence notify the Assistant Secretary of
  Labor or his duly authorized representative
  for employees not protected by the Atomic
  Energy Commission  by  meim* of 10 CFR
  Part 20; paragraph (p){2) of this section, or
  the requirements of the laws and applicable
  regulations of States named in paragraph
  (p)(3) of this section,  by  telephone  or tele-
  graph of any incident involving radiation
  which  may have caused or threatens to
  cause:

    (i)  Exposure  of the whole  body of any
    individual to 5 rems or more of radiation;
    exposure of the skin of the whole body
    of any individual to 30 rems or more of
    radiation; or exposure  of the feet, ankles,
    hands, or forearms to 75 rems  or more
    of radiation; or
    (m) Reports  of overexposure  and excessive
    levels and concentrations.

     (1) In addition to any notification required
     by  paragraph  (1) of this  section each
     employer shall  make a report in  writing
     within 30 days to the Assistant Secretary
     of Labor or his duly authorized represent-
     ative,  for employees  not protected by the
     Atomic Energy Commission  by means of 10
     CFR Part  20; or under  paragraph (p}(2) of
     this section, or the requirements of the laws
     and regulations of States named  in parn-
     ' graph (p)(3)-of this section, of each exposure
     of an individual to radiation or concentra-
     tions of radioactive material  in excess of any
     applicable  limit in this section. Each report
     required  under  this subparagraph shall
     describe the extent of exposure of persons
     to radiation or  to radioactive material;
     levels  of radiation  and concentration of
     radioactive material involved, the  cause ol
     the exposure, levels of concentrations; and
     corrective  steps taken or planned to assure
     against a recurrence.

     (2)  In  -any  case where an  employer  is
     required pursuant to the provisions of this
     paragraph to report to the U.S. Department
     of Labor any exposure  of an individual to
     radiation or to concentrations of radioactive
     material,  the  employer shall also notify
     such individual of the nature and extent of
     exposure.  Such notice  shall be in  writing
     and shall contain the following statement:
     "You should preserve this report for future
     reference."
    (n) Records.


      (1) Every employer shall maintain records
      of the radiation exposure of all employees
      for whom personnel monitoring is required
      under paragraph (d) of this section and
      advise each  of  his employees of  his
      individual exposure on at least an annual
      basis.
                                               (2) Every employer shall maintain records
                                               in the  same units used in tables in para-
                                               graph (b) of this section and Appendix B
                                               to 10 CFR Part 20.  •
1910.96
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OCCUPATIONAL SAFTTT AND HEALTH
 (o) Disclosure to former employee of individ-
 ual employee's record.

   (]J At the reouest of a former employee an
   employer shall furnish to the employee a
   report of the employee's exposure to radia-
   tion as shown in records maintained by the
   employer pursuant to  paragraph (n)(l)  of
   this section. Such report shall be furnished
   \vithin 30 days from the time the request
   isniade, and shall cover each calendar quar-
   ter of the individual's, employment involv-
   ing exposure  to radiation or such lesser
   period us inny be ri-ijiiesteri by the employee.
   The report shall also include the results of
   any calculations and analysis of radioactive
   material deposited in  the body of the
   employee. The re port shall be in writing-and
   contain the  following statement:  "You
   should  preserve  this report for future
   reference."
(p)  Atomic  Energy   Commission   licen-
sees—AEC contractors operating AEC plants
and  facilities—AEC  agreement   State
licensees or registrants.

   (1) Any employer who possesses or uses
   source material, byproduct material, or spe-
   cial nuclear material, as defined  in the
   Atomic Energy Act of 1954, as  amended,
   under a license issued by the Atomic Energy
   Commission and in accordance with the
   requirements of 10  CFR  Part  20 shall be
   deemed  to be in  compliance  with the
   requirements of this section with respect
   lo such possession and use.

   (2) AEC contractors operating  AEC plants
   and facilities; Any employer who possesses
   or  uses  source  material,, byproduct
   material, special nuclear material, or other
   radiation sources under a contract with the
   Atomic Energy Commission  for the opera-
   tion of AEC plants and facilities  and  in
   accordance with the standards, procedures,
   and other requirements for  radiation pro-
   tection established  by the Commission for
   such  contract pursuant to  the Atomic
   Energy Act of 195-i  as amended (42 U.S.C.
   2011 et seq.), shall be deemed to  be in com-
              STANDAJIDS A.KD INTERPRETATIONS


     pliance with the requirements of this sec-
     tion with respect to such possession and use.

     (3) AEC-agreement State licensees or regis-
     trants:

       (i) Atomic Energy Act tourcei. Any employer
       who  possesses or uses source material,
       byproduct material, or special nuclear
       material, as defined in the Atomic Energy
       Act of 1954, as amended (42 U.S.C. 2011
       et seq.), and  has either registered such
       sources with, or is  operating under  a
       license issued by, a State which has an
       agreement in effect with the Atomic
       Energy Commission pursuant to section
       274(b) (42 U:S.C. 202l(b)) of the Atomic
       Energy Act of 1954, as amended,  and in
       accordance with the requirements of that
       State's laws and regulations shall be
       deemed to be in compliance with  the
       radiation  requirements of this section,
       insofar as his possession and use of such
       material  is concerned, unless the Secre-
       tary of Labor, after conference with the
       Atomic Energy'Commission, shall deter-
       mine that the State's program for control
       of these radiation sources is incompatible
       with the requirements of this section.
       Such agreements currently are in effect
       only in the States of Alabama, Arkansas,
       California, Kansas,  Kentucky, Florida,
       Mississippi, New Hampshire, New York,
       North  Carolina, Texas,  Tennessee,
       Oregon,  Idaho, Arizona,  Colorado,
       Louisiana, Nebraska, Washington, Mary'
       land, North Dakota, South Carolina, and
       Georgia.

       (ii) Other source^. Any employer who pos-
       sesses or uses radiation sources other
       than  source   material,  byproduct
       material, or special nuclear material, as
       defined in the Atomic Energy Act of 1954,
       as amended (42 U.S.C. 201l"et seq.), and
       has either registered such sources with,
       or is operating under a license issued by
       a State which has an agreement in effect
       with the Atomic Energy Commission pur-
       suant to section.274(b) (42 U.S.C. 2021(b))
       of the Atomic Energy Act of 1954, as
       amended, and in accordance with  the
       requirements of that State's laws and
       regulations shall be deemed to be in com-
       pliance with  the radiation requirements
       of this section, insofar as his possession
 Chanft
153
                                                                             1910.96
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 10.»6(pXJXll)
      OCCUPATIONAL SAFTrY AND HEALTH
STANDARDS AND  INTERPRETATIONS
    and use of such material  is concerned,
    provided the State's program for control
    of these radiation sources  is the subject
    of a currently effective determination by
    the Assistant Secretary of Labor that
    such program  is compatible with  the
    requirements of this section. Such  deter-
    minations currently are in effect only in
    the  States  of Alabama, Arkansas,
    California, Kansas, Kentucky,  Florida,
    Mississippi, New Hampshire, N'ew York,
    North  Carolina,  Texas, Tennessee,
    Oregon,   Idaho, Arizona, Colorado,
    Louisiana, Nebraska, Washington, Mary-
    land, North Dakota, South Carolina, and
    Georgia.

(q) [Reserved]

(r) Radiation standards for mining. Revoked
                 1910.97—NONIONIZ1NG  RADIATION
 (a) Electromagnetic radiation.

   (1) Definitions applicable lo this paragraph.

     (i) The term "electromagnetic radiation"
     is restricted to that portion of the spec-
     trum commonly defined as the radio fre-
ouency region, which for the purpose of
this specification shall include the micro-
wave frequency region.

(ii) Partial body irradiation. Pertains to the
case in which part of the body is exposed
to the incident electromagnetic energy.

                               Ch&nre 7

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The following document entitled Federal Radiation Protection



Guidance for Occupational Exposure, has been retyped from the



Federal Register dated Tuesday, January 27, 1987 (Vol. 52, No. 17).

-------
FEDERAL RADIATION PROTECTION GUIDANCE FOR OCCUPATIONAL EXPOSURE

This memorandum transmits recommendations that would update previous guidance to Federal
agencies for the protection of workers exposed to ionizing radiation.  These recommendations
were developed cooperatively by the Nuclear Regulatory  Commission, the Occupational Safety
and Health Administration,  the Mine Safety and Health Administration,  the Department of
Defense,  the Department of Energy, the National Aeronautics and Space Administration,  the
Department of Commerce,  the Department of Transportation, the Department of Health and
Human Services, and the Environmental Protection Agency.  In addition, the National Council
on Radiation Protection and Measurements (NCRP), the National Academy of Sciences (NAS),
the Conference of Radiation  Control Program Directors (CRCPD) of the States, and the Health
Physics Society were consulted during the development of this guidance.

Executive Order  10831, the  Atomic Energy Act, as amended, and Reorganization Plan No. 3
of 1970 charge the Administrator of the Environmental Protection Agency (EPA) to ". . advise
the President with respect to radiation matters, directly or indirectly affecting health, including
guidance  for  all Federal  agencies in the  formulation  of radiation  standards and  in  the
establishment  and execution of programs of cooperation with  States."   This  guidance  has
historically taken the  form  of  qualitative  and  quantitative  "Federal  Radiation Protection
Guidance." The recommendations transmitted here would replace those portions of previous
Federal guidance (25 FR 4402), approved by President Eisenhower on May 13, 1960, that apply
to the protection of workers exposed to ionizing radiation. The portions of that guidance which
apply to exposure of the general public would not be changed by these recommendations.

These recommendations are based  on consideration of (1) current scientific understanding of
effects on health from ionizing radiation, (2) recommendations of international  and national
organizations  involved in  radiation protection,  (3)  proposed "Federal Radiation Protection
Guidance for Occupational Exposure" published on January 23, 1981  (46 FR 7836) and public
comments on that proposed  guidance, and (4) the collective experience of the Federal agencies
in the control of occupational exposure to ionizing radiation.  A summary of the considerations
that led  to these recommendations is provided below.  Public  comments  on  the previously
proposed  guidance and a response to those comments are contained in the document "Federal
Radiation Protection Guidance for Occupational Exposure-Response to Comments" (EPA 520/1-
84-011).  Single copies of this report are available from the Program Management Office (ANR-
458), Office of Radiation Programs, U.S. Environmental Protection Agency, Washington, D.C.
20460; telephone (202) 475-8388.

Background

A review of current radiation protection guidance for workers began in 1974 with the formation
of a Federal interagency committee by EPA.  As a result of the deliberations of that committee,
EPA published an "Advance Notice of Proposed Recommendations and Future Public Hearings"
on September 17, 1979 (44 FR 53785).  On January 23, 1981, EPA published  "Federal
Radiation Protection Guidance for Occupational Exposures; Proposed Recommendations, Request

-------
for Written Comments, and Public Hearings"  (46 FR 7836).  Public hearings were held in       •
Washington, D.C. (April 20-23, 1981); Houston, Texas (May 1-2, 1981); Chicago, Illinois (May
5-6, 1981), and San Francisco, California (May 8-9, 1981) (46 FR 15205).  The public comment
period closed July 6,  1981 (46 FR 26557).  On December 15,  1982, representatives of the ten
Federal agencies noted above, the CRCPD  and NCRP convened under the sponsorship of the
EPA to  review  the issues raised in public comments and to  complete development of  these
recommendations.  The issues were carefully considered during a series of meetings, and the
conclusions of the working group have provided the basis for these recommendations for revised
Federal guidance.

EPA has  also  sponsored or conducted four major studies  in support of this  review  of
occupational radiation protection guidance.  First, the Committee on the Biological Effects of
Ionizing Radiations, National Academy of Sciences-National  Research Council reviewed the
scientific data on health risks of low levels of ionizing radiation in a report transmitted to EPA
on July 22, 1980:   "The Effects on Populations of Exposure  to Low  Levels  of Ionizing
Radiation:  1980," National Academy  Press,  Washington, D.C. 1980.  Second,  EPA has
published  two studies of occupational  radiation exposure:  "Occupational Exposure to Ionizing
Radiation  in the United States: A Comprehensive Summary for the Year 1975" (EPA 520/4-80-
001) and "Occupational Exposure to Ionizing Radiation in the United States:  A Comprehensive
Review  for the Year  1980 and Summary of Trends for the Years 1960-1985" (EPA  520/1-84-
005).  Third, the Agency sponsored  a study  to  examine the changes in previously derived
concentration limits for intake of radionuclides  from  air or water that result from use of
up-to-date dosimetric and biological transport models.  These are presented in Federal Guidance
Report No. 10, "The Radioactivity Concentration Guides: A New Calculation of Derived Limits
for the  1960 Radiation Protection Guides  Reflecting Updated  Models for Dosimetry and
Biological Transport" (EPA 520/1-84-010).  Finally,  the cost of implementing the changes in
Federal  guidance proposed on January 23,  1981 was surveyed and the findings published in the
two-volume report:   "Analysis  of Costs  for Compliance with Federal Radiation Protection
Guidance  for Occupational Exposure:  Volume I-Cost of Compliance"  (EPA 520/1-83-013-1)
and "Volume II-Case Study Analysis of the Impacts" (EPA 520/1-83-013-2). These EPA reports
are available from National Technical Information Service, U.S. Department of Commerce, 5285
Port Royal Road, Springfield, Virginia 22161.

The interagency review of occupational radiation protection has confirmed the need for revising
the previous Federal guidance, which  was promulgated in 1960.  Since that time knowledge of
the effects of ionizing radiation on humans has  increased substantially. We now have a greatly
improved  ability to estimate risk of harm due to irradiation of individual organs and tissues.  As
a result, some of the old numerical guides are now believed to be less and some more protective
than formerly.  Other risks,  specifically those to the unborn,  are now  considered to be more
significant and were not addressed by the old guidance.  These disparities and omissions should
be corrected.   Drawing on this improved knowledge,  the International  Commission  on
Radiological Protection (ICRP) published, in 1977, new recommendations on radiation protection
philosophy and  limits for occupational exposure.  These recommendations are now in use, in
whole or substantial pan, in most other countries.  We have considered these recommendations,

-------
among others, and  believe that it is  appropriate to adopt the  general  features of the ICRP
approach in radiation protection guidance to Federal agencies for occupational exposure. These
recommendations are now in use, in whole or substantial pan, in most other countries.  We have
considered these recommendations, among others, and believe that it is appropriate to adopt the
general features of the ICRP approach in radiation protection guidance to Federal agencies for
occupational exposure. In two cases, protection of the unborn and the management of long-term
exposure to internally deposited radioactivity, we have found it advisable to make additions.

There are four types of possible effects on health  from exposure to ionizing radiation.  The first
of these is cancer.  Cancers  caused by  radiation are not different from  those that have been
historically observed,  whether from known or unknown causes.  Although radiogenic cancers
have been observed in humans over a range of higher doses,  few useful data are available for
defining the effect of doses at normal occupational levels of exposure. The second type of effect
is  the induction of hereditary effects in descendants of exposed persons.   The severity  of
hereditary effects ranges from inconsequential  to  fatal.   Although  such  effects have been
observed in  experimental animals  at high  doses, they have not  been  confirmed in  studies of
humans.  Based on extensive but incomplete scientific evidence, it is prudent to assume that at
low levels of exposure the risk of incurring either cancer or hereditary effects is linearly related
to the dose received in the relevant tissue.  The severity  of any such effect is not related to the
amount of dose received.  That is, once a cancer or an hereditary effect has been induced, its
severity is independent of the doses.  Thus, for  these two types of effects, it is assumed that
there is no completely risk-free level of exposure.

The third type includes a variety  of  effects for which  the degree of damage (i.e., severity)
appears to depend on the amount of dose received and for which there is an  effective threshold
below which clinically  observable effects  do not occur.   An  example  of  such an effect is
radiation sickness syndrome,  which is observed  at high  doses and is fatal at very high doses.
Examples of lesser effects include opacification of the lens of the eye, erythema of the skin, and
temporary impairment of fertility.  All of these  effects occur at relatively high doses.  At the
levels  of  dose  contemplated  under  both  the  previous  Federal  guidance  and  these
recommendations, clinically observable examples of this third type of effect are not known to
occur.

The fourth type includes effects on children who were  exposed in utero.  Not  only may the
unborn be more sensitive than adults to the induction of malformations, cancer, and  hereditary
effects, but recent studies have drawn  renewed attention to the risk of severe mental retardation
from exposure of the unborn during certain periods of pregnancy.  The risk of less severe mental
retardation  appears  to be similarly elevated. Although it is not yet  clear  to  what  extent the
frequency of retardation  is proportional to the amount of dose (the data  available at occupational
levels of exposure are limited), it is prudent to assume that proportionality exists.

The risks to health from exposure to low levels of ionizing radiation were reviewed for EPA by
the NAS in  reports published in  1972 and in 1980. Regarding cancer there continues to be
divided opinion on how  to interpolate  between the absence of radiation effects at zero dose and

-------
the observed effects of radiation (mostly at high doses) to estimate the most probable effects of
low doses.  Some scientists believe that available data best support use  of a linear model for
estimating such effects.  Others,  however, believe that other models, which  usually predict
somewhat lower risks, provide better estimates. These differences of opinion have not been
resolved to date by studies of the effects or radiation in humans, the most important of which
are those of the Hiroshima and Nagasaki atom bomb survivors.  Studies are now underway to
reassess radiation dose calculations for these survivors and in turn to provide improved estimates
of risk. It will be at least several years before these reassessments and estimates are completed,
and it is not likely that they will conclusively resolve uncertainties in estimating low dose effects.
EPA is monitoring the progress of this work.  When it is completed we will initiate reviews of
the risks of low levels of radiation, in order to provide the basis for any indicated reassessment
of this guidance.

In spite of the above uncertainties, estimates of the risks from exposure to low levels of ionizing
radiation are reasonably well bounded, and the average worker is believed to incur a relatively
small risk of harm from radiation.  This situation has resulted from a system of protection which
combines limits on maximum dose with active application of measures to minimize doses within
these limits.   These  recommendations  continue that approach.   Approximately 1.3 million
workers were employed in occupations in which they were potentially exposed to radiation in
1980,  the latest year for which  we have comprehensive assessments.   About half of these
workers received  no measurable occupational dose. In that year the average worker measurably
exposed to external radiation received an occupational dose equivalent of 0.2 rem to the whole
body,  based on the readings of individual dosimeters worn on the surface of the body.   We
estimate (assuming a linear non-threshold model) the increased risk of premature death due to
radiation-induced cancer for such a  dose is approximately  2 to  5  in  100,000 and that the
increased risk of serious hereditary effects is somewhat smaller.   To put these estimated  risks
in perspective with other occupational hazards, they are comparable to the observed risk of
job-related accidental death in the safest industries, wholesale and retail trades, for which the
annual  accidental  death rate averaged about 5 per  100,000 from  1980 to  1984.  The  U.S.
average for all industries was 11 per 100,000 in 1984 and 1985.

These recommendations  are based on the assumption that  risks of injury from exposure to
radiation should  be considered in relation to the overall benefit derived from  the activities
causing the exposure.  This approach is similar to that used by the Federal Radiation Council
(FRC) in developing the 1960 Federal guidance.  The FRC said  then, "Fundamentally, setting
basic radiation protection standards  involves passing judgment  on the extent of the possible
health hazard  society is willing to accept in order to realize the  known benefits of radiation."
This leads to three basic principles that have governed radiation protection of workers in recent
decades in the United States and in most other countries.  Although the precise formulation of
these principles has evolved over the years, their intent has continued unchanged. The first is
that any activity involving occupational  exposure should be determined to be useful enough to
society to warrant the exposure of workers; i.e., that a  finding be made that the  activity is
"justified".  This same principle applies to virtually any human endeavor which involves  some
risk of injury. The second is that, for justified activities, exposure of the work force should be

-------
as low as reasonably achievable (commonly designated by the acronym "ALARA");  this has
most recently been characterized as "optimization" of radiation protection  by the International
Commission on Radiological Protection (ICRP). Finally,  to provide an upper limit on risk to
individual workers, "limitation" of the maximum allowed  individual dose is  required  This is
required  above and beyond  the protection provided  by the  first two principles because  tneir
primary objective is to minimize the total harm from occupational exposure in the entire work
force; they do not limit the way that harm is distributed among individual  workers.

The principle  that activities causing  occupational  exposure should produce a net benefit is
important in radiation protection even though the judgement of net benefit is not easily made.
The 1960 guidance says:   "There should not be any man-made radiation exposure without the
expectation of benefit resulting from  such exposure  ..."  And "It is basic that exposure to
radiation should result from a real determination of its necessity."  Advisor)' bodies other than
the FRC have used language which has essentially the same meaning. In its  most recent  revision
of international guidance (1977) the ICRP  said ".  .  .no  practice  shall be adopted unless its
introduction produces a positive net benefit," and  in slightly different  form  the NCRP, in its
most recent statement  (1975)  on this matter, said "... all exposures should  be kept to a
practicable minimum; . .  . this principle involves value judgments based  upon perception of
compensatory benefits commensurate  with risks, preferably  in the  form of realistic numerical
estimates of both  benefits and risks from activities involving radiation and alternative means to
the same benefits."

This principle is set forth in these recommendations in a simple form:  "There should not be any
occupational exposure  of workers to  ionizing radiation without  the expectation of an overall
benefit from the activity causing the exposure."  An obvious difficulty in making this judgement
is the difficulty of quantifying in comparable terms cost (including risks) and benefits. Given
this situation, informed value judgements are necessary and  are usually all that is possible. It
is perhaps useful  to observe, however,  that throughout history  individuals and societies  have
made risk-benefit judgements, with their success usually depending upon the amount of  accurate
information available.  Since more is known about  radiation now than in previous decades, the
prospect is that these judgments can now be better made than before.

The preceding discussion has implicitly focused on major activities, i.e.,  those instituting or
continuing a general practice  involving radiation  exposure  of workers.   This principle also
applies to detailed management of facilities  and direct supervision of workers.  Decisions on
whether or not particular  tasks should be carried  out  (such as inspecting control systems or
acquiring  specific experimental data)  require judgments which  can, in the aggregate,  be as
significant for radiation protection as those justifying the basic activities these tasks support.

The principle of reduction of exposure to  levels  that are "as low as  reasonably achievable"
(ALARA) is typically implemented in  two different ways.  First, it is applied  to the engineering
design of facilities so as to reduce, prospectively, the anticipated exposure of workers.  Second,
it is applied  to actual operations; that is, work practices are designed and carried out to reduce

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the  exposure  of  workers.    Both  of  these  applications  are  encompassed  by  these
recommendations." The principle applies both to collective exposures of the work force and to
annual and cumulative individual exposures.  Its application  may therefore require complex
judgments, particularly when tradeoffs between collective  and individual doses are involved.
Effective implementation of the ALARA  principle involves most  of  the many facets of an
effective radiation protection  program:  education of workers  concerning the  health risks of
exposure to radiation; training in regulatory requirements and  procedures to control exposure;
monitoring, assessment.,  and  reporting of exposure levels and doses; and management  and
supervision  of radiation protection activities,  including the  choice  and implementation of
radiation control measures.  A comprehensive radiation  protection program will also include,
as  appropriate,  properly  trained  and qualified  radiation  protection   personnel; adequately
designed,  operated, and maintained facilities and equipment; and quality assurance and audit
procedures. Another important aspect of such programs is maintenance of records of cumulative
exposures  of  workers and implementation of  appropriate measures  to assure that lifetime
exposure of workers repeatedly exposed  near the limits is minimized.

The types of work and activity which involve worker exposure to radiation vary greatly and are
administered  by many different Federal and State agencies under a wide variety of legislative
authorities.  In view of this complexity, Federal radiation protection guidance can address only
the broad prerequisites of an effective ALARA program,  and regulatory authorities must ensure
that more detailed requirements are identified and carried out.  In doing this, such authorities
may find it useful to establish or encourage  the use of 1) administrative control levels specifying,
for specific categories of workers or work situations, dose levels below the limiting numerical
values recommended in this guidance; 2) reference levels to indicate the need  for such actions
as  recording, investigation, and intervention; and 3)  local goals for  limiting individual and
collective occupational exposures.  Where the enforcement of a general ALARA requirement
is not practical under  an agency's statutory authority, it is sufficient that an agency endorse and
encourage  ALARA, and establish such regulations which result from ALARA findings as  may
be useful and appropriate to meet the  objectives of this guidance.

The numerical radiation protection guidance which has been in  effect since 1960 for limiting the
maximum allowed dose to an individual worker is based on the concept of limiting the dose to
the  most  critically exposed part of the body.   This  approach was  appropriate, given  the
limitations of scientific information  available  at the time,  and resulted in  a  set of  five
independent numerical guides for maximum exposure of a) the whole body,  head and trunk,
    *  The recommendation that Federal agencies,  through their regulations, operational
 procedures and other appropriate means, maintain doses ALARA is not intended to express, and
 therefore should not be interpreted as expressing, a view whether the ALARA concept should
 constitute a duty  of care in  tort litigation.  Implementation of the ALARA concept requires a
 complex, subjective balancing of scientific, economic and social factors generally resulting in
 the attainment of average dose  levels significantly  below  the  maximum permitted  by this
 guidance.

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active blood-forming organs, gonads, and lens of eye; b) thyroid and skin of the whole body;
c) hands and  forearms, feet and ankles; d) bone,  and e) other organs.  A consequence of this
approach when several different parts of the body are exposed simultaneously is that only  the
part that receives the highest dose relative to  its  respective  guide is decisive for limiting  the
dose.

Current knowledge permits a more comprehensive approach that takes into account the separate
contributions  to the total risk from each exposed part  of the body.  These  recommendations
incorporate the dose weighting system introduced  for this prrpose by the ICRP in 1977.  That
system assigns weighting factors to the various pans of the body for the risks of lethal cancer
and serious prompt genetic effects (those in the first two generations); these factors are chose
so that  the sum  of  weighted  dose equivalents represents  a  risk the  same as that  from a
numerically equal dose equivalent to the whole body. The ICRP recommends that the effective
(i.e. weighted) dose equivalent incurred in any year be limited to 5 rems. Based on the public
response to  the  similar proposal  published by EPA in  1981 and Federal experience with
comparable exposure limits,  the Federal agencies concur.  These recommendations therefore
replace  the  1960 whole body numerical guides of 3 rems per quarter and 5(N-18)  rems
cumulative dose equivalent (where N is the age of the worker) and associated critical organ
guides with  a  limiting  value of 5 rems effective dose equivalent incurred  in any  year.
Supplementary limiting values are  also recommended to provide protection against those health
effects for which an effective threshold is believed to exist.

In recommending a limiting value of 5 rems  in any single year, EPA has had to balance a
number of considerations.  Public comments  confirmed that, for some beneficial activities,
occasional doses approaching this value are not  reasonably avoidable.  On the other hand,
continued annual exposures at or near this level over substantial portions of working  lifetime
would, we believe, lead to unwarranted risks. For this reason such continued annual exposures
should be avoided, and these recommendations provide  such  guidance.  As noted  earlier,  these
recommendations also continue a system of protection  which combines  limiting values  for
maximum dose with a requirement for active application of measures to  minimize doses-the
ALARA requirement. This has resulted in steadily decreasing average annual doses to workers
(most recently to about one-fiftieth of the recommended limiting value) and, to date, only a few
hundred out of millions of workers have received planned cumulative doses that are a substantial
fraction of the maximum previously permitted  cumulative dose over and occupational lifetime.
EPA anticipates that the continued application of the ALARA requirement, combined with new
guidance on avoidance of large cumulative doses, will result in maintaining risks to all workers
at low  levels.   EPA will  continue to review  worker doses with  a view  to initiating
recommendations for any further modifications of the dose limitation system that are warranted
by future trends in worker exposure.

Certain radionuclides, if inhaled or ingested, may remain in  and continue to irradiate the body
for many years.  These recommendations provide that radionuclides should be contained so as
to minimize intake, to the extent reasonably achievable.  When avoidance of situations that may
result in such  intake is not practical, the recommendations distinguish between pre-exposure and

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post-exposure situations.  With respect to the former, Federal agencies should base control of
prospective internal exposure to radionuclides (e.g. facility design, monitoring, training, and
operating procedures) upon the entire future dose that may result from any intake (the committed
dose), not just  upon the  dose accrued  in the year of intake.  This is to assure that,  prior to
exposure to such materials, proper account is taken of the risk due to doses in future years.

With respect to post-exposure  situations, most significant internal  exposure to radionuclides
occurs as the result of inadvertent intakes. In the case of some long-lived radionuclides, it may
also be difficult to measure accurately  the small quantities corresponding to  the recommended
numerical  guidance  for  control  of committed  doses.   In  such  cases,  when workers  are
inadvertently exposed  or it  is not otherwise possible  to  avoid intakes  in excess of these
recommendations  for control of committed dose, it will  be necessary  to  take  appropriate
corrective action to assure control has been reestablished and to properly manage future exposure
of the worker.  In regard to the latter requirement,  provision should be made to  continue to
monitor the annual dose received from radionuclides in the  body as long as they remain in
sufficient amount to deliver doses significant compared to the limiting values for annual dose.
These recommendations extend those of the  ICRP, because it is appropriate  to maintain active
management  of workers who exceed the guidance for committed dose in order that individual
differences in retention of such materials in the body be monitored, and to  assure, whenever
possible, conformance to the limiting values for annual dose.

These recommendations also incorporate guidance  for limiting exposure of the unborn as a result
of occupational exposure of female workers.  It has long been suspected that the embryo and
fetus are  more sensitive to a variety  of effects  of radiation than are adults.   Although our
knowledge remains incomplete, it has now become clear that  the unborn are especially subject
to the risk of mental retardation from exposure to radiation at a  relatively early phase of total
development.  Available scientific evidence appears  to indicate that this sensitivity is greatest
during the period  near the end of the first trimester and the beginning of the second trimester
of  pregnancy,  that is,  the  period from  8 weeks to about  15 weeks  after  conception.
Accordingly, when a woman has declared her pregnancy, this guidance recommends not only
that the total exposure of the unborn be more limited than that of adult workers,  but that the
monthly rate of exposure be  further limited in order to  provide additional protection.  Due to
the incomplete  state of knowledge of the transfer of radionuclides from the mother to the unbom
(and the resulting  uncertainty in dose to the unborn), in those  few work situations where intake
of radionuclides could normally be possible it may also be necessary to institute measures to
avoid such intakes by pregnant women in order to satisfy these recommendations.

The  health protection objectives of this  guidance  for the unborn  should be achieved in
accordance with the provisions of Title VTI  of the Civil  Rights Act of 1964, as amended, with
respect to discrimination in employment practices.** The guidance applies only to situations
     **       The Civil Rights Act of 1964, as amended provides that "It shall be an unlawful
 employment practice for an employer (1) to fail or refuse to hire or to discharge any individual,

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in which the worker has voluntarily made her pregnancy known to her employer.  Protection of
the unborn may be achieved through such measures as  temporary job rotation, worker self-
selection, or use of protective equipment.  The guidance recognizes that protection of the unborn
is a joint responsibility of the employer and worker. Workers should be informed of the risks
involved and encouraged to voluntarily make pregnancies  known as early as possible so that any
temporary arrangements necessary to modify exposures can be made. Conversely, employers
should make such arrangements in a manner that minimizes the impact  on the worker.

The recommenced numerical guidance for limiting dose to workers applies to the sum of dose
from external and internal sources of radiation. This procedure is recommended so as to provide
a single limit on the total risk from radiation exposure.  Therefore, in those cases where both
kinds of radiation sources are present, decisions about to control of dose from internal sources
should not be made without equal consideration of their implication for dose from  external
sources.

The guidance  emphasizes  the importance of recordkeeping for  annual,  committed, and
cumulative (lifetime)  doses.  Such  recordkeeping  should  be designed to avoid burdensome
requirements for cases in which doses are insignificant.  Currently, regulatory records are not
generally required for  doses small compared to regulatory limits for annual external and internal
doses. Under this guidance such regulatory practices would continue to be appropriate if due
consideration is given to  the implications of summing internal  and  external doses and  to
recordkeeping needs for assessing cumulative doses.  To  the extent reasonable such records
should be established on the basis of individual dosimetry  rather than on monitoring of exposure
condition.

In summary, many of the important changes from  the  1960 guidance are structural.   These
include introduction of the concept of risk-based weighting of doses to different parts of the body
and the use of committed dose as the primary basis  for control  of internal exposure.  The
numerical values of the guidance for  maximum radiation doses  are also modified.  These
changes bring this guidance into general conformance with international recommendations and
practice. In addition, guidance is provided for protection of the unborn, and increased  emphasis
is placed  on eliminating unjustified  exposure  and on keeping justified  exposure as low  as
reasonable achievable, both long-standing tenets of radiation  protection.    The  guidance
emphasizes the importance of instruction of workers and their supervisors, monitoring and
or otherwise to discriminate against any  individual with respect to his compensation, terms,
conditions or privileges of employment, because of such individual's sex ... or (2) to limit,
segregate, or classify his employees  or applicants for employment in  any way  which would
deprive or tend to deprive any individual of employment opportunities  or otherwise adversely
affect his status as  an employee, because  of such individual's ...  sex  ..." [42 U.S.C.
2000e(a)].  The Pregnancy Discrimination  Act of 1978 defines "because  of sex" to include
because of or on the basis of pregnancy,  childbirth,  or related medical conditions [42 U.S.C.
2000 e(k)].

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recording of doses to workers,  and the use of administrative control and reference levels for       m
carrying  out ALARA  programs.

These recommendations apply to workers exposed to other than  normal background radiation
on the job. It is sometimes hard to identify such workers because everyone is exposed to natural
sources of radiation and many occupational exposures are small. Workers or workplaces subject
to this guidance will be identified by the responsible implementing  agencies.  Agencies will have
to use care in determining when exposure of workers does not need to be regulated.  In making
such determinations agencies should consider both the collective dose  which is likely to be
avoided through regulation and  the maximum individual doses possible.

Implementation of these recommendations will require changes that can reasonably be achieved
only over a period of time.  It is expected that Federal agencies will identify any problem areas
and provide adequate flexibility and the necessary transition periods to avoid  undue impacts,
while at  the same time assuring reasonable prompt implementation of this new guidance.

Upon implementing these recommendations, occupational exposure should be reduced. It is not
possible  to quantify the overall exposure reduction that will be realized because it cannot be
predicted how efficiently these recommendations will be implemented or how much of existing
exposure in unnecessary. These recommendations reduce  the maximum whole body dose that
works  may receive in any one year by more than half (i.e., from 3 rems per quarter to 5 rems
per year), require that necessary exposure to internal radioactivity be controlled on the basis of
committed dose,  require that internal and external doses  be considered together rather  than
separately, and provide increased protection of the unborn.  We also expect the strengthened and
more explicit  recommendations for maintaining occupational exposure "as  low as reasonably
achievable" will improve the radiation protection of workers.  Finally, these recommendations
would  facilitate the practice of radiation  protection by introducing a self-consistent system of
limits in accordance with that in practice internationally.

Recommendations

The following recommendations are made for the guidance of Federal agencies  in their conduct
of programs for the protection of workers from ionizing radiation.

1.     There should not be any occupational exposure of workers to ionizing radiation without
       the expectation of an overall benefit from the activity causing the exposure.  Such
       activities may  be allowed provided exposure of workers  is limited in  accordance  with
       these recommendations.

2.     No exposure is acceptable without regard to the reason for permitting it, and it should
       be general practice to maintain doses from radiation to levels below the limiting values
       specified in these recommendations. Therefore, it is fundamental to radiation protection
       that a  sustained effort be  made  to ensure that collective doses,  as   well as annual,
                                           10

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       committed, and cumulative lifetime individual doses, are maintained as low as reasonably
       achievable (ALARA), economic and social factors being taken into account.

3.      In  addition to the above recommendations, radiation doses  received  as a result  of
       occupational exposure should  not  exceed  the limiting  values  for  assessed dose  to
       individual workers specified below.  These  are given separately for protection against
       different  types of effect  on health  and  apply  to  the sum of doses from external and
       internal  sources  of radiation.  For cancer and  genetic effect, the limiting value is
       specified  in terms of a derived quantity  called  the effective dose equivalent.  For other
       health effects, the limiting  values  are  specified  in terms of the dose  equivalent1  to
       specific organs or tissues.

Cancer and Genetic Effects.  The effective dose  equivalent, HE, received in any year by an adult
worker should not exceed 5  rems (0.05 sievert).2 The effective dose equivalent is defined as:
where WT is a weighting factor and HT is the annual dose equivalent averaged over organ  or
tissue T.  Values of WT and their corresponding3 organs and tissues are:

                           Gonads	0.25
                           Breasts	0.15
                           Red bone marrow	0.12
                           Lungs  	0.12
                           Thyroid   	0.03
                           Bone surfaces	0.03
                           Remainder	0.30
    1 "Dose equivalent" is the product of the absorbed dose, a quality factor which varies with
the energy and type of radiation, and other modifying factors, as defined by the International
Commission on Radiation Units and Measurements.

    2 The unit of dose equivalent in the system of special quantities for ionizing  radiation
currently in use in the United States is the "rem".  In the recently-adopted international system
(SI) the unit of dose equivalent is the " sievert".  One sievert =  100 rems.

    3 "Remainder" means the five other organs (such as liver, kidneys, spleen, brain, thymus,
adrenals, pancreas, stomach, small intestine, upper large intestine, and lower large intestine, but
excluding skin, lens of the eye,  and extremities) with the highest doses.  The weighting factor
for each such organ is 0.06.

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For the case of uniform irradiation of the whole body, where HT may be assumed the same for
each organ or tissue, the effective dose equivalent is equal to the dose equivalent to the whole
body.

Other  Health Effects.   In addition  to the limitation  on effective dose equivalent, the  dose
equivalent, HT , received in any year by an adult worker should not exceed  15  rems (0.15
sievert) to the lens of the eye, and 50 rems (0.5 sievert) to any other organ, tissue (including the
skin), or extremity4 of the body.

Additional limiting values  which apply  to the control of dose  from internal exposure  to
radionuclides in the workplace are specified in Recommendation 4.  Continued exposure of a
worker at or near the limiting values for dose received in any year over substantial  portions of
a  working  lifetime  should  be  avoided.   This should normally be accomplished  through
application of appropriate radiation protection practices established under Recommendation  2.

4.  As the primary means for controlling internal  exposure to radionuclides, agencies should
require that radioactive materials be contained,  to the extent reasonable achievable,  so as to
minimize intake.   In controlling internal exposure  consideration should also be given  to
concomitant eternal exposure.

The control of necessary exposure of adult workers to  radioactive materials in the workplace
should be designed, operated, and monitored  with sufficient frequency to ensure  that, as the
result  of intake of radionuclides in a year, the following limiting values for control of the
•workplace are satisfied: (a) the anticipated magnitude of the committed effective dose equivalent
from such intake plus any annual effective dose equivalent from eternal exposure will not exceed
5 rems (0.05 sievert), and (b) the anticipated magnitude of the committed dose equivalent to any
organ  or tissue from  such intake plus any annual dose equivalent from external exposure will
not exceed 50 rems (0.5 sievert).  The committed effective dose equivalent from internal sources
of radiation, HE 50 is defined as:

                                   •^i.so =  T  WT "r.»
 where WT is defined as in Recommendation 3 and the committed dose equivalent, HT t J0, is the
 sum of all dose equivalents to organ or tissue T that may accumulate over an individual's
 anticipated remaining lifetime (taken as 50 years) from radionuclides that are retained in the
 body.  These conditions on committed doses should provide the primary basis for the control
 of internal exposure to radioactive materials.5
    4 »
       Extremity" means the forearms and hands, or the lower legs and feet.

    5 When these conditions on intake of radioactive materials have been  satisfied, it is not
 necessary to assess contributions from such intakes to annual doses in future years, and, as an

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In circumstances where assessment of actual intake for an individual worker shows the above
conditions for control of intake have not been met, agencies should require that appropriate
corrective action be taken  to assure control has been reestablished and that future exposure of
the worker  is  appropriately  managed.   Provision  should  be made to  assess annual dose
equivalents due to  radionuclides retained in the body from such intake for as long as they are
significant for ensuring conformance with the limiting values specified in Recommendation 3.

5.  Occupational dose equivalents  to individuals under the age of eighteen should be limited to
one-tenth of the values specified in Recommendations 3 and 4 for adult workers.

6.  Exposure of an unborn child should be less than that of adult workers.  Workers should be
informed of current knowledge of risks to the unborn6  from radiation and of the responsibility
of both employers and workers to  minimize exposure of the unborn. The dose equivalent to an
unborn as a result of occupational exposure of a woman who has  declared that she is pregnant
should be maintained as low as reasonably  achievable,  and in any case should not exceed  0.5
rem (0.005  sievert)  during  the entire gestation  period.   Efforts should be  made  to avoid
substantial variation above the uniform monthly exposure rate that would satisfy  this limiting
value.  The limiting value  for the unborn does not create a basis for discrimination, and should
be achieved  in conformance with the provision of Title VII of the Civil Rights Act of 1964, as
amended, regarding discrimination in  employment  practices,  including  hiring, discharge,
compensation, and terms,  conditions, or privileges of employment.

7. Individuals occupationally exposed to radiation and managers of activities involving radiation
should be instructed on the basic risks to health from ionizing radiation and on basic radiation
protection principles. This should, as a minimum, include instruction on  the somatic (including
in uiero) and genetic effects of ionizing radiation, the recommendations set forth in Federal
radiation protection guidance for occupational exposure and applicable regulations and operating
procedures which implement this guidance,  the general  levels of risk and appropriate radiation
protection practices for their work situations, and the  responsibilities  of individual  worker to
avoid and minimize exposure. The degree and type of instruction that is appropriate will depend
on the potential radiation exposures involved.

8. Appropriate monitoring of workers and the work place should be performed and records kept
to ensure conformance with these recommendations.   The types and  accuracy of monitoring
methods  and procedures utilized  should be periodically reviewed to assure that appropriate
techniques are being competently applied.
operational procedure,  such doses may be assigned to the year of intake for the purpose of
assessing compliance with Recommendation 3.

    6 The term "unborn" is defined to encompass the period commencing with conception and
ending with birth.

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Maintenance  of a  cumulative  record of lifetime  occupational  doses for each  worker  is
encouraged.  For doses due to intake of radioactive materials, the committed effective dose
equivalent and the quantity of each radionuclide in the body  should be assessed and recorded,
to the extend practicable.   A summary of annual, cumulative, and committed effective dose
equivalents should be provided each  worker  on no less than  an annual basis; more detailed
information concerning his or her exposure should be made available upon the worker's request.

9. Radiation exposure control  measures should be designed,  selected, utilized, and  maintained
to ensure that anticipated and actual doses meet the objectives of this guidance. Establishment
of administrative control  levels7  below the  limiting values  for control may be  useful and
appropriate for achieving this objective.  Reference levels* may also be useful to determine the
need  to take  such  actions as  recording,   investigation,  and  interventions.    Since such
administrative control and  reference levels will often involve ALARA considerations, they may
be developed for specific categories of workers or work situations.  Agencies should encourage
the establishment of measures  by which management can assess the effectiveness of ALARA
efforts,  including,  where  appropriate, local goals for limiting individual  and  collective
occupational  doses.   Supervision should be provided on a pan-time, full-time, or task-by-task
basis as necessary to maintain effective control over the exposure of workers.

10.  The  numerical  values recommended  herein  should not  be deliberately exceeded  except
during  emergencies, or under  unusual circumstances  for which  the  Federal  agency  having
jurisdiction has carefully considered the reasons for doing so in light of these recommendations.
If Federal agencies authorize dose equivalents  greater  than  these values  for unusual
circumstances,  they should make any generic procedures specifying conditions under which such
exposures may occur publicly available or make specific instances in which such  authorization
has been given a matter of public record.

The following notes are provided  to clarify application  of the above recommendations:

1. Occupational exposure of workers  does not include that due to normal background radiation
and  exposure as a patient of practitioners of the healing arts.

2.  The existing Federal guidance (34 FR 576 and 36 FR  12921) for limiting  exposure for
underground miners to radon decay products  applies independently of, and  is  not changed by,
these recommendations.
    7 Administrative control levels are requirements determined by a competent authority of the
 management of an institution or facility.  They are not primary limits,  and may therefore be
 exceeded, upon approval of competent authority or management as situations dictate.

    8 Reference levels are not limits, and  may be expressed  in terms of any useful parameter.
 They are used to determine a course of action, such as recording, investigation, or intervention,
 when the value of a parameter exceeds, or is projected to exceed, the reference level.

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3. The values specified by the International Commission on Radiological Protection (ICRP) for
quality factors and dosimetric conventions for the various types of radiation, the models for
reference persons, and the results of their dosimetric methods and metabolic models may be used
for determining conformance  to these recommendations.

4. "Annual Limits on Intake"  (ALIs) and/or "Derived Air Concentrations" (DACs) may be used
to limit radiation exposure from intake of or immersion in radionuclides. The ALI or DAC for
a single radionuclide is the maximum intake in a year or average air concentration for a working
year, respectively, for a reference person that, in the absence of any external dose, satisfies the
conditions  on  committed  effective dose equivalent  and  committed  dose  equivalent  of
Recommendation 4.  ALIs and DACs may be derived from different chemical or physical forms
of radioactive materials.

5.   The numerical  values provided  by  these recommendations do  not apply to workers
responsible for the management of or response to emergencies.

These recommendations would replace those portions of current Federal Radiation  Protection
Guidance (25 FR 4402) that apply to the protection of workers from ionizing radiation.  It is
expected that individual Federal  agencies,  on the basis of their knowledge of specific worker
exposure situations, will use  this new guidance as the basis upon which to revise or  develop
detailed standards and regulations to  the  extent  that they have regulatory or administrative
jurisdiction.  The Environmental Protection Agency will keep  informed  of Federal agency
actions to implement this guidance, and will issue any necessary clarifications and interpretations
required to reflect new information, so as to promote the coordination necessary to achieve an
effective Federal program  of worker protection.
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                                                                                                  Jur* 1974
                   U.S.  ATOMIC  ENERGY COMMISSION
                   REGULATORY      GUIDE
                   DIRECTORATE  OF  REGULATORY  STANDARDS
                                        REGULATORY GUIDE 1.86

                             TERMINATION OF  OPERATING LICENSES
                                      FOR NUCLEAR REACTORS
                 A.  INTRODUCTION

   Section 50.51, "Duration oflicensc, renewal," of 10
CFR Part 50, "Licensing of Production and Utilization
Facilities,"  requires thai  each  license  to operate  a
production  and  utilization  facibly  be   issued  for  a
specified duration  Upon  expiration  of  the  specified
penod, the license may be either renewed  or terminated
by  the  Commission. Section 50.82,  "Applications for
termination  of licenses," specifies the  requirements that
must  be satisfied to terminate  an operating license,
including the requirement that the dismantlement of the
facility  and  disposal- of the component  pans not  be
inimical to the common  defense and security  or to the
health  and safety of the pubbc. This guide  describes
methods and  procedures considered acceptable by the
Regulatory  staff  for  the  termination   of  operating
licenses for  nuclear reactors. The Advisory Committee
on  Reactor  Safeguards has  been consulted concerning
this guide and has concurred in  the  regulatory position.

                   B. DISCUSSION

   When a  licensee  decides  to  terminate his nuclear
reactor operating license, he may, as  a first step in the
process, request that his operating licensr be amended to
restrict him  to possess but not operate the facility. The
advantage to  the licensee  of  converting to such  a
possession-only license is  reduced surveillance require-
ments  in  that periodic  surveillance of equipment im-
portant to  the safety of  reactor operation is no longer
required. Once  this possession-only  license  is  issued,
reactor  operation is  not  permitted.  Other  activities
related to cessation of operations such at unloading fuel
from the reactor and placing it in storage  (either onsne
of offsite) may be continued.
    A  licensee  having  a  possession-only  license must
 retain, with the Part 50 license, authorization for special
 nuclear  matenal  (10  CFR  Part  70, "Special  Nuclear
 Material"), byproduct matenal (10 CFR Pan 30, "Rules
 of General  Appbcability to  Licensing  of Byproduct
 Material"), and   source  matenal  (10  CFR Part  40,
 "Licensing of  Source Material"), until  the  fuel, radio-
 active components, and sources  are  removed from  the
 facility.  Appropriate administrative controls and facility
 requirements are imposed by the Pan 50 license and the
 technical specifications to assure that proper surveillance
 is performed and that the reactor facility  is maintained
 in a safe condition and not operated.

   A  possession-only license  permits various options and
 procedures for decommissioning, such  as mothbaliing,
 entombment, or  dismantling. The requirements imposed
 rt»«T\AnrA nn tV^»> nntinn cr»lr»rM**r^
 depend on the option selected.
   Section  50.82 provides that the licensee may  dis-
 mantle and dispose of the component pans of a nuclear
 reactor  in  accordance  with  existing regulations.  For
 research reactors and  critical  facilities, this has usually
 meant  the  disassembly  of a  reactor  and its shipment
 offsite,  sometimes  to  another  appropriately licensed
 organization for further  use. The site  from which  a
 reactor has been removed must  be decontaminated, as
 necessary, and  inspected by  the Commission to  deter-
 mine whether  unrestricted access  can be approved. In
 the  case of nuclear  power  reactors., dismantling  has
 usually  been  accomplished  by shipping fuel  offsitc,
 making the reactor inoperable, and disposing of some of
 the radioactive components.

   Radioactive components may be either shipped  off-
 site  for burial at an authorized burial  ground or secured
              USA.EC REGULATORY GUIDES
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       9Ui6*t *•>.-l tx trviM< p*MO0'C*t'T  at *pp/
                                          to •ccommoon*
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O«*«r« th« Stcrtisry

Aiicmio.-!- Chxt. Pwbltc ^roQMdtnpi Stiff.

The puiQn »rc titwvtf •« the following l*n tvo*d O»vi»K>ni.

 1  l*o*«*t R»»cion                6 l*rocx*cti
 2  f\vt**icf\ »nd 7*il Re»etO'i

 <  f nvtfon«T»*nljtJ »«d Si»*n5
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                                                                                    &. Occup«l»orwl Hettth
                                                                                    fi. Aht.ttuit R*v*v>

-------
on the site. Those ladioactivr materials remaining on the
site must be isolated from the public by physical barriers
01 other means to prevent  public access to  hazardous
levels of radiation. Surveillance is necessary to assure the
long  term  integrity of  tne  barrier*  The amount  of
surveillance  required  depends  upon  (1)  the  potential
hazard  to  the  health  and  safety of  the public from
radioactive material  remaining  on the sue and (2) the
integrity of the physical barriers. Before  areas may be
released  for  unrestricted  ust,  the)   mus!  have been
decontaminated or tne radioactivity must have decayed
to less than prescribed limits (Table I).

   The  hazard  associated  with  the  retired  facility is
evaluated  by  considering  the  amount  and type  of
remaining contamination, the degree of confinement of
the remaining radioactive materials, the physical security
provided  by  the  confinement,  the   susceptibility  to
release  of  radiation  as  i  result of natural phenomena,
and the duration of required surveillance

             C  REGULATORY POSITION

 1. APPLICATION FOR A LICENSE TO POSSESS BUT
   NOT OPERATE (POSSESSION-ONLY LICENSE)

   A  request  to   amend  an  operating license  to a
 possession-only license should be made to the Director
 of Licensing, U.S. Atomic Energy Commission, Washing-
 ton,  D.C.  20545.  The  request  should  include   the
 following information.

   a. A description of the current status of the facility.

   b. A description of measures that will be taken to
 prevent cnticality or reactivity  changes and to minimize
 releases of radioactivity from the facibty

   c. Any proposed cnanges 10 the technical specifica-
 tions that  reflect  the possession-only facility status and
 the  necessary   disassembly/retirement  activities  to be
 performed.

   d. A safety analysis of  both the  activities  to be
 accomplished and the proposed changes to the technical
 specifications.

   e. An inventory  of activated  materials  and their
 location in the  facility.

 2. ALTERNATIVES FOR  REACTOR  RETIREMENT

   Four alternatives  for retirement of nuclear  reactor
 facilities are considered acceptable  by  the  Regulatory
 staff. These are:

   a.  Mothballing  Mothballing  of a  nuclear  reactor
 facility  consists of putting the facility in  a state of
 protective storage. In general, the  facility may be  left
 intact except that all fuel assemblies and the  radioactive
fluids  and  waste  snould  be  removed  from  the  sue
Adequate  radiation monitoring, environmental survei)-
lance,  and  appropriate  security procedures should be
estabhshed under a possession-only license to ensure that
the healtri and safety of the public is not endangered

   b  ln-Place  Entombment in-place entombment  con-
sists  of sealing all  tnt remaining highly radioactive or
contaminated  components  (e.g , the pressure vessel and
reactor  internals' within i  structure  integral  with the
biolopcal shielc after naving all fuel  assemblies,, radio-
active  fluids   and  wastes,   and  certain  selected   com-
ponents shipped offsite  The  structure should provide
integnu over the  period  of  time in  which significant
quantities (greater  thar  Table 1 levels)  of radioactivity
remain  with  tne   material  in  tne  entombment   An
appropriate anc continuing surveillance  program should
be established  under a possession-oniy license.

   c. Removal  of   Radioactive  Components  and  Dis-
mantling.  All   fuel assemblies,  radioactive  fluids  and
waste,  and other materials  having  activities above ac-
cepted  unrestricted activity levels  (Table I) should be
removed from the site. The facility owner may tnen  have
unrestricted  use of the site  with no requirement  for  a
license. If the  facility  owner so desires, the remainder of
the reactor facility may be  dismantled  and all vestiges
removed and disposed of.

   d. Conversion to a New  Nuclear System or a Fossil
Fuel System.  This alternative,  which applies  only to
nuclear power plants, utilizes the existing turbine system
with a  new  steam supply system.  The  original nuclear
steam  supply  system  should  be  separated  from the
electric generating  system  and disposed of in accordance
with one of the previous three  retirement alternatives.

3. SURVEILLANCE AND SECURITY FOR THE RE-
   TIREMENT  ALTERNATIVES  WHOSE   FINAL
   STATUS   REQUIRES   A  POSSESSION-ONLY
   LICENSE

   A  facility  which has  been  licensed  under  a posses-
sion-only license may contain a significant amount of
radioactivity in the form of activated  and contaminated
hardware  and structural  materials.  Surveillance  and
commensurate security should be provided to assure that
the public health and safety are not endangered.

   a.  Physical security to prevent inadvertent  exposure
of personnel  should  be  provided  by multiple locked
barriers. The  presence of these  barriers should make  n
extremely difficult for an  unauthorized person to gain
access to areas where radiation or contamination  levels
exceed those  specified  in  Regulatory  Position C.4. To
prevent inadvertent exposure, radiation areas above  5
mR/hr, such as near  the activated  primary system of a
power plant, should be appropriately marked and should
not be accessible except by cutting of welded closures or
the  disassembl) and  removal of substantial structures
                                                    1.86-2

-------
and/o:  smelding  material   Means  such  as  a  lemote
readout intrusion alarm  system should be provided to
indicate to designated personnel when  a physical barrif
is  penetrated  Secunt)   personnel  thai  provide access
control to the  facility  may  be used instead  of the
physical barriers and the intrusion alarm systems

   t  The  physical  barriers to  unauthorized entrance
into the facility, e.g., fences, Duildings,  welded doors.
and  access openings, should   be  inspected  a:  leas:
quarterly lo assure  thai these barriers have noi deterior-
ated and tnat IOCKS and locking apparatus  are intact.

   c  A facility radiation survey should be performed at
leasi quarter))' to verify  that no radioactive  material  is
escaping or oemg transported through the contammeni
barriers in  the facility Sampling should  be  done along
the mos;  probable  path  by which  radioactive  material
such  as thai stored in  the inner containment  regions
could be transported  to the outer regions of the facility
and ultimately to the environs.

   d.  An  environmental  radiation  survey   should  be
performed  at least  semiannual!)'  to verify  thai no
signficant amounts of radiation have been released to the
environment  from  the  facility.  Samples such  as soil,
vegetation, and water should be taken at locations for
which statistical  data has been established dunng reactor
operations.

   e.  A sue representative should be  designated to be
responsible for controlling authorized access into and
movement within the facility.

   f.  Administrative  procedures  should  be  established
for the notification and reporting of abnormal occur-
rences such .as (1) the  entrance of  an unauthorized
person or persons into the facility and (2) a significant
change in the radiation or contamination levels in the
facility or the offsite environment

   g.  The following reports should be made:

      (1) An annual report to the Director of Licensing,
U.S.  Atomic  Energy  Commission,  Washington,  D.C.
20545, describing the results of the environmental and
facility radiation surveys, the status of the facility, and
an  evaluation  of  the  performance   of security  and
surveillance measures.

      (2) An abnormal occurrence report to the Regula-
tory Operations Regional Office by telephone within 24
hours  of discovery  of an abnormal  occurrence. The
abnormal occurrence will also be reported in the annual
report described in the preceding item.

   h.  Records  or logs relative  to the following items
should be kepi and retained  until the license is termi-
nated, after which they may be stored  with  other plant
records:
      (}) Environmental surveys.

      (2) Facility radiation surveys,

      (3) Inspections of the physical barriers, and

      (4) Abnormal occurrences
4  DECONTAMINATION  FOR RELEASE  FOR  UN-
   RESTRICTED USE

   If it  is desired to terminate a license and to eliminate
any  further surveillance requirements, the facility should
be  sufficiently decontaminated  to prevent  risk to-the
public health  and safety. After  the  decontamination is
satisfactorily  accomplished  and the  site  inspected by
tne  Commission,  the  Commission may authorize the
license  10 be terminated and the facility abandoned or
released for  unrestncted  use. The licensee should per-
form the  decontamination using  ihe following  guide-
lines

   a.  The licensee  should  make a reasonable effort to
eliminate residual contamination.

   b. No covering  should  be   applied to  radioactive
surfaces of equipment  or structures by paint, plating, or
other covering material until it is known that contamina-
tion levels (determined by a survey and documented) are
below  the limits  specified in  Table I. In addition, a
reasonable  effort should be made (and documented) to
further   minimize  contamination  pnor to-any  such
covering.

   c.  The radioactivity of the interior surfaces of pipes,
drain lines,  or  ductwork  should  be  determined  by
making  measurements at all traps and other appropriate
access points, provided contamination at these locations
is  likely to be representative of contamination  on the
interior  of the pipes, dram lines, or ductwo'rk. Surfaces
of premises, equipment, or scrap which are'likely to be
contaminated  but  are  of  such size, construction, or
location as to make the surface inaccessible for purposes
of measurement  should be  assumed to be contaminated
in excess of the permissible radiation limits.

   d.  Upon request, the Commission may authorize a
licensee to relinquish possession or control of premises,
equipment, or scrap having surfaces contaminated in
excess of the limits specified. This may include, but is
not limned to, special circumstances such as the transfer
of premises to another licensed organization that will
continue to work with radioactive materials. Requests
for such authorization should provide:

      (1) Detailed,  specific information  describing the
premises, equipment,  scrap, and  radioactive contami-
nants and  the nature, extent,  and  degree  of residua]
surface  contamination.
                                                   1.86-3

-------
     (2) A OetaUed  health  ind safety in»)yas indi-
cating mat  the residual amounu of materials on jurface
areas, together with othet consideiations  such  as  the
prospective use of the premises, eouipmeni. or scrap, are
unlixely to result in an unreasonable  risk to the health
ana safety of the public.

   e  PriOJ  to release of  th?  premises  for unrestricted
use,  trie licensee should  make a comprehensive tadutior.
survey  establishing that  contamination is  within  tht
limits specified in Table  1 A survey report  should  be
filed with the Director of Licensing, U.S. Atomic Energ)
Commission, Washington, D.C. 20545,  with, » copy 10
the  Director of  the  Regulatory Operations  Regional
Office having jurisdiction. The report  shoidd be filed at
least 30 days prior to the planned date of abandonment.
The  survey  report should:

     (1) Identify the premises;

     (2) Show that reasonable effon has been made to
reduce residual contamination to as  low as practicable
levels,

     (3) Describe  the  scope  of the survey and  the
general procedures followed; and

     (4) State  the  finding  of the  survey  in units
specified in Table 1.

   After review  of  the  report,  the  Commission may
inspect  the facilities to  confirm  the  survey pnor to
granting approval for abandonment.

5. REACTOR RETIREMENT PROCEDUR£S

   As  indicated  in  Regulatory  Position  C.2,  several
alternatives are acceptable for reactor faci'ity retirement
If minor disassembly or "mothballin^" is planned, this
could be  done by the  existing operating and mainte-
nance  procedures  under the  license  in  effect.  Any
planned actions involving an unreviewed safely question
or a change  in  the  technical specifications should be
reviewed and approved  in accordance with the require
mentsof 10CFR §50.59.

   If maior structural changes to radioactive components
of  the  facility  are  planned,  such  as  removal of  the
pressure vessel  or  major components of  tne pnmars
system, a dismantlement plan including the information
required bv 550.82 should be submitted to the Commu
sior,  A dismantlement plan should be suomjitec  fo; aL
the  alternatives  of  Regulator)  Position  C.I excep!
mothballmg  Howeve:, minoi disassembly activities ma>
stili  be  performed  in   the  absence  of  such i  plan,
provided they are permitted by existing  operating and
maintenance  procedures A  dismantlement  plan snould
include the following'

   a.  A description of the ultimate status of the facility

   b.  A description of the dismantling activities and the
precautions to be taken.

   c.  A  safety  analysis of the  dismantling  activities
including any effluents which may  be released.

   d.  A  safety  analysis  of  the  facility in  Us ultimate
status.

   Upon satisfactory review  and  approval  of the dis-
mantling  plan,  a  dismantling order is  issued  by  the
Commission  in  accordance with  §50.82. When dis-
mantling  is completed  and  the Commission has been
notified by letter, the appropriate  Regulatory Opera-
tions  Regional Office  inspects the  facDity and verifies
completion in accordance with the dismantlement plan.
If residual radiation levels do not  exceed the values in
Table I, the  Commission may terminate  the license. If
these   levels  are  exceeded,  the   licensee  reiains  the
possession-only  license  under  which  the dismantling
activities have been conducted or, as an alternative, may
make application to the Slate (if an Agreement State)
for a byproduct  materials license.
                                                   1.86-4

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                                                          TABLE!

                                  ACCEPTABLE SURFACE CONTAMINATION LEVELS
NUCL1DE8
U-nai.U-235.U-23o.and
associated oecay prooucts
Transuramcs, Ri-226, Ra-22E,
Th-230. Th-22E.Pa-23!.
Ac-227, 1-125, J-I29
Th-nat,Th-232.Sr-90.
Ri-223.Ri-224,U.232,
1-126,1-131,1-133
Beta-gamma emJtters (nuchdes
wuh decay modes other tnan alpha
emission or spontaneous fission)
except Sr-90 and others noied above.
AVERAGE6 c
5.000 dpma/100 cm-
)00apm/100 cm-

lOOOdpm/lOOcm2

5 000 dpm H/ 1 00 cm-
MAXJMUMbd
15.000apma/10Gcm:
300dpm/100cm:

3000 dpm/ 100 cm2

15,000 dpm A-7/100 cm-
REMOVABLE6 e
1.000 dpm o/l 00 cm2
20 dpm/! 00 cm-

200 dpm/ 100 cm2

lOOOdpme-Y/IOOcm2
*Whcrc  surface contamination  by  both  alpha- and beia-gamma-emiitin{  nucbdcs exisu  the  limits edablished  for  alpha- and
 bcta-gamma-emitling nuchdei should apply independently.
 As used in Ihu table, dpm (diunicpations per minute) means the rale o! emission by radioactive maienal ai determined by correcting
 the counts per minute observed by an appropriate detcctoi for background, efficiency,  and (eomeiric factors associated with the
 instrumentation.
Measurements of average contaminant  should  not  be averaged  over more than 1 square  metei. For objects  of less surface area, the
 average should be derived for each such  object.
''The maximum conumination level applies to an area of not mere than 100 err, .
eThc amount of removable radioactive material  per  100 crrr of surface area should be determined by wipm;  that area with dry filter or
 soft absorbent papei, applying moderate pressure,  and assessing the amount of radioactive material  on the wipe with an appropriate
 instrument of known efficiency.  When removable  contamination on objects of  less surface  area is determined, the perunenl levels
 should be reduced proportionally and the entire surface should be wiped.
                                                         1.86-5

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-L'K FOP.:
                             NOV  4 1983
                          Aoministrators
                 Branch Chiefs
                 Division of Fuel Cycle
  DISTRIBUTION
  .RECunningham
  BSinger
  VLMiller
  LCRouse
  CEMacDonald
  WTCrow
  FCUF R/F
  NMSS R/F   FC  	
and Haterlal Safety,
                                                     DRChapel1
                                                     JGlenn, RI
JPotter,
BMallett
JEverett
RThomas,
                                                 Central  File
                                                      flhSS
 RII
,  Rill
,  RIV
 RV
FROM:
SUBJECT:
         Richard E. Cunningham. Director
         Division of Fuel Cycle and Uaterial Safety,  MMSS

         POLICY AND GUIDANCE DIRECTIVE FC 83-23 :
         TERMINATION OF BYPRODUCT, SOURCE AKD SPECIAL
         NUCLEAR HATERIAL LICENSES
Tnc enclosed final rule specifies licensee responsibility and requirements
for terminating a license Issued under 10 CFft Parts 30, 40 and  70.   Among
other things, a licensee is required to submit on or before the expiration
date a radiation survey report to confirm the absence of radioactive materials
or to specify existing levels of residual radioactive contamination present frwi
past operations.  A survey report is not required 1f a licensee can demonstrate
the absence of radioactive contamination in some other manner,  such as  the use
only of sealed sources that never showed evidence of leakage.   If detectable
levels of residual radioactive contamination attributable to licensed operations
are found, tht license continues in force until the Cotnraisslon  notifies the
licensee in v.-riting that the license is terminated.  The purpose of this
Memorandum is to provide guidance to the Regions and Headquarters staff on
the findings tnat need to be made before written notification is given ttiat
the license is terminated.

Pevjjew Procedure

Before terminating a license where residual  radioactive material contcmination
is present from past licensed operations, f.'RC should determine whether:

1.  a reasonable effort has been made to eliminate residual  contamination, and

C.  residual  radioactive contamination is acceptably low to permit unrestricted
    release of the affected facilities.

If the levels of residual  radioactive contamination on surfaces and In soil are &
snail  fraction of those normally acceptable  for unrestricted release (see Section
!>elow). it is not necessary for the licensee to describe the efforts he has made
to reduce contamination levels.
                  ,•

Policy and Guidance Directive FC H3-3:  Standard Review Plan (SRP) for
Termination of Special  Nuclear Material Licenses for Fuel  Cycle FedHtles,
contains  Information that  is generally useful  for terminating any byproduct,
source or special  nuclear  material  license.
orncc^
SXIMNAMCk
D»TC^

	

	

	

	 '• 	
	 1 	
	 MOV-1-0-4983 	

	
                       OFFICIAL RECORD  COPY

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                                             -  2--
          !n ,.iost cases involving short n.ilf-life  radlonucliCes  or  operations  involving
          only iCdleo sources, an independent  confirmatory  survey  by  II'^C will  not  be
          necessary.  Confirmatory surveys  should  always  be nade  if the  licensee's survey
          report appears suspect or past  licensee  operations  involved  the chemical processing
          cf hundreds of ni111 grans of plutonium,  tens  of  kilograms of  enriched uraniun  235
          or hundreds of kilograms of  source material.   For materials  licensees whicn  used
          ano processed hundreds ot nil 1icuries  of  long half-life  radlonuclides (> 1  yr).
          cont in*iatory surveys snoulii  also  be  made  in  all  cases.   If  it  is determined  that
          o confirmatory survey will t;e made,  a  notice  should  be  sent  to the licensee
          informing Inn that the equipment  and facilities  should  be held for f.'kC Inspection.
          Discretion nay te exercised  as  to whether a  confirmatory  survey is to be naae  if
          there is information dvailable, such as  inspection  reports,  v*i1ch provides  A
          basis for acceptance or Uie  licensee's survey.

                Containnaj-ion_ Levels C-enerajly Acceptable  for  Unrestricted Release

          o   SJTT ace Contamination - See  Enclosure  2

          o   Soil Contawiiiatlon - See  Enclosure  3

          o   '.iater Contamination -  If  surface  or groundwater  contamination 1s  below
              MPA's ilatlonal Interim Prinary Drinking Uater Kegulations  (EPA 570-9-76-003),
              the contaninfltion is acceptable for unrestricted  release.

          If  the levels of contamination  exceed  the levels discussed  above and a judgment
          is nade that further efforts to reduce the contamination  is  not necessary for
          termination of the license,  an  environmental  impact  assessment should be nade
          to  support the ternination.  Such cases  should be  brought to the attention  of   •
          the Director of the lHvision of Fuel Cycle and Material  Safety, NHSS. before 'the
          tsminaticn is dispatched.


                                                 Original Sirica ^7
                                                     D. R. uliT?ell
                                                                                                      4
                                                        L.  Cunningham. Director
                                             JMiivision of  Fuel  Cycle end
                                                  riatcrial  Safety,  fir'^S
     Enclosures :
     1.  Final  Rule:   /s:
-------
                   Acceptable  Soil Contamination Levels
        Kind of Material
Soil  Concentration Level
  for unrestricted area
  i)  Natural Uranium (U-238 +
      U-234) with daughters present
      and in equilibrium

 1i)  Depleted Uranium or Natural
      Uranium that has been separated
      from its daughters Soluble or
      Insoluble

iii)  Natural Thorium (Th-232 + Th-228)
      with daughters present and in
      equilibrium

 iv)  Enriched Uranium Soluble or
      Insoluble

  v)  Plutonium (Y) or (W)  compounds

 vi)  Am-241 (W)  compounds

vii)  All  Byproduct Material
viii) External  Radiation
      10 (pCi/grn of soil)
      35  (pCi/gm of soil)
      10  (pCi/gm of soil)

      30  (pCi/gm of soil)


      25  (pCi/gm of soil)

      30  (pCi/gm of soil)

      Soil concentrations
         shall  be determined
         on a case  by  case
         basis

      10  microroentgens/hr
         above  background
         measured at one
         meter  from the
         ground .surface
                                                   ENCLOSURE  3

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                                                                                                   Revision 2
                                                                                              D»c«mb«r 1987
                  U  S  NUCLEAR  REGULATORY COMMISSION

                  REGULATORY    GUIDE
                  OFFICE OF NUCLEAR REGULATORY  RESEARCH
                                         REGULATORY  GUIDE 8.13
                                               (Task OP  031-4)
                 INSTRUCTION  CONCERNING PRENATAL RADIATION  EXPOSURE
                A.   INTRODUCTION

  Section 19.12, "instructions to Workers," of 10 CFR
Part  19, "Notices, Instructions, and .Reports to Workers,
Inspections," requires  that alj individuals  working  in  or
frequenting any portion of a restricted ares1 be  instructed
in the  health  protection  problems associated with  expo-
sure  to radioactive  materials  or radiation,  in precautions
or  procedures to minimize exposure, and  in the regula-
tions  that they  are  expected  to observe. The present
10  CFR   Part  20,  "Standards  for  Protection  Against
Radiation," has  no  special limit  for exposure of  the
embryo/fetus.2  This guide describes  the instructions  an
employer  should  provide  to  workers  and supervisors
concerning biological  risks to the  embryo/fetus exposed
to  radiation, a  dose  limit for the embryo/fetus that is
under  consideration,  and  suggestions  for reducing  radia-
tion  exposure.

  This  regulatory  guide   takes   into  consideration  a
proposed revision to 10 CFR Part 20, which incorporates
the  radiation  protection guidance for the embryo/fetus
approved  by  the  President  in  January  1987 (Ref.  1).
This revision to Part  20  was issued in January 1986 for
comment  as  a  proposed rule. Comments on the guide as
it pertains to  the  proposed  Part  20 are  encouraged. If
the  new  Part  20 is  codified, this regulatory  guide  will
be revised to  conform to the new  regulation and  will
incorporate appropriate public comments.

  Any information  collection  activities   mentioned  in
this  regulatory  guide  are  contained  as requirements in
10  CFR  Parts 19  or 20, which  provide  the  regulatory
    Restricted area means any area thai ha* controlled access to
protect'individuals from being exposed to radiation and radioactive
material*.

    In conformity with the proposed revision to 10 CFR Part 20,
the term "embryo/fetus"  is  used throughout this document to
represent all stages of pregnancy.
basis  for  this  guide.     The  information  collection
requirements  in  10  CFR  Parts  19  and 20  have  oeen
cleared  under  OMB Clearance   Nos   3150-0044   and
3150-0014, respectively.

                  B.  DISCUSSION

   It has been known since 1906 that cells that are divid-
ing very rapidly and are undifferentiated in their structure
and function  are generally  more sensitive  to radiation. In
the embryo  stage,  cells  meet  both these criteria  and
thus would be  expected to  be- highly sensitive to radia-
tion.  Furthermore,   there  is  direct  evidence  that  the
embryo/fetus  is radiosensitive. There  is  also evidence
that it is  especially sensitive  to certain  radiation effects
during  certain   periods  after  conception,  particularly
during the first 2  to 3 months after conception  when a
woman  may not be aware  that she is pregnant.

   Section  20.104  of 10  CFR Part 20  places different
radiation  dose  limits on  workers  who  are minors  than
on adult  workers.  Workers  under the  age  of  18  are
limited  to one-tenth  of the  adult radiation dose limics.
However,  the present NRC regulations  do not establish
dose limits specifically for the embryo/fetus.

   The  NRC's  present limit  on  the radiation dose that
can  be  received  on the  job is  1,250 millirems  per
quarter (3 months).3 Working minors (those under 18)
are limited to  a dose equal  to one-tenth that of adults,
125 millirems  per  quarter.  (See  §  20.101  of  10  CFR
Part 20.)

   Because of  the  sensitivity of the unborn  child, the
NationaJ  Council on Radiation  Protection and  Measure-
ments (NCRP) has recommended  that the dose equivalent
    •*The limit is 3,000 millirems per quarter i! the worker's occups-
 tionaJ dose history is known and the average dose does not exceed
 5,000 miiiifgnu per year.
             USNRC REGULATORY GUIDES
 Regulatory Guloes are njued to Describe ana mane available to the
 public methods acceptable  to  tne NRC stall ol  Implementing
 specific pjrts  of the Commission's  regulations, to oelineate lecn-
 nlques usea by tne stiff In evaluating specific  problems or postu-
 lates accloents, or to  brovloe guloance to applicants. Regulatory
 Guides are  not substitutes lor regulations,  ano compliance  with
 them Is not reaulreo. Methods anc solutions oltterent from tnose set
 out In the guides will be acceptable If tney provide ' Oasis lor trie
 findings reoulslte  to trie Issuance or continuance ol 2 permit or
 license by tne Commission.
 Tnls guioe was issued after consideration ol comments received  Irom
 trie public.  Comments  ano suggestions lor improvements in tnese
 guides are encouragec  a! all  times,  and guides will  be revised, as
 appropriate, to accommodate comments ano to reflect new Informa-
 tion or experience.
 Tne guides are issued in the following ten broad divisions:
 1. Power Reactors
 2 Research an<3 Test Reactors
 3. Fuels and Materials Facilities
 <. Environment*) and Siting
6. Products
7. Transportation
E. Occupational Health
  Antitrust and Financial Reviev
 Written comments
                   DC iutxmiued to the Rules anc ProccCuie'.
 S. Materials and Plant Protection 10. General

 Copies o<  issued guides may be purchased from the Government
 Printing Office 41 me currenl  GPO price. Inlormatlon on current
 GPO  prices may  be obtained by contacting the Suoerlntenoent 01
 Documents, U.S.  Government  Printing  Office  Post  Office  Box
 370B2. Washington. DC 2001?-70B2.  telephone (202)275-2060 or
 (202)275-2171.

 issued guides rr\i\ also be purchased  from ihe rvationai Technical

-------
to tne  unborn  clula  from occuptuonii exposure of tne
expectant motnei be  iirrmed  to  500  miliirems fo; tnf
entire pregnancy  (Ref. 2). The 1987 Presidential guidance
(Ref  1) specifies an  effective dose equivalent limit of
500  millirems to the  unborn  child  if  the pregnancy has
beer declared by the motner, the  guidance  also recom-
mends  that  substantial variations in the rate  of exposure
be avoided  Tne  NRC (m § 20.208 of its proposed  revi-
sior.  to  Pan  20) has proposed  adoption  of the  above
limits on dose and rate of exposure

   Ir. 1971,  the  NCRP  commented on the  occupational
exposure  of  fertile  women  (Ref. 2) and  suggested  that
fertile women should be employed only where the annual
dose would be unlikely to exceed 2 or 3  rems and would be
accumulated  at 2 more or less steady  rate. In  1977, the
1CRP recommended  that, when pregnancy has been diag-
nosed,  the woman work  only where it is unlixely that the
annual dose would exceed 0.30 of the dose-equivalent limit
of 5 rems (Ref. 3).  In otner words, the 1CRP has recom-
mended  that pregnant women  not work where the annual
dose might exceed 1.5 rerr..

             C.  REGULATORY POSITION

   Instructions  on   radiation  risks should  be  provided
to  workers,  including  supervisors, in  accordance   with
§ 19.12  of  10 CFR Pan  19  before they are allowed  to
work in a  restricted  area.   In providing instructions on
radiation risks, employers should include specific instruc-
tions  aoout   tne  nsks  of  radiatior.  exposure  to  the
embryo/fetus.

   The instructions should be  presented  both orally anr1
in printed  form,  and the instructions should  include,  a.
a  minimum,  the  information  provided  in  Appendix  A
(Instructor's  Guide)  to  this  guioe  Individuals should  be
giver,  the  opportunity  to  ask  questions  and in  turn
should be  questioned to  determine wnether they  under-
stand  the instructions. An  acceptable method of ensuring
that the information is understood  is  tc give  a  simple
written   test  covering the  material  included  in Appen-
dix B (Pregnant Worker's  Guide).  This approach  should
highlight for  instructors  those  pans  of the instructions
that cause  difficulties  and  thereby  lead  to appropriate
modifications in the instructional cumculu:.-..

                D.  IMPLEMENTATION

   The  purpose of this section is to provide information
to  applicants  and  licensees  regarding  the  NRC  staffs
plans for using this regulatory guide.

   Except  in  those  cases  in  which  an  applicant  or
licensee  proposes  an  acceptable  alternative method  for
complying with specified  portions  of the  Commission's
regulations,  the  NRC  will use  the material  described
in  this  guide  to  evaluate   the  instructional program
presented  to individuals.,  including  supervisors, working
in or frequenting any portion of a restricted area.
                                                       £.13-2

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                                                 APPENDIX  A
                                            INSTRUCTOR'S GUIDE

                      EFFECTS  ON  THE EMBRYO/FETUS OF  EXPOSURE TO  RADIATION
                                  AND  OTHER ENVIRONMENTAL HAZARDS
   In   order  to  decide  whether  to  continue  working
while   exposed  to  ionizing  radiation  during  her  preg-
nancy, a woman  shoulc  understand the potential effects
on  an embryo/fetus,  including those  that  may be  pro-
duced  by various environmental  risks such as  smoking
and drinking. This will  allow her to compare  these risks
with  those  produced  by exposure to ionizing radiation.

   Table  1  provides information on the potential effects
resulting  from  exposure of an embryo/fetus to radiation
and  nonradiation  risks   The  second  column  gives the
rate  at  which  the effect  is produced  by natural  causes
in terms of the number per  thousand cases  The  fourth
column  gives  the  number  of  additional  effects per
thousand cases  believed  to be produced  by exposure to
the specified amount of the  risk facto:

   The  following  section  discusses  the  studies  from
which  the  information  in  Table  1  was  oenved.  The
results  of  exposure  of the  embryo/fetus  to  the  risk
factors  and  the   dependence  on the  amount  of the
exposure are explained.

1. RADIATION RISKS

1.1   Childhood Cancer

  Numerous studies of radiation-induced childhood cancer
have  been  performed,  but a number of them are con-
troversial. The National Academy of Science (NAS) BEIR
report  reevaluated  the data from  these studies and even
reanalyzed the results. Some of the strongest support for
a  causal  relationship is  provided by twin data from the
Oxford survey  (Ref.  4). For maternal radiation doses of
1,000  millirems, the excess number of deaths (above those
occurring  from natural causes)  was  found  to be  0.6
death  per thousand children (Ref. 4).

1.2   Mental Retardation »nd Abnormal Smallness  of the
      Head (Microcephaly)

   Studies of Japanese children who were exposed while in
the womb to the atomic bomb radiation at Hiroshima and
Nagasaki have shown evidence of both small head size and
mental retardation. Most of the  children  were  exposed  to
radiation doses in  the range of 1  to 50 rads. The  impor-
tance  of the  most  recent  study  lies  in  the fact that
investigators were able  to show  that the gestationaJ age
(age of the embryo/fetus after conception) at the time the
children were exposed was z  critical factor (Ref. 7). The
approximate nsk  of  small  head  size as c function  of
gestational age  is  shown in Table 1.  For  a  radiation dose
of  1,000 milliiems at A to 7 weeks after conception, the
excess  cases of small head  size was 5 per thousand; at 8
to 1) weeks, it was 9  per  thousand  (Ref. 7).

   In another  study, the highest risk of mental  retarda-
tion  occurred  During  the  8  to  15  week  period after
conception (Ref  8). A recent EPA  study (Ref.  16)  has
calculated  that  excess  cases  of mental retardation  per
live  birth he  between  0.5  and  4  per  thousand per rad.

1.3  Genetic  Effects

   Radiation-induced genetic effects have not been observed
to date  in  humans. The  largest  source of material for
genetic  studies involves  the  survivors  of Hiroshima  and
Nagasaki, but  the  77,000 births that  occurred  among
the survivors showed  no evidence of genetic effects.  For
doses received  by the  pregnant worker  in the course of
employment  considered in this guide, the dose  received
by the  embryo/fetus  apparently  would  have  a negligible
effect on descendants (Refs. 17 and 18).

2. NONRADIATION RISKS

2.1  Occupation

   A recent study (Ref. 9) involving the birth records of
130,000 children  in  the  State of Washington  indicates
that  the risk  of  death  to  the unborn child is related to
the  occupation of the  mother.  Workers in  the metal
industry, the  chemical industry, medical technology, the
wood industry, the textile industry, and farms exhibited
stillbirths or spontaneous  abortions  at  a rate of 90 per
thousand above  that  of workers in the control group,
which  consisted  of  workers  in several  other industries.

2.2  Alcohol

   It has been recognized  since  ancient times that alco-
hol  consumption had an effect  on the unborn child.  Car-
thaginian' law  forbade the consumption of wine on the
wedding night so  that  a  defective  child  might  not be
conceived.   Recent  studies  have  indicated  that small
amounts  of alcohol  consumption have only  the minor
effect  of reducing the  birth  weight slightly, but when
consumption increases to  2  to 4 drinics per  day, a pat-
tern  of  abnormalities  called  the  fetal  alcohol syndrome
(FAS)  begins  to appear (Ref. 11). This syndrome consists
of reduced growth in the unborn  child,  faulty brair, func-
tion, and abnormal  facial  features.  There is  & syndrome
that  has the  same symptoms as  full-blown FAS  that
occurs  in   children  born  to  mothers  who have  not
consumed  alcohol  This  naturally  occurring syndrome
occurs  IT.  about  1 to  'I  cases per  thousand (Ref.  10).
                                                     6.13-3

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                                                   TABLE 1

                            EFFECTS OF RISK FACTORS ON PREGNANCY OUTCOME
 Effect
 Number Occurring
from Natural Causes
 Risk Factor
Excess Occurrences
 from Risk Factor
 Cancer death in children
 Small head sue
 Small head size
 Mental retardation
 Stillbirth or spontaneous
 abortion
Fetal alcohol syndrome
Fetal alcohol syndrome
Fetal alcohol syndrome
Perinatal infant death
(around the time of birth)
Perinatal infant death
Perinatal infant death
 1 4 per thousand
 (Ref. 5)
 40 per thousand
 (Ref.6)

 40 per thousand
 (Ref. 6)

 4 per thousand
 (Ref. 8)
 200 per thousand
 (Ref.  9)
 1 to 2 per thousand
 (Ref. 10)

 1 to 2 per thousand
 (Ref. 10)

 1 to 2 per thousand
 (Ref. 10)

 23 per thousand
 (Refs. 13, 14)
23 per thousand
(Refs 13, 14)

23 per thousand
(Refs 13, 14)
 RADIATION RISKS

 Childhood Cancer

 Radiation dose of 1000 milhrems
 received before birth

 Abnormalities

 Radiation dose of 1000 milkrads
 received during specific periods
 after conception.

   4-7 weeks after conception


   8-11 weeks after conception


 Radiation dose of 1000 milkrads
 received 8 to  15 weeks after
 conception

 NONRADIAT1ON RISKS

 Occupation

 Work in high-risk occupations
 (see text)

 Alcohol Consumption (see text)

 2-4 drinks per day


 More than 4 dnnks per day


 Chronic alcoholic (more than
 10 drinks per  day)

 Chronic alcohobc (more than
 10 drinks per  day)

 Smoking

 Less than 1  pack  per day


One pack or more per day
0.6 per thousand
(Ref. 4)
5 per thousand
(Ref. 7)

9 per thousand
(Ref. 7)

4 per thousand
(Ref. 8)
90 per thousand
(Ref. 9)
100 per thousand
(Ref. 11)

200 per thousand
(Ref. 11)

350 per thousand
(Ref. 12)

170 per thousand
(Ref. 15)
5 per thousand
(Ref. 13)

10 per  thousand
(Ref. 13)
                                                   S.I 3-4

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   For mothers  who  consume 2  to  4 drinxs pe: d«l ,
the excess  occurrences number »bout  100  per tnousanc,
and  (or  those  who consume  more  than  4 drinks per
day,  excets  occurrences  number  200  per  thousand
The  most  sensitive  period   for  this  effect  of  alcohol
appears  to  be  the  first  few  weeks  after  conception,
before the  mother-io-be realizes she is pregnant (Refs  1C
and 1))  Also, 17% or  170 per thousand of the embryo/
fetuses of chronic  alcoholics develop FAS  and die before
birth (Ref  15).  FAS  was first identified in 1973 in tht
United States where less  than full-blown effects  of the
syndrome   are now referred  to as fetal alcohol   effects
(FAE) (Ref.  12).

2.3  Smoking

   Smoking   during  pregnancy  causes  reduced  birth
weights  in  babies  amounting  to  5 to 9 ounces  on the
average.   In addition,  there  is an  increased  nsk  of  5
infant  deaths per  thousand  for  mothers  who   smoke
Jess than one pack per day and  10  infant  deaths per
tnousand  for mothers  who smoke  one or  more- packs
per day (Ref. 13).

2.4   Miscellaneous

   Numerous other risks  affect the embryo/fetus,  only ^
few  of  which  are  touched  upon  here  Mo»t  people  are
familiar with  the  drug thalidomide  (a sedative  given to
some pregnant  women),  which  causes  children  to  be
born  with missing  limbs,  and the  more recent use  of  the
drug  diethylstilbestrol (DES),  a synthetic estrogen given
to  some  women   to  treat  menstrual  disorders,  which
produced  vaginal   cancers   in   the  daughters  born  to
women  who took   the drug. Living at high altitudes also
gives  rise to an increase in the number of low-birth-weight
children  born, while  an increase in Down's Syndrome
(mongoksm) occurs in children born to  mothers who  are
over  35  years of  age.  The  rapid growth in  the  us- of
ultrasound   in  recent  years  has  sparked   an  ongoing
investigation into   the   risks  of   using  ultrasound  for
diagnostic procedures (Ref.  19).

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                                                   APPENDIX B
                                          PREGNANT WORKER'S GUIDE

                         POSSIBLE HEALTH RISKS TO CHILDREN OF WOMEN WHO ARE
                                 EXPOSED TO RADIATION DURING PREGNANCY
   During pregnancy, you should be  aware of things  in
your  surroundings  or in your style  of life that could
affect  your unborn child. For those of you  who  work
in or  visit  areas designated  as Restricted  Areas (where
access  is  controlled  to   protect  individuals  from  being
exposed  to  radiation  and   radioactive  materials),  it  is
desirable  that  you  understand the  biological  risks  of
radiation to  your unborn child.

   Everyone  is exposed  daily  to  various kinds of radia-
tion:  heat,  light, ultraviolet, microwave, ionizing, and  so
on. For  the purposes of this  guide, only  ionizing radia-
tion  (such  as  x-rays',  gamma  rays,  neutrons, and  other
high-speed   atomic  particles)   is  considered.  Actually,
everything  is radioactive  and -all human activities involve
exposure  to radiation.  People  are  exposed to different
amounts  of  natural  "background"  ionizing  radiation
depending on where  they live. Radon gas in homes is a
problem of  growing concern. Background radiation comes
from three sources:
                                          Average
                                        Annual Dose

                                           50 millirem

                                           50 millirem

                                           25 millirem

                                          125 millirem*

                                   75 to 5,000 millirem
Terrestrial - radiation from soil
   and rocks
Cosmic - radiation from outer
   space
Radioactivity normally found
   within the human body
Dosage range (geographic and
   other factors)
   The  first two of these sources expose the  body from
the outside, and the last one exposes it  from the inside.
The  average  person is thus exposed  to  a total  dose  of
about  125 millirems  per  year  from natural background
radiation.

   In addition  to  exposure from  normal  background
radiation,  medical  procedures  may  contribute  to  the
dose people receive. The following table  lists the average
doses received  by the bone marrow  (the  blood-forming
cells) from different medical applications.
   'Radiation do*et in this document tit described in two different
uniU. The rad u * measure of the amount of energy absorbed in «
certain amount of material (100 erg: per gram}. Equal amount] of
energy abaorbed  from  different types  of radiation may lead  to
different biologic*] effectt. The rem  u s  unit that reflects the
biological damage done to the body. The nuUirad and millirem refer
to 1/1000 of a rad uid a rem, respectively.
                                                                X-Ray Procedure

                                                            Normal  chest examination
                                                            NormaJ  dental examination
                                                            Rib cage examination
                                                            Gall bladder examination
                                                            Banum  enema examination
                                                            Pelvic examination
                                       Average Dose*

                                          10 millirem
                                          10 millirem
                                         140 miliirem
                                         1 70 millirem
                                         500 millirem
                                         600 miliirem
                                                              •Variations by » factor of 2  (above and below) are not unusual.
                    NRC POSITION

   NRC  regulations  and  guidance  are  based  on  the
conservative  assumption that any  amount  of radiation,
no  matter how small, can  have  a  harmful effect  on an
adult,  child,  or unborn child. This assumption is said to
be  conservative because  there  are  no data  showing ill
effects  from  small  doses;  the  National  Academy  of
Sciences recently expressed  "uncertainty as to whether a
dose  of,  say,  1  rad  would  have any  effect  at all."
Although  it  is known that  the unborn  child  is more
sensitive  to  radiation  than  adults, particularly  during
certain  stages  of development,  the NRC  has not estab-
lished  a  special dose limit  for protection  of the unborn
child.  Such a limit  could  result in  job  discrimination for
women of child-bearing age and perhaps in the invasion
of  privacy (if pregnancy  tests were  required) if a sepa-
rate regulatory  dose  limit  were specified for the unborn
child.  Therefore, the  NRC  has  taken  the position  that
special protection  of the  unborn child should  be volun-
tary and  should be based on decisions made by workers
and employers  who  are  well  informed about  the risks
involved.

   For the NRC position to be effective, it is important
that both the  employee and the  employer understand
the risk  to  the unborn child from radiation received as
a  result  of  the occupational exposure of the mother.
This document tries  to  explain the  risk  as clearly as
possible  and   to  compare  it with  other  risks to the
unborn child  during  pregnancy.  It  is hoped   this  will
help pregnant  employees  balance the risk  to the, unborn
child  against  the  benefits  of employment  to  decide  if
the risk  is  worth  taking. This document  also  discusses
methods of  keeping  the dose, and  therefore the risk, to
the unborn child as low as is  reasonably achievable.
                                                      8.13-6

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               RADIATION DOSE LIMITS

   The NRC's present limit on the radiation dose that cm
I be received on  the job is 1,250 millirems per quarter (3
 months).* Working minors (those under 18) are limited to *
 dose equal to one-tenth  that of adults,  125  miHirems per
 quarter. (See § 20.101 of 10 CFR Part 20.)

   Because of the sensitivity of the unborn child, the National
 Council on Radiation  Protection and Measurements (NCRP)
 has  recommended that the dose equivalent to the unborn
 child from occupational exposure  of the  expectant mother
 be limited to 500 millirerns for the entire pregnancy (Ref. 2).
 The 1987 Presidential guidance (Ref. ]) specifies an effective
 dose equivalent limit of 500 millirems  to the unborn child if
 the pregnancy has been declared by the mother; the guidance
 also recommends  that substantial  variations in the rate of
 exposure be avoided.  The NRC(in § 20.208 of its proposed
 revision to Part 20)  has proposed adoption  of  the above
 limits on dose and rate of exposure.

      ADVICE FOR EMPLOYEE AND EMPLOYER

   Although  the risks to the unborn  child  are small under
 normal working conditions, it is still advisable to limit the
 radiation  dose from occupational exposure to no more than
 500 millirems  for the  total  pregnancy.  Employee and
 employer should work together to decide the best method
 /or accomplishing this goal Some methods that might be
 used include reducing the  time spent in  radiation areas,
 wearing some shielding over the abdominal area, and keeping
   extra distance from radiation sources when possible. The
 employer or  health physicist will  be  able to estimate the
 probable dose to the  unborn child during the normal nine-
 month pregnancy period and to inform the employee of the
 amount. If the predicted dose exceeds 500 millirems, the
 employee and employer should work out schedules or proce-
    The limit is 3,000 milliremj per quarter if the worker'* occupa-
 tional dole history is known ind the average  doie doe* not exceed
 5,000 miUirenu per yen.
dures to limit the dose to the 500-millirem recommended
limit.

   It   is  important  that  the  employee   inform  the
employer of her condition  as  soon as  she realizes she  is
pregnant  if  the dose  to  the  unborn  child  is  to  be
minimized.

                INTERNAL HAZARDS

   This document has been directed primarily  toward  a
discussion of radiation doses received from sources outside
the body.  Workers should  also be aware that  there is a
risk  of radioactive material entering the-body in work-
places  where unsealed radioactive material is used.  Nuclear
medicine  clinics, laboratories,  and certair. manufacturers
use radioactive material in bulk form, often as a liquid or a
gas.  A  list  of the commonly used materials and safety
precautions for each is beyond the scope of this document,
but certain general precautions might include the following:

   1.  Do not smoke, eat, drink, or'apply cosmetics
      around radioactive material.

   2.  Do not pipette solutions by mouth.

   3.  Use  disposable gloves  while handling radioactive
      material when feasible.

   4.  Wash hands after working around radioactive
      material.

   5.  Wear lab  coats or  other  protective clothing when-
      ever there is a possibility of spills.

   Remember that  the  employer  is  required  to have
demonstrated that  it  will have safe  procedures  and
practices  before the  NUC issues  it  a license  to use
radioactive  material.  Workers are urged to follow estab-
lished  procedures and  consult  the employer's radiation
safety  officer or health physicist  whenever problems  or
questions arise.
                                                       8.13-7

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                                                 REFERENCES
    "Federal  Radiation Protection  Guidance for Occupa-
    tional Exposure," Federal Register t p. 2822, January 27,
    1987.
11   D  W  Smith, "Alcohol Effects on the Fetus," Prop-ess
    ir. Clinical and Biological Research, Vol. 36, pp. 73-82,
    1980
                                                       4
2.   National Council on Radiation Protection and Measure-
    ments, "Basic  Radiation Protection  Cnteru," NCRP
    Report No  39, 197i

3.   International Commission on Radiolopcal Protection,
    "Recommendations of the International  Commission
    on Radiological Protection," 1CR? Publication No. 26,
    Vol  1, No  2, 1977.

A.   National Academy of Sciences, "The Effects on Popula-
    tions of Exposure  to Low Levels of Ionizing Radiation
    (BEIR III)," Nauonal Academy Press, Washington, DC,
    1980.

5   J   L  Young  anc1  R  W. Miller, "Incidence  of Malig-
    nant  Tumors  ir.  U.S.  Children," Journal of Pedia-
    trics,pp. 2S4-25E, 1975.

6.   W.  J.  Blot,  "Growth  and  Development  Following
    Prenatal and  Childhood  Exposure to Atomic Radia-
    tion,"  Journal of Radiation  Research (Supplement),
    pp. 82-85, 1975.

7.   R. W. Miller and J. J. Mulvihill, "Small Head Size After
    Atomic Radiation,"  Teratology, Vol. 14,  pp.  355-
    358, 1976.

8.   M. Otake  and W, J.  Schull, "In Utero  Exposure to
    A-bomb Radiation and Mental Retardation; a Reassess-
    ment,"  TTie Bntish Journal  of Radiology, Vol. 57,
    pp. 409-414, 1984.

9.   T. L  Vaughan  et  al,  "Fetal  Death and  Maternal
    Occupation,"  Journal   of  Occupational  Medicine,
    Vol. 26, No. 9, pp. 676-678, 1984.

10. J.  W.  Hanson,  A.  P.  Streissguth,  and D. W.  Smith,
    "The   Effects   of  Moderate  Alcohol Consumption
    During Pregnancy on Fetal Growth and Morphogenesis,"
    Journal of Pediatrics, Vol. 92, pp. 457-460, 1978.
12  L.  E  Robe,  "Alcohol and Pregnancy," The American
    Medical Association, box 10946, Chicago, 1984
13  M. B. Meyer and J,  A  Tonascia, "Maternal Smoking,
    Pregnancy Complications,  ant)  Pennatal  Mortality,"
    American Journal of Obstetrics and Gynecology , Vol.
    121,  Nc. 5, pp. 494-502, 1977

14.  R  H Molt, "Radiation Effects  on  Pre-NataJ Devel-
    opment  and  Tneir  Radiological Significance," Tne
    British Journal of Radiology, Vol  52,  No.  614, pp.
    89-101, February 1979.

15.  D  A Roe,  Alcohol and  the  Diet,  AVI Publishing
    Company inc., Westport, Connecticut, 1979.

16.  Environmental Protection  Agency,  "Radionuclides,"
    Background   Information   Document   EPA   520/
    1-84-022-1, pp. 8-56 -  8-63.
17. G. W.  Beebe,  "The Atomic Bomb Survivors and the
    Problem  of  Low-Dose  Radiation Effects," American
    Journal  of  Epidemiology,  VoL  114,  No.  6,  pp.
    761-783, 1981.
18. W.  J. Blot et al.,  "Reproductive  Potential of Males
    Exposed in Utero or Prepubertally to  Atomic Radia-
    tion," in Atomic  Bomb Casualty  Commission Tech-
    nical Report  TR-39-72, Radiation Effects Research
    Foundation, Hiroshima, Japan, 1972.
19. National Council on Radiation Protection and Measure-
    ments, "Protection in  Nuclear  Medicine and  Ultra-
    sound  Diagnostic  Procedures  in  Children," NCRP
    Report No. 73, 1983.
                                                    8.13-8

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                                          VALUE/IMPACT STATEMENT

    A  draft value/impact  statement  was  published  with      final guide  has no! been prepared  A  copy of the draft
 the  proposed  Revision  1  to  Regulator)  GuiOt  S.I 3      value/impact  statement  is  available  for  inspection  and
 (Tasx OP  031-4) wher. the  draft puioe  was published for      copying for  a fee  a: the Commission's Public Document
 public  comment   in  August   19ti   No  changes  were      Roorr,  at  1717  H  Street NW., Washington, DC,  unoe:
 necessary, so  t  separate  value/impact  statement fo: the      Task OF 031-4
.U.S. C.T-.O.  198?-202-29ii603ie


                                                    8.13-9

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                                            DARCOWrP 385-1
            Fundamentals of Health
            Physics for the Radiation
            Protection Officer

                 EXEMPT AR 325-15, PARAGRAPH 5-2D
HQ, U.S. ARMY MATERIEL DEVELOPMENT AND" READINESS COMMAND  SEPTEMBER 19E3

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                           DISCLAI.V :R

This report was prepared as an account of work sponsored by an agency of the
United States Government. Neither the United States Government nor any
agency thereof, nor any of their employees, makes any warranty, express or
implied, or assumes any legal liabilnv or responsibility (or the accuracy, com-
pleteness, or usefulness ol anj information, apparatus, product, or process
disclosed, or represents that us use would not infringe privately owned rights.
Reference herein to any specific commercial  product, process, or service by
trade  name, trademark, manufacturer, cr otherwise, does  not necessarily
constitute or imply its endorsement, recommendation, or favoring by the
United Stales Government or any agency thereof. The views and opinions of
authors expressed herein do not necessarily state or reflect those of the United
Mates Government or any agency thereo!.
               PACIFIC NORTHWEST LABORATORY
                           operated by
                            BATTEi i,E
                              for th;
            UNITED STATES DEPAR MEK'T OF ENERGY
               under Contract D£-/ C06-76RLO TB30

-------
FUNDAMENTALS OF HEALTH PHYSICS FOR
THE RADIATION PROTECTION OFFICER
B. L. Murphy
R. 0. Traub
R. L. Gilchrist
J. C. Mann
L. H. Munson
E. H. Carbaugh
J. L. Baer
Contributors:
  L.  W. Brackenbush
  T.  H. Essig
  D.  E. Hadlock
  W.  N. Herrington
  M.  P. Moeller
D.  W. Murphy
0.  M. Set by
J.  E. Tanner
J.  M. Taylor
C.  M. Unruh
March 1983
Prepared for the U.S. Department of the Army
Pacific Northwest Laboratory
Richland, Washington 99352

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                                                                     DARCOK-P 385-1
                               DEPARTMENT OF THE ARMY
          HEADQUARTERS, US ARMY MATERIEL DEVELOPMENT AND READINESS COMMAND
                    5001 Eisenhower Avenue, Alexandria, VA  2233i
DARCOM PAMPHLET
NO.       385-1
18 April 1984
                                      Safety

                        FUNDAMENTALS OF HEALTH PHYSICS FOR
                         THE RADIATION PROTECTION OFFICER

 1.  Purpose.  This pamphlet provides the fundamentals of health physics for
 radiation protection officers.

 2.  Applicability.  This pamphlet is applicable to elements who possess and use
 radioactive materials.

 3.  Explanatory.  A consolidated "contents" page appears on page iv; "contents"
 pages appear at the beginning of each chapter.
      The proponent of this pamphlet is the US Army Materiel Development and
      Readiness Command.  Users are invited to send comments and suggested
      improvements on DA Form 2028 (Recommended Changes to Publications and
      Blank Forms) to Commander, DARCOM (DRCSF-P), 5001 Eisenhower Avenue,
      Alexandria. VA  22333.
FOR THE COMMANDER:
OFFICIAL:
                                          CLAUDE M. K1CKL1GKTER
                                          Major General, USA
                                          Chief of Staff
DONNA H.
CPT, CS  I/
HQ Adjutant

DISTRIBUTION
HQDA (DASG-PSP-E) WASH DC  20310 (3)
Initial Distr (78) i es HQ Dir/Actv/Staff Ofc)
A Pub Distr (50)
DRXAK-ABS Stock (50)
B LEAD Distr (2,279):  Safety officers ONLY 2 ea
                       Cdr, Belvoir R&D Center (STR3E-VR), Ft
                         Belvoir, VA  22060 (200)

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                                   CONTENTS







CHAPTER 1.   PROPERTIES OF RADIOACTIVE MATERIALS  •	1.1



CHAPTER 2.   RADIATION INSTRUMENTATION   	    2.1



CHAPTER 3.   RADIATION PROTECTION PROGRAM     	    3.1



CHAPTER 4.   RADIATION SURVEY PROGRAMS   	    4.1



CHAPTER 5.   INTERNAL EXPOSURE 	    5.1



CHAPTER 6.   EXTERNAL EXPOSURE 	    6.1



CHAPTER 7.   DECONTAMINATION   	    7.1



CHAPTER 8.   SELECTION AND DESIGN OF RADIATION FACILITIES    .     .     .8.1



CHAPTER 9.   TRANSPORTATION OF RADIOACTIVE MATERIALS    .    .     .     .9.1



CHAPTER 10.  MANAGEMENT OF LOW-LEVEL RADIOACTIVE WASTE      ,     .     .10.1



CHAPTER 11.  RADIATION ACCIDENTS AND EMERGENCY PREPAREDNESS .     .     .11.1



CHAPTER 12.  TRAINING	12.1



CHAPTER 13.  RECORDKEEPING    	   13.1



CHAPTER 14.  OUALITY ASSURANCE PROGRAM  	   14.1



CHAPTER 15.  APPRAISAL OF RADIATION PROTECTION PROGRAMS     .     .     .   15.1



CHAPTER 16.  REFERENCE DATA	16.1



Appendix A  ...     ••........   A-l
                                      IV

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                                                               DARCOM-P  385-1
               CHAPTER 1.  PROPERTIES OF RADIOACTIVE MATERIALS
1.1  ATOMIC STRUCTURE    	    1.5
     1.1.1  The Nucleus	1.6
     1.1.2  Electrons	              1.8
1.2  RADIOACTIVITY AND RADIOACTIVE DECAY     	    1.9
     1.2.1  Characterization of Radionuclides     .....    1.10
            A.  Rate of Decay	1.10
            B.  Energy of Decay    .     .     .     .     .     .    .     .    1.11
            C.  Type of Radiation Emitted     .                    .         1.15
     1.2.2  Decay Pathways	1.19
     1.2.3  Quantification of Radioactivity  	    1.20
1.3  INTERACTIONS OF RADIATION WITH MATTER   	    1.21
     1".3.1  Alpha and Beta Particles	1.21
            A.  Energy Transfer Processes     .     .     .     .    .     .    1.21
            B.  Alpha Particle Interactions  	    1.22
            C.  Beta Particle Interactions	1.23
     1.3.2  Photons	1.24
            A.  Energy Transfer Processes     .     .     .     .    .     .    1.25
            B.  Photon Interactions     .......    1.26
     1.3.3  Neutrons	1.26
            A.  Energy Transfer Processes     	    1.27
            B.  Neutron Interactions     .......    1.28
1.4  RADIATION QUANTITIES AND UNITS     	    1.29
     1.4.1  Exposure	1.29
     1.4.2  Absorbed Dose	1.30
                                      1.1

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DARCOK-P 385-1

     1.4.3  Relative Biological  Effectiveness     .....    1.31

     1.4.4  Dose Equivalent   .........    1.31

1.5  BIOLOGICAL EFFECTS OF RADIATION    	    1.33

     1.5.1  Genetic Effects	1.34

     1.5.2  Somatic Effects	1.34

            A.  Prompt Effects	1.34

            B.  Delayed Effects	1.34

            C.  Relationship Between Exposure and Delayed Effects      .    1.36

     1.5.3  Environmental Dose and Occupational Dose Limits .     .     .    1.36

1.6  PROPERTIES OF RADIOACTIVE MATERIALS IMPORTANT IN THE
     DEVELOPMENT OF RADIATION PROTECTION PROCEDURES    ....    1.38

     1.6.1  External Versus Internal Exposure     .....    1.38

            A.  External Exposure  .    .    .    .    .    .     .     .    1.38

            B.  Internal Exposure  ........    1.38

     1.6.2  Dispersibility	1.39

            A.  Nondispersible     ........    1.39

            B.  Limited Dispersibility  	    1.39

            C.  Dispersible	1.40

            D.  Highly or Readily Dispersible     	    1.40

     1.6.3  Chemiccl Toxicity 	    1.40

     1.6.4  Radiotoxicity	1.41

     1.6.5  Criticslity	1.41

            A.  The Double-Contingency Rule	1.43

            B.  Factors That Affect Criticality    	    1.44

REFERENCES	1.46

APPENDIX A  -  DECAY SCHEMES	1.47
                                       1.2

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                                                               DARCOM-P 385-1

                                    FIGURES


1.1  Numbers of Neutrons and Protons in Stable  Nudities      .     .     .     1.7

1.2  Schematic Diagram of an Atom Showing  Nucleus  and
     Electron Shells     	     1.8
               198
1.3  Number of    Au Atoms Present as a Function  of
     Half-Lives Elapsed  	     1.11

1.4  Energy-Velocity Relationships for Alpha  and  Beta  Particles   .     .     1.13

1.5  The Electromagnetic Spectrum  .    .     .     .     .     .     .     .     1.14

1.6  Relative Importance of the Photoelectric Effect,
     the Compton Effect, and Peir Production	1.25
                                    TABLES


1.1  Radiation Characteristics     ........     1.16

1.2  Effect of Common Decay Types on the Parent Nucleus     .     .     .     1.17

1.3  Classification of Neutrons    ........     1.27

1.4  Relationship of LET and Quality Factor	1.32

1.5  Recommended Values of Q for Different Types of
     Radiation	1.32

1.6  Dose-Effect Relationship for Acute Whole-Body Irradiation   .     .     1.25

1.7  Dose-Effect Relationship for Acute Partial-Body Irradiation      .     1.35

1.8  U.S.  General-Population Dose Estimates (1978)     ....     1.37

1.9  Maximum Dose Equivalent Per Celendar Quarter .....     1.37

1.10 Radiotoxicity of Various Nuclides  .......     1.42

1.11 Fissionable Materials    .    .    .     .     .    .     .     .     .     1.43
                                      1.3

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                                                               DARCOM-P 385-1
                CHAPTER  1.   PROPERTIES OF RADIOACTIVE MATERIALS
      The  world  around us  is  composed of elements and combinations of elements,
 each  with its own  unique  chemical  properties.  Only about 100 elements are
 known to  man.   Some examples  are hydrogen, oxygen, cerbon, gold, and silver.
 Substances  such  as water,  wood, rock,  rubber, coal, and hundreds of thousands
 more  are  combinations of  the  comparatively few elements.  Tnese combinations
 are called  compounds.
      Each element  can be  denoted by a  one- or two-letter chemical symbol; for
 example,  H  is the  symbol  for  hydrogen, 0 is the symbol for oxygen, and Au is
 the symbol  for  gold.  Compounds are denoted by' combinations of element symbols
 and numbers  that refer to  the proportion of each element in the compound.
 Water,  for  example, which  has two  units of hydrogen for every unit of oxygen,
 is designated H^O.  A list of all  of the known elements and the chemical
 symbol  for  each  can be found  in Chapter 16, "Reference Data."
      Some atoms  are unstable  and undergo transitions that result in the forma-
'tion  of a more  stable atom and the release of some energy.  This process is
 called  radioactive decay,  and substances that are unstable and subject to
 decay are called radioactive  materials.
      This chapter  provides a  review of the fundamental characteristics of radio-
 activity.   The  initial portion covers  basic information about atomic structure
 and radioactive  decay.  The properties of ionizing radiation are then reviewec1,
 followed  by  a discussion  of radiation  quantities and units.  Information on
 the biological effects of  radiation is presented.  The chapter concludes with
 the presentation of concepts  important to the development of radiation protec-
 tion  procedures.
                         Section 1.1  ATOMIC STRUCTURE

     The smallest unit of an element is the atom.  An atom consists of a small,
dense, positively charged nucleus surrounded by a cloud of negatively charged
electrons.
                                      1.5

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DARCOM-P 385-1
1.1.1  The Nucleus
     The nuclei1? consists of two fundamental  particles,  protons  end neutrons.
The proton is a positively charged particle that has  a  unit  charge of 1.6 x
   19                          "                  -24
10"   coulombs.  The mass of a proton is 1,67 x 10    gram.  Tne  number of
protons in the nucleus, the atomic number, Z,- is unique  for  each element; for
example, if a nucleus contains six protons, the atom is  a carbon atom; on tne
other hand, if a nucleus contains eight protons, the atom is an  oxygen atom.
     The neutron is a particle that has no electrical charge and has a mass
slightly greater than that of a proton.  The nuclei of the atoms that make up a
given element may contain varying numbers of neutrons.   The  number of neutrons
in the nucleus, the neutron number, N, influences the stability  of the nucleus;
that is,- it determines whether the atom is radioactive.   If  the  N number of a
nucleus is plotted as a function of the Z number of-the nucleus, as shown in
Figure 1.1, stable, or nonradioactive, nuclei tend to be clustered about e line
called the 1ine of stability.  In the case of nuclei of low Z, the most stable
nuclei have approximately equal numbers of protons and neutrons.  In the case
of very heavy nuclei (those with many protons, or high Z), the nucleus is most
stable if the number of neutrons in the nucleus is about 1.5 times the number
of protons.
     Isotopes are atoms of one element that have the same atomic number but
that differ in neutron number.  The isotopes of a given element have the same
•chemical properties and cannot be separated by chemical methods.  However, the
nuclear characteristics of the isotopes may be quite different; for example,
some isotopes of en element may be radioactive while others are not.   Isotopes
of a given element are identified by their mass number, A, which  is the  total
number of protons plus neutrons in the nucleus; that is, A  = Z +  N.
     Individual atoms are called nuclides; the  radioactive  forms  are  called
redionuclides.  An isotope or nuclide may  be  identified by  its chemical  symbol,
with the atomic number, Z, as a presubscript  and the nass number,  A,  as  a  pre-
superscript:  -X, where X represents the  chemical  symbol.   Because  the atomic
number, Z,  is unique to a given element,  it  is  often omitted  from this ncta-
tion.  Sometimes a nuclide is designated  by  the full name of  the  element,
                                       1.6

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                                                             DARCOM-P 385-1

120


100

•
8C
s
i
ac.
5 60
LU
Z
o
CO
1 40
2

20

0
..r
j.1"
. r"
. »". *
— Mr; /
. "1 ' X
— . .: . x
LINE OF STABILITY ^ If /"
~ . us : ^
• r' • X
.?!'•
:P. ,'N-Z
.'iV1' /
— . .n ° x
~r: x
**" /
!•• X
Pv
:;v
"xv
— — *»•>*•
y
.:' 1 1 I 1 I 1 1 I 1
                          20        40        60        80

                               NUMBER OF PROTON S (Z)
100
       FIGURE 1.1.   Numbers of Neutrons and Protons  in Stable Nuclides
                                                                      12
or its chemical  symbol,  followed by a hyphen and the A number.  Thus,   C,
l ?
 tC, C-12, and carbon-12 ere four ways of designating the same nuclide.   In
 o
the past, the A number was written with the chemical symbol as a postsuper-
         12
script, C  .

     The natural elements of the earth's crust or atmosphere are composed of

mixtures of the isotopes of each element.  The isotopes vary in their percent

natural abundance; that is, they do not ell occur in ecjal amounts.  For

example, of all  the oxygen atoms that occur on earth, 99.756* ere   0, 0.034%
                                      1.7

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DARCOM-P 385-1
    17
              18,
are "0, and 0.205°* are   0.   The relative abundance  of stable  isotopes remains
fairly constant over a wide geographic range.
1.1.2  Electrons
     The nucleus is surrounded Dy electrons,  which have a  negative cnerge that
is equal in magnitude, but opposite in sign,  to that  of the proton.   In the
neutral, uncharged atom, there is one electron outside the nucleus for every
proton in the nucleus.  The electrons can be  thought  of as occupying orbits, or
shel1s, as shown in Figure 1.2.  Because the  photons  give  the nucleus a
positive charge and the electrons have a negative charge,  and because opposite
charges tend to attract each other, there is  an attractive force between an
atom's nucleus and its electrons.  The shells represent the strength ot the
attractive force between the nucleus and the  electrons, not the exact location
of the electrons.
     The shells form a series of energy or quantum levels.  The diameters of
the shells are large in comparison with the diameter of the nucleus.  The
                                                        M N  0
FIGURE 1.2.
                       Schematic Diagram of an Atom Showing Nucleus and
                       Electron Snells
                                      1.8

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                                                              DARCOM-P  385-1
shells are identified by either e  letter (K,  L,  K,  N, 0,  P,  C) or  a  quantum
number (1, 2, 3, £, 5, 6, 7).   The energy state  of  each electron  in  a  shell
is completely described oy four independent quantum letters  (n, 1, m,  end  s),
and the Pauli Exclusion Principle  sets  an upper  limit on  the  number  of
possible electrons in each shell.
     Because of the attractive force between  the nucleus  and  the  electrons,
it takes a certain amount of energy to  remove the electrons  from  the stor..
The amount of energy required  to completely remove  an electron from  the  atom
is called the electron binding energy.   This  energy is different  for each
shell  in the atom of any one element, and different for the  same  shell in
different elements.  The electrons closest to the nuclei.-:.,  in the  K  shell,
have a greater attraction to the nucleus than electrons farther from the
nucleus.  The electron binding energy associated with an  inner shell is
therefore greater than that of en  outer shell.
     If an electron is removed from an  inner  shell, a vacancy, or "hole,"  is
formed in that shell.  An electron from one of the  outer  shells may  then
"jump" or "fall" into the vacancy.  When this happens, energy equal  to the
difference between the electron binding energies of the two  shells is emitted
from the atom in the form of electromagnetic  radiation.   This radiation  is
celled cnaracteris t i c radiation because the amount  of energy released is
characteristic of a given element.  Characteristic  radiation may  be  given  off
in the form of light, heat, or x rays,  depending upon  its energy.
              Section 1.2  RADIOACTIVITY AND RADIOACTIVE DECAY

     Radioactivity is the tendency of unstable nuclides to undergo radioac-
tive decay.   Radioactive decay is Defined as a spontaneous, energy-re";casing
atomic transition that involves a change in the state of the nucleus of an atom.
This change  means that the atom changes from one nuclide (the parent) into a
second nuclide (the daughter)  or from one nuclear energy level to a lower
energy level.   The difference  in the energy levels determines the amount of
energy recessed by the transition.   The transition must be spontaneous, that
is, free from  the influence of outside forces.  It is possible to use machines
                                      1.9

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DARCOM-P 3S5-1
such as cyclotrons, linear accelerators,  or even nuclear reactors to change the
nucleus of an atom; however, such transitions are not  considered radioactive
decay.
1.2.1  Characterization of Radionuclides
     A radioactive ruclide, or radionuclide,  can be characterised by its rate
of decay, ttie energy rele.--.sed during  the  decay,  and the type of raciation emit-
ted by the decay.
     A.  R;:te of Decay.  All radionucl ides do not decay at the same rate.  Some
decay very quickly, in a matter cf a  few  seconds.  Others may take days, weeks,
or millions of years to decay.  The rate  of decay of a radionuclide is measured
in terms of a helf-1ife.
     The half-life of a radionuclide, symbolized t, ,^ "is the time required
for the number of radioactive atoms present to decrease by one half.  After one
half-life, 50* of the original radioactive atoms remain; after two half-lives,
2.5% of the original radioactive atoms remain; etc.  Figure 1.3 illustrates the
                           198
concept of half-life using    Au, an  isotope of gold,  as an example.  The
half-life of a particular radionuclide  may be found in the Table of Isotopes
(Lederer and Shirley 1978) or the Radiological Health  Handbook (1970).
     The rate of radioactive decay can  also be expressed in terms of the decay
constant, x, of the radionuclide.  The  decay constant  indicates the fraction of
radioactive atoms present that will undergo radioactive decay in a given period
of time.  It is numerically equal to  the  natural logarithm of 2 (0.693) divided
by the half-life of the radionuclide.  That is,

                X = (In 2)/t1/2 = 0.693/t1/2                          (1.1)

The decay constant is used when calculating the number of radioactive atoms
present in a sample at any time, using  the equation

                N = N  e'U                                           (1.2)
                                     1.10

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                                                                DARCO:---P  385-1
      10,000
       8000
   a:
   c_
o   6000
   o
   c;
   I  4000
       2000 -
             0 HALF-LIVES
                100*
                                  19E
                                    'AuHALF-ilFL =2.695 DAYS
                         ! HALF-tlFE
                                    2HALF1IVES
                                        25*
                                                             5 HALF-LIVES
                                           3 HALF-LIVES           3.13*
                                             12.5*    4 HALF-LIVES
                                                         6.25%
               0    1    23    45    6   78   9   10  11   12   13  14  15
                                       TIME (days)
                                198
         FIGURE1.3.   Number of    Au  Atoms  Present  as  e Function of
                      Half-Lives Elapsed

where            N =  the number of radioactive  atoms  present  at time t
                N  =  the number of radioactive  atoms  originally present
                 e =  the base of the natural  logarithms  (2.71828)
                 >. =  the decay constant of the  given  radionucl ide
                     = (In 2)/t./2 = 0.693/t./2
                 t =  the time elapsed.
     The halr-life end decay constant are inversely  related.  A radionuclide
with a long half-life has e  small decay constant; a  radionuclide with  a short
half-life has  a  relatively large decay constant.
     E-  Energy  of De::y.   Tne unit of energy  used  in  radioactive  decay is
the electron volt (eV).   The electron volt,  which is the energy acquired by an
electron when  it  falls through e potential  difference  of 1  volt, is equal  to
                                      i.li

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DARCO.M-P 3S5-1
              -19
about 1.6 x 10    joules.   Multiples  of the electron  volt,  such  as  the  kilo-
electron volt (keV) and the millionelectron volt (MeV),  are  also  used.   One  keV
equals 1000 eV and 1 MeV equals 1,000,000 eV.
     The energy of radioactive decay  is observed as  the  kinetic  energy  of par-
ticulete radietion, electromagnetic radiation,  or as  both.
     (1)  Kinetic Energy of Particles.   Radioactive  decay  can  change the state
of an atom's nucleus through the  emission of particles from the  nucleus.  These
particles have kinetic energy, or energy of motion.   The kinetic energy (T)  of
a particle is a function of its mass  (m) end its velocity  (v).   According to
classical physics,

                T = j mv2                                             (1.3)

From this equation we learn that, if  two particles have  the same velocity,
their kinetic energies are related by a simple ratio of  their  masses.  Con-
versely, two particles of equal kinetic energy have velocities that are related
to the square root of their masses.  That is, a light particle has a velocity
greater than that of a heavy particle of equal  kinetic  energy.
     Equation (1.3) is valid if the velocity of the particle is not comparable
to the velocity of light.  When the speed of the particle becomes faster than
one-tenth the speed of light, the mass of the particle  increases, and the
equation cannot be used.  Particles traveling at velocities comparable  to the
speed of light are said to be traveling at relativistic velocities, and the
equation relating the kinetic energy of a particle and  its velocity is  then
                       .2
T • m0c
- 1
                                                                         (1.4)
where           T » the kinetic energy of the particle
               m0 * the mass of the particle
                c » the velocity of light
                5 • v/c
                v * the velocity of the particle.
                                     1.12

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                                                              DARCOM-P 385-1
     An important conseouence of this equation is that no particle, whatever
its energy, can travel  faster than the speed of light  in a vacuum.   Figure l.£
illustrates the energy-velocity relationship for alphe. and bete  particles.
     (2')  Electrom??net'c Energy.   The energy released from a decaying rariio-
nuclide can also take the form of  oscillating (vibrating) electric  and magnetic
fields, or electromagnetic radiation.  This radiation  travels in the forrr, of
waves that have a characteristic frequency, u, and wavelength, )..   The frequency
of electromagnetic radiation is expressed in terms r*  cycles per second,  or
hertz (Hz).  The wavelengths of various electromagnetic radiations  ere expressed
in units of measure appropriate to their length.  For  example, wavelengths of
ultraviolet radiation are measured 1n nonometers or meters, whereas radio waves
are measured in centimeters or meters.  All electromagnetic radiations travel
                                               o
at the velocity of light, which is about 3 x 10  meters per second  (m/sec) in
a vacuum.   'The wavelength times the frequency is equal to the the  velocity of
light.  The electromagnetic wave spectrum consists of  wavelengths  ranging from
       100R  -
                   1.2       3        4       5       6       7
                                  PARTICLE ENERGY (MeV)
            FIGUR:.1.4.   Energy-Velocity Relationships for Alpha and
                         Beta  Particles
                                     1.13

-------
DARCOM-P 385-1
                                                        ,-9
several  kilometers to e  small  fraction  of a  nanometer  (10   m).   Between
these limits lies a continuous range of electromagnetic waves.
     Figure 1.5 shows the electromagnetic spectrum.  This  spectrum is  divided
into 2 number of regions, each representing  wavelength intervals  within which
there is a common state-of-the-art in radiation sources  and detectors.  All of
these regions overlap; thct is, the characteristics  of the radiation change
slowly with the change in frequency, and it  is difficult  to know  exactly where
one region ends and the next begins.  Examples of electromagnetic radiation
include radio waves and microwaves, infrared and visible  light,  and x and
gamma radiation.
     Electromagnetic  radiation exists as waves; however,  wr.en discussing the
enerc_\ of electromagnetic radiation, it  is often convenient to think of the
WAVELENC
(meter
icf18 -i
ID'14 -
10'10 -
-A
10 6 H
itf2 -
io2 -
io6 -
;TH FREQL
>) (he
r 3xio26 -
- 3xl022 '

3x10
, irJ4
— 3x10
~ 10
- 3xlOiU -
- 3x1 06 -
- 3xl02 -
1ENCY ENE1
rtz) (e\
"2 1.24xl012~
~ 1.24xl08 '


*
— 1 . 24 -
- 1.24xlO~4 -
- 1.24xlC'8 -
I1.24xlO~12'
*GY
/)

i
GAMMA COSMIC
f
X-RAY 5
-ULTRAVIOLET ^ f

* "^VISIBLE
_ INFRARED
T i
MICROWAVES
RADIO
TV
f
4 TV
SHORTWAVE
T RADIO
—
I ELECTRIC POWER
                    FIGURE  1.5.   :he  Electromagnetic  Spectrum
                                      1.14

-------
                                                              DARCOX-F 385-1
waves as existing in the forrr. of wave packets., celled quanta or photons.   At
high energies, these wave Dockets or photons beneve cr. if they were smell
particles.  This phenomenon of electromagnetic radiation acting like particles
or particles acting like weves is called the wave-particle duel'ty of
electromagnetic rsdiatio*,.
     The energy of a single photon or quantum is related to the frequency of
the radiation and ranges from very small values at low frequencies to very
large values at high frequencies.  That is,

                E = hu                                                (1.5)

where           E = the energy associated with a photon of electromagnetic
                    radiation
                h = Planck's constant (4.136 x 10    eV-sec)
                u = the frequency.
     C.  Type of Radiation Emitted.   When particulate or electromagnetic radi-
ation has an energy greater than about 30 eV, it is able to strip an electron
from a molecule in a process celled ionization.  Photon energy is sufficient
to cause ionizetion at frequencies greater than that of light.  Radiation with
this high an energy level is called ionizing radiation.  The most important
characteristic of ionizing radiation is its localized release of large amounts
of energy, approximately 32 eV per ionizing event.  This amount of energy is
more than enough to break a strong chemical bond; for example, the energy
associated with a C=C bond, commonly found in body tissues, is 4.9 eV. The
ability to break chemical bonds makes ionizing radiation of concern because it
can disrupt the function of living cells.
     Radioactive decay results in five tyoes of ionizing radiation:  alpha
particles, beta particles, gamma rays, x rays, and neutrons.  These types of
radiation can be distinguished by their physical characteristics, such as
mess,  electrical  charge, and path length or range, as shown in Table 1.1.  The
two mejc*- modes of  decay result in the emission of alpha particles or bete
particles.   Both  of  these decay modes can also be accompanied by the emission
of gamma  rays.   The  five types of ionizing radiation and tne tyoes of decay
that  produce them are described below.
                                     1.15

-------
 DARCQ.M-P 385-1
                     r^SLt 1.1.   Radiation  Characteristics
                                         /  *             Path Lenoth
n .
r\a
Alpha

Beta
Gamma
diati
part

parti
rays
on
ides

cles
and
Mass
6.64 x

9.11 x
• — ••
(9
10

10

)
-24
.97


Chargevc '
-2

-1
0

5-

0-
0.

10

18
1-
Air
cm

m
100 nT(b;
Sol id
25-40 urn

0-1 cm
1 mm-1 nT




(b)
   x rays
 Neutrons
1.67 x 10
                            -24
0-100 m
0-100 cm
 (a) Tne unit charge is approximately 1.6  x  10"    coulombs.
 (b) There is no real  endpoint for  electromagnetic  radiation;  however,  its
     intensity is reduced as it travels  farther  and  passes  through materials.

     (1)  Alpha Particles.   Alpha particles  are  emitted  only  from  very  heavy
nuclei that have an atomic number,  2, of 82  or more,  except in some  artifi-
cially produced nuclides.  An alpha particle (a) is  a helium  nucleus.   It  has
two protons and two neutrons and a  net charge of +2.   When  a  parent  nucleus
decays by alpha emission, the atomic number  of the daughter nucleus  is  two
less than that of the parent, and the mass number,  A, of the  daughter nucleus
is four less than that of the parent.  This  reaction is  summarized  in
Table 1.2.
     (2)  Bete Particles.  Beta particles  result when a  proton is  converted  to
a neutron or a neutron is converted to a proton  in the nucleus.  These  transi-
tions help an unstable nucleus establish a more  favorable neutron-proton
ratio.  After such a transition, two types of particles  are ejected  from the
nucleus:  a neutrino and a beta particle.  A neutrino, symbolized  u, has no
charge and essentially no mass and  travels at the velocity of light. The
neutrino does not easily interact with matter and presents no radiation hazard.
A beta particle can have either positive or  negative charge,  depending  on the
type of transition in the nucleus.   If the beta  particle is negatively  charged,
it is an electron; if positively charged,  it is  a positron.
     When the nucleus has an excess number of neutrons,  it undergoes a  neutron-
to-proton transition, and a negatively charged beta particle, or electron, is
                                     1.16

-------
   Decay Type
                                                      I-ARCOX-P 385-1
TABLE 1.2.  Effect of Common Decay Types  on the  Parent Nucleus

           Che-"'e frorr Parent to Deuohter Nucleus

                                                     Reaction Suircna-v
 Alpha
 Bets  negative
 (electron)
 Beta  positive
 (positron)
 Electron  capture
Atomic
Numoer. Z
-2
-1
-1
-1
Neutron
Number, l\
-2
-1
+ 1
+ 1
Mass
Number. A
-4
No change
No change
No change
                                                            Av
                                                                     .
                                                                     n,


                                                               A .,
                                                            + 7 , X - nu
                                                              L ™ 4.
                                        No  change    Jx  + e"  - 7A,X + hu
                                                    Z         Z-i
 (a)   hv  * energy of protons  (see  Equation  (1.5)).

 ejected.  As shown in Table  1.2,  this beta negative decay results  in  the atomic
 number,  Z, of the nucleus  increasing by one.  The mass number, A,  remains
 constant.
      Proton-to-neutron transition occurs when the nucleus has an excess number
 of protons.  In this case, a positively charged beta particle, or  positron,  is
 ejected  in what is called  bets positive decay or positron decay.   As  a  result
 of this  decay, the atomic  number  Z decreases by one while the mass  number  A
 remains  constant.
      Sometimes a nucleus has an excess number of protons but is unable  to  emit
 a positron.  In this case, the nucleus captures an orbiting electron, which
 combines with e proton to  forrr a neutron.   This process is called  elect-on
 capture decay,  and the resulting nuclear change is identical to that  of posi-
 tron emission:   the atomic number decreases by one and the mess number  remains
 constant.  Because an electron has been removed from its orbit, x  rcys  are
 produced as the electrons  become rearranged (see Section (4) below).
     Beta particle:  are emitte: from the nucleus with a spectrum of energies.
The beta particle and the  neutrino are emitted together and share  a given
                                     1.17

-------
DARCOM-P 365-1
amount of energy, but the sharing is  not in a  constant  ratio.   The  beta  par-
ticle may tnerofore be ejected from the nucleus  with  essentially  no kinetic
energy, or with a nigr, kinetic energy.   The average energy  of  the beta  parti-
cles emitted is aocut one-third of the  highest kinetic  energy  for beta
particles.  Tables of beta energies indicate the highest  energy level  for
betas; but only a small fraction of beta particles  possess  this highest  energy
level.
     (3)  Gamma Rays.  When radioactive decay  results in  the  emission  of a
particle from tne nucleus, the nucleus  is often  left  in an  excited  state.  The
excited nucleus then releases its excess energy  in  the  form of gamma rays
(photons, or wave packets of electromagnetic radiation) until  the ground energy
state of the nucleus has been reached.   Sometimes the energy  is emitted in one
jump; at other times it is emitted in a series of jumps.   The  number and energy
of gamma rays given off following particle emission is  characteristic  of a
given radionuclide.
     Gamma rays are usually emitted immediately after the particle is  ejected,
but sometimes the nucleus remains in a  high-energy state  for a measurable
period of time, up to several hours.  The excited nucleus is  then  in an
unstable, transient condition and is called an isomer of  the ground-state
nucleus.  Isomers are nuclei that are identical  to each other in ell respects
except for their energy state   The excited state is designated by writing "m"
af-
                                                   9Sir-
  ter the mass number of the  nuclide;  for example,  "  "Tc  is  an  isomer of
                            oc
technetium-99 and decays to "Tc by releasing a gamma  ray.
     (4)   X Rays.  The capture of en orbital  electron  by  a  nucleus with excess
protons (electron capture decay) results in a vacancy  in  the shell that the
electron  occupied.   The shell  most commonly vacated is the  K shell, that
closest to the nucleus.  Because an electron  from an  outer  shell  jumps down to
fi'il  the  vacancy, electron capture is  always  accompanied  by the emission of
characteristic radiation in the form of x rays.  Like  gamma rays, x rays are
photons,  or quanta of electromagnetic  radiation.
                                      1.18

-------
                                                              DARCOM-P 385-1
     (5)  Neutrons .   Neutrons ?re not emitted from the ripre common rpdio-
nuclides.  Some of the heavier radionucl ides enit neutrons by spontaneous
fission, or splitting of the nucleus.  The most common example of e spon-
-                                               252
tanecusly fissioning radionucl ide is cal ifornium-252 (   Cf).  Other sources
of neutrons ere listed below.
 1.  Some isotopes of boron, beryllium, lithium, sodium,  fluorine, and other
     elements with e low atomic number emit neutrons when irradiated by alpha
                 t
     particles or qemne rays.  These neutron sources are  prepaid by mixing a
     radioactive nuclide and a finely Divided powder of the target substance.
                                                       24 1
     Examples of neutron sources are the  mixed powders   ' Am:Be (americium
     and beryllium) and    PorBe (polonium and beryllium), and the chemical
              23°
     compound   "'PuF. (plutoniun fluoride).  Neutron sources are kept in
     seeled metal containers, and the neutrons emitted have a spectrum of
     energies.
 2.  When high-speed charged particles irradiate a suitable target material,
     the resulting nuclear reactions yield neutrons.  These high-speed
     particles, or accelerator sources, can be used to produce neutrons of
     nearly the same energy.
 2.  The fission process in nuclear reactors produces large numbers of neutrons
     with a spectrum of energies.
1.2.2  Decay Pathways
     A radionucl ide can undergo radioactive decay vie more then one decay
path way.  Each decay pathway consists of the emission of a particle followed,
in most cases, by the emission of one or more gamma rays.  Pathways differ in
the manner in which the energy of decay is distributed among the particle
                                                                 o o *
emitted end the subsequent gamma rays.  For example, radium-226 (  cRe) can
decay by five pathways.  The most common pathway is the emission of an alpha
particle with £.78 MeV of kinetic energy; the resulting (dauahter) nucleus,
           222
radon-222 (   Rn), is not in an excited state, so no gamma is emitted.  The
next most common pathway is the emission of an alpha particle wiTh s kinetic
                         2?'
energy of 4.60 MeV.   The  "Rn daughter nucleus is in an excited state, and a
gamma ray is emitted.  For three additional pathways with alpha energies of
                                       10

-------
DARCOM-P 385-1
4.34, 4.19, and 4.16 MeV,  the emission  of  gamma  rays follows.  A single nucleus
can decay by only one of the various  pathways, but because there are five
                                              OOC
potential patnways,  it is  sometimes  said  that    Re has five alphas, or five
potential elpna energies.   Appendix  A contains a more detailed discussion  of
oecay pathways.
1.2.3  Quantification of Radioactivity
     Radioactive materials are not always  measured by their ness or the number
of atoms present.  They are usually  measurec  by  the number of nuclear  decays  or
disintegrations occurring  in a sample at  any  time.  The number of  disintegra-
tions occurring in a sample per unit time  is  the activity of the sample.   The
traditional unit of activity is the  curie, aboreviated Ci.  One curie  is  the
amount of material undergoing 3.7 x  10   disintegrations per second  (dps).
Severs! fractions of the curie are in common  usage:   the microcurie, abbrevia-
ted yd, is one millionth  of a curie (3.7  x 10  dps), and the picocurie,
                             o
abbreviated pCi, is 3.7 x  10   dps or 2.22 disintegrations per minute  (dpm).
The international system (SI) unit of activity,  the becouerel, abbreviated Bq,
is 1 dps.
     A radionuclide's activity, A, is related to its  decay constant, X,  and the
number of radioactive atoms present, N, by the equation  A *  AN.   Remember that
X » (In 2)/ty2-  From th-is equation, we  learn that for  a given  sample activity,
fewer radioactive atoms are present if the half-life  is  snort than if  the half-
life is long.
     The activity represents the disintegration  rate  of  the  sample;  for every
disintegration, one or more radiations may be emitted.   As  a  result,  two samples
of equal activity may emit different amounts  of  radiation.   For  example,  each
disintegration of cobalt-60 (Co) Involves the  emission  of  one  electron
                                                       3       14
followed by two gammas, whereas each disintegration  of   H  and    C Involves
the emission of only one electron, without gammas.
     The activity of a radioactive sample 1s  directly related  to the number of
radioactive stems present.  For this reason,  the activity of the sample
decreases exponentially as the number of radioactive  atoms  present decreases.
That is, the activity of a sample of a radlonuclide  can  be  determined  at any
time'using the following equation:
                                      ..20

-------
                                                              DAKCOM-F 3S5-1
               "o
where           A =  the  activity present  at  time  t
                  =  tne  activity originally  present
                e =  the  base of the  natural  logarithms  (2.71826)
                ), =  the  decay constant of the  radionucliae  =  (In  2)/t, .r
                ,    =  0.693/t1/2
                t =  the  elapsed time.
     The specific activity is defined  as  the activity of  1  gram of  radioactive
material and is usually  expressed as Ci/g of the  materiel.  The specific
activity of a radionuclide is inversely proportional to its half-life;  that is,
a radionuclide that  has  a short half-life will  have  a higher  specific activity
thsn a radionuclide  that hes a long  half-life.
             Section 1.3  INTERACTIONS OF RADIATION WITH MATTER

     All radiation, whether participate or electromagnetic,  possesses energy.
The reduction of this energy, or of the radiation's Intensity, as It passes
through some matter 1s called attenuation.   Attenuation 1s  a combination of
two processes, absorption end scattering.  Absorption involves the dissipation
of the radiation energy Into the absorbing medium;  scattering involves the
deflection of the radiation from its original  path.  The mechanisms of radia-
tion attenuation are described in this section.
1.3.1  Alpha and Bete Particles
     The transfer of energy from radiation to the atoms of an absorbing
material can occur by several processes.   Alpha  and beta particles transfer
energy primarily by the absorption processes of  excitation and ionization.
     A.  Energy Transfer Processes.  Excitation  is  the raisin? of an electron
in an £tom or molecule of the absorbing materiel to a nighe* energy 'eve!
without the electron being ejected from the atom or molecule.  The electron
                                     1.21

-------
DARCOM-P 385-1
then returns to its original  energy  state,  at  the  same  time  releasing  electro-
magnetic radiation in the form of light  or  x rays.
     lonizr.tion involves the  transfer of sufficient  energy  to  an  electron  to
remove it from the electronic structure  of  the atom  or  molecule.   Depending on
the degree of the interaction, the ejected  electron  may possess  anywhere  from a
negligible up to a very large amount of  kinetic energy.   If  the  electron  is
given sufficient kinetic energy as it is ejected,  it may cause excitation  and
ion-.zation in other atoms of  the absorbing  materiel  it  is passing through; it
is then termed a delta ray.   The isolated electron and  the  remaining atom
together are called an ion pair.  The average  number of ion  pairs formed  by
radiation per unit length of  the matter  it  passes  through is called the
specific ionizetion of the radiation.
     As alpha and beta radiations move through an  absorbing medium and their
energy of motion is transferred to the orbiting electrons of the absorbing
medium by excitation and ionization, the alpha and beta particles gradually
lose all kinetic energy until virtually  no energy  is left.   The  rate of energy
loss as the radiation traverses a material  is  called the linear  energy transfer
(LET) of the radiation and is measured in joules per meter (J/m).  (Histori-
cally, LET has also been expressed in terms of keV/um.)  In general, the higher
the LET of the radiation, the shorter its range (the distance it travels) and
the greater the biological hazard it presents  because all its energy is
deposited over e smaller volume of tissue.
     E.  Alpha Particle Interactions.  An alpha particle is emitted from the
nucleus cr a rcdioactive atom with a velocity about one-twentieth that of
light.  Because of its low velocity and  double positive charge,   the alpha
particle interacts readily with atomic electrons by excitation and  ionization,
and has a very high specific   ionization  and LET.  The alpha particle loses
kinetic energy very rapidly,   so it has a low  penetrating ability and travels
only a few centimeters in air.  (Refer back to Table 1.1.)  An alpha particle
can usually be stopped by s.everal sheets of paper or a  sheet  of  aluminum  foil.
After an alpha particle loses all of its energy,  it attracts  two electrons and
becomes a he!ium atom.
                                     1.22

-------
                                                               DARCOM-F 385-1
     The range of en elpha particle in f'ssue is 35 to 70 uf., Depending on its
original energy.  Because -this range is about tne same i~ the thickness of the
dead skin  layers on the human booy, an elphc-enitting racionuclide is con-
sidered to present little hazard c^tside of tne body.  There are a few excep-
tions to this general rule.  First, if the skin is broken, living tissue may
be  irradiated.  Second, in the case of the eye, the living tissues ere very
close to the surface end can be harmed by alpha radiation.
     The greatest' biological hazard due to alphe-eritting radionuclides occurs
when the material enters the body by inhalation or ingestion.  In this case,
there e*"e no dead cells to absorb the energy, end living tissue is irradiated.
     C.  Beta Particle Interactions.  Beta particles are emitted from the
nucleus with a velocity much greater than that of alpha particles, even
approaching the velocity of light.  Beta particles are more penetrating than
alpha particles and can travel up to 18 meters in air, depending on their
energy; however, they can be stopped by a few millimeters of materials such as
plastic, aluminum, and iron.
     Beta particles lose their energy primarily by interacting with the elec-
trons of the absorbing medium.  Bete particles can elso slow down in the elec-
trical field of atomic nuclei to produce x rays.  The x rays produced in these
interactions ere called bremsstrehlung (from the German word for "braking," so
named because this radietion results from the slowing down of bete particles).
The energy of the bremsstrahlung mey range from negligible up to the energy of
the bete particle.  The probability of this interaction occurring is greater
for radionuclides that emit high-energy bete particles, such as phosphorus-32
( "P) end yttrium-90 (  Y), and for absorbing materials with a high atomic
number, such as iron or lead, than for redionuclides that emit beta perticles
with lower energies and for ebsorbing materials with a low atomic number.  The
radiation produced is identical in ell  respects to gamma or x radietion of the
same energy.   Bremsstrahlung photons can present a significant radietion
hazard when radionuclides thet emit high-energy beta particles ere stored in
Retellic containers.   In orde>- to reduce the production of b'-ensstrenlung,
emitters of high-energy bete ^articles should be kept in thick-walled plastic
containers.   Tne plestic containers may then oe placed in iron or lead

-------
DARCQM-P 385-1
containers to protect against any photons  other than  bremsstrchlung  that  mey              ^
be emitted.  Bremsstr?hlunc is not produced in  any  sic;.ificant  amount  in  bio-
logical materials because the element:  of  which human tissues are  composed
have low atomic numbers.
     The LET of beta particles is much  lower than that of  alpha particles
became betas have only a single charge and travel  at high velocities.   In
many cases, beta radiation is considered to be  only a slight hazard  outside
the body, because even though betes with an energy  higher  then  70  keV  can pene-
trate to living skin tissue, they still cannot  reach the major  organs  of  the
body.  However, beta particles can cause severe damage to  the skin and the
eye. Thus we can say that beta particles present more of an external hazard
than do alpha particles.
     Inside the body, beta radiation is less hazardous than elpha  radiation.
Because the LET of beta particles is less than  that of alphe particles, the
energy deposited by the beta radiation  is dissipated over a larger volume than
is the1 energy deposited by alpha radiation.
     After a negatively charged beta particle (an electron) loses  all  of its
kinetic energy, it becomes attached to a positive Ion, becoming an orbital
electron.  A postlvely charged beta particle (or positron), on  the other hand,
1s antimatter and cannot exist for long 1n nature,   After 1: loses all of its
kinetic energy, 1t fuses (coalesces) with an electron, the ".wo  particles annihi-
late each other, and their mass 1s converted into energy,   "his energy 1s
observed as two photons, called annihilation radiation, each of which his
0.511 MeV of energy.  The two photons are emitted in opposite directions.
1.3.2  Photons
     Gamma rays and x rays are both forms of electromagnetic radiation and they
have identical properties.  The only difference between then 1s that gamma rays
are emitted from the nucleus and x rays arise from processes outside the
nucleus.  X rays oroduced as a result of radioactive decay tend to  have  lower
energies than gamma rays, while x rays produced by x-ray machines can have
energies much higher than the energies of gamma rays.
                                     1.24

-------
                                                              DARCOM-P 385-1
     A.   Energy Trer.sfer Processes.   Ionizing photons "interact with metier by
th-ee mejor nechanisr.s:   t^e photoelectric effect,  tne Comrton effect, and
pair production.   Which  interaction  takes place depends on the photon energy
and on the atomic number,  Z, of the  absorbing medium.  Figure 1.6 shews the
relative importance of thes? interactions es a function of Z end photon energy.
The end result of all  three interactions i; the procuction of high-energy
electrons, which  interact  with matter in the same way beta particles do.
     The photoelectric effort is en  interaction between a photon and an
orbital  electron.  In  this process,  the photon ceases to exist and its energy
is transferred to the  electron, whirh is ejected from the atom with a kinetic
energy equal  to the energy of the photon minus the binding energy of the
electron.   The photoelectric interaction is dependent on the energy of the
photon and strongly dependent on the atomic number,of the absorbing material.
It is most likely to occur in high-Z materials, such as iron and lead, and at
low photon energies, less  than 100 keV (0.1 MeV).  The photoelectric effect is
not an important  interaction in biological systems,  which are made up primarily
                  PHOTOELECTRIC EFFECT
                      DOMINANT
PAIR PRODUCTION
   DOMINANT
                                   COMP10N EFFECT
                                     DOMINANT
                                         1
                                PHOTON ENERGY (MeV)
 10
100
         FIGURE  1.6.   Relative  Importance of the Photoelectric Effect,
                      the  Compton  Effect, and Pair Production
                                     1.25

-------
DARCON-? 385-1
of carbon, oxygen, hydrogen, and nitrogen,  all  low-Z elements.   However,  it  is
important in high-Z materials and is  useful  for  identifying and  Quantifying
gamma-emitting radionuclides.
     The Compton effect (or Compton scattering)  is  the  predominant  interaction
between biological materials and photons  from 30 keV  (O.Oi MeV}  to  10 MeV.   In
the Compton interaction,  the photon interacts with  an orbital  electron that  is
not tightly bound to the  nucleus.  The photon transfers part  of  its energy  to
the electron, which is ejected from the atom.  The  photon  is  then scattered  by
(deflected from) the eton at reduced energy.  The scattered photon  can go on to
interact with electrons of other atoms.
     High-energy photons can interact with the electrical  field  of  the atom's
nucleus via pair production.  In this process, when a  photon  passes close to
the nucleus of an atom, the  photon ceases to exist, .and 1.02  MeV of energy is
converted into an electron  (negatron) and a positron.   If  the original photon
had an energy greater than  1.02 MeV, the remaining energy  is  shared by the
electron and the positron in the form of kinetic energy.  This interaction,
which does not occur if the  original photon energy is  less than  1.02 MeV, is
of greatest importance in high-Z materials and does not often occur in bio-
logical tissue.
     B.  Photon  Interactions.   Photon-emitting radionuclides  outside the body
can present a severe hazard  for  several reasons.  First, photons can penetrate
through thick layers of lead and concrete,  so it is difficult to shield  the
body against them.  Second,  they can penetrate great distances through air and
may therefore constitute a  hazard  even far  from a source of radiation. Finally,
photons can easily penetrate the skin  and can irradiete organs within the body;
in fact, they can irradiate  the  whole  body.  However,  photons ere  less of an
internal hazard  than either alpha  or beta radiation because they have a  low
LET and distribute their energy  throughout  the body rather than  concentrating
it in one small  area.
1.3.3  Neutrons
     Neutrons,  like photons, are very  penetrating.  Because  they have no elec-
trical charge,  neutrons, unlike  other  types  of  radiation, do  ret interact with
electrons.  They do interact with  atomic nuclei, yielding charged  particles
                                     1.26

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                                                              DARCO>5-P 385-1
that can then deposit energy in er  absorbing medium by excitation  and  ionize-
tion.   Neutrons are not  stable outride  the  nucleus.  They  r.ave a half-life  of
10.6 min and decay to a  proton and  an electron.
     Neutrons car be classified by  their  energies; one such  classification
scheme is shown ir Table 1.?.   All  neutrons are  fast neutrons when produced.
Neutrons that have lost  most of their energy are called  thermal neutrons.
Thermal  neutrons receive their name from  the fact that they  are in approximate
thermal  equilibrium with their environment.
     A.   Energy TrePS'er Processes. To a large  extent,  the  type of interac-
tion that a neutron undergoes  depends on  its energy.  Most fest neutrons  lose
their energy by colliding with nuclei in  whet  art termed elastic collisions or
"billiard ball" collisions.   For neutrons with energies  between 100 keV  and
20 MeV,  this is the predominant interaction with biological  materials.   When
incident neutrons collide with the  nucleus  of  an atom of the absorbing mate-
rial,  pert of the neutron's  kinetic energy is  transferred to the nucleus  and
part is  retained by the  deflected neutron,  which may then undergo  additional
collisions until it has  lost virtually  all  of  its energy.
     Fest neutrons may also  lose their  energy  by inelastic scatter.  In  this
process, a neutron transfers part of its  kinetic energy  to the  nucleus of an
etom.   The nucleus is then in  an excited  state and emits a gamma  r?y to  return
to its ground state.  Inelastic scatter is  e phenomenon  more closely associ-
ated with high-2 absorbers,  such a?, iron  or lead, than with  low-2  absorbers,
such as  hydrogen or carbon.

                    TABLE 1.3.  Classification of Neutrons

              Neutron Classification                Enerov
              Thermal  neutrons                _<0.025 eV
              Slow neutrons                  0.025 eV to 100 eV
              Intermediate neutrons           100 eV to 10 keV
              Fast neut-ons                  >10 keV
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DARCOM-P 385-1
     P neutron ma> enter the nucleus  of an atom and  undergo  radiative  capture.
The resultant nucleus, one mass unit  heavier than the  original,  is  in  an
excited state and emits a gamma ray.   Because the gamma  rays  arising fron this
type of interaction nay have energies up to several  MeV,  they contribute  to
the shielding difficulties encountered with neutrons.
     In radiative capture with particle emission, a  neutron  nay  be  captured by
a nucleus that subsequently ejects a  charged particle, for example, a  proton or
an alpha particle.,  This interaction  is used to infer the presence  of  neutrons
and to produce radioactive isotopes.
     The capture of a neutron by certain heavy nuclei  may result in fission,
the splitting of the nucleus into two lighter nuclei of approximately  equal
mass, called fission fragments.  As the nucleus disintegrates, an average of
two or three neutrons is emitted.  If one of these causes a  subsequent fission,
a steady-state chain reaction may take place.  Some nuclei undergo fission
after absorbing a thermal neutron, others after absorbing a  fast neutron.
Fission fragments are radioactive and present a radiation hazard of their own.
     B.  Neutron Interactions.   In soft tissues of the body, the predominant
interaction  is collisions between incident neutrons and hydrogen nuclei,  which
are single protons.  This interaction is  important because a large fraction of
the neutron  energy is transferred to the  proton, since its mass is almost the
same as that of the neutron.  Furthermore, hydrogen is the most abundant atom
in the tissues.  The protons that are set  into motion by this process lose
energy by the excitation and ionization of atoms as they pass through biolog-
ical material.  These protons have a high  LET and can produce significant
biological damage.
     At kinetic energies below a few hundred keV, radiative  capture of neu-
trons becomes important.  The capturing nuclei are  primarily those  of hydrogen
and nitrogen.  Neutron capture by hvdrogen,  H,  results  in the emission  of  a
                                          1                            2
2.2-MeV gamma ray; at the same tine, the   H  nucleus is converted to a  H
nucleus.  Neutron capture by a   N nucleus  leads to the  emission of a 660-keV
                                     14               14
proton and the transformation of the   N  nucleus to a    C nucleus.  The  prob-
ability of neutron capture by other  elements  in  the body  is  small.
                                     1.28

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                                                              DARCO.M-? 385-1
      If  the  body  is exposed  to e high concentration of neutron? , two reactions
 occur the: can  be  used  to  estimate the neutron exposure:  sediur in the tissues
                             n *s      o f
 end  blood  is  converted  from  "f.'a to " N'e, and sulfur in the heir chances from
 ??     70
  S  to  P.   The  radiation dose from these radioective nuclides is smell com-
 pared to the  radiation  dose  received fro^ the large number of neutrons required
 to act'vete  en  appreciable number of teroet atoms.
                  Section  1.4  RADIATION QUANTITIES 0ND UNITS

     Radiation measurements and units of radiation dose are based primarily on
 the energy  deposited by  radiation as it travels through matter.  The Interna-
 tional  Commission on Radiation Units and Measurements (3CRU) selects and defines
 the units and quantities of radiation.  Information provided in this section is
 b?sed on  ICRU Report 33, Radiation Quantities end Units (ICRU 1980).
 !.*.!   Exposure
     The  term "exposure" has two levels of meaning.  The first level is that.
 of an object or person being subjected to the action of radiation.  It is in
 this context that the word is most commonly employed, especially by the public.
 "or example, a person might say "I was exposed to neutrons."  In radiation
 protection, on the other hand, the term exposure .is used to quantify the emount
 of x or earns radiation present.   In a given situation, the meaning cf the
 word is determined from the context in which it is used.
     In the context of radiation protection, exposure is a measure o- the ior.i-
 zation produced by x or gamma radiation in sir.  The ionization is measured by
 collecting all  the electrons liberated by the photons through photoelectric,
 Compton, and pair production interactions.   Note that exposure, in this sense
 of the word, is defined only for x and gamma radiation, and that the measure-
ment must be made in air.  In practice, exposure is difficult to measure pre-
 cisely when the photon energies involved are below a few keV or above a few
MeV.
                                     1  70

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DARCOM-? 385-1
  -  The special unit for exposure is  the  roentgen  (R).   One  roentgen  is  equal
to 2.58 x 10"  coulomb/kg of 'air.   This  seemingly arbitrary value  is equiva-
lent to 1 electrostatic unit of electricity (esu) per  cubic centimeter of air
at standard temperature and pressure (STP), which was  the original  definition
                                                                    9
of the roentgen.  One roentgen results in  the production  of 2.08 x  10   ion
pairs/cc of dry air at STP.  The energy  required to produce these  ion  pairs is
approximately 87.7 ergs/g of air.
1.4.2  Absorbed Dose
     The absorbed dose describes the Quantity of radiation energy  transferred
to any absorbing material  (tissue, air,  shielding,  etc.).  The ICRU has defined
absorbed dose, symbolized D, as
where dc is the mean energy imparted by ionizing radiation to matter of mass
dm.  The advantage of absorbed dose as a measure, as compared with exposure,
is that absorbed dose can be applied to any radiation and any absorbing medium.
The unit of absorbed dose is called the rad and is equal to 100 ergs/g of the
absorbing material.  In the international system of units (the SI unit), the
absorbed dose is the gray (Gy) and is equal to 1 joule/kg.

            1 rad = 100 ergs/g = 10"2 J/kg = 0.01 Gy

     For x and gamma rays, the exposure (expressed in units of roentgens) can
be related to the absorbed dose in tissue (expressed in rad) by the equation

          Dtissue s °'97 X                                             (1'8>

where X is the exposure in roentgens.  This equation holds for x  or gamma
radiation of energies from 0'. 1 to  10 MeV.  From this equation, the absorbed
dose, in rad, to an individual exposed to x or gamma radiation can be  deter-
mined by measuring the exposure, in roentgens, at the location where  the
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                                                              DARCOM-P 385-1
 individual was exposed.   Potential  radiation doses to individuals working with
 sources of x or gamrs  rachetion can also be estimated, as will be discussed in
 subsequent chapters.
 i.4.3  Relative Biological Effectiveness
     Because radiations  interact with matter in varying ways, equal doses of
 different types of  radiation do not always produce equal biological effects.
 Men comparing the  effects of different radiations, it is customary to use
 250-kV >; rays ;.s the standard.  This radiation was cnosen as a standard because
 its effects were well  documented and it was the only type of radiation widely
 available at the time  this convention was adapted.
     The formal Definition of relative biological effectiveness (RBE) is as
 follows:  the RBE of a test radiation is defined by the ratio Dcn/D , where
      is the absorbed dose of 250-kV x rays and D  is the absorbed dose of
 the test radiation required to produce an equal biological  effect.
      The RBE TS often used in radiation biology, but the concept has limited
 usefulness in radiation protection because the RBE of a given radiation is
 influenced by the specific conditions of the experiment.  The dose rate used,
 the dose fractionation (or the number of increments in which the dose is
 received), the biological tissue irradiated, and the radiation effect measured
 all effect the RBE.
 1.4.£  Dose Equivalent
      The results of biological experiments have shown that the absorbed dose by
 itself is insufficient for predicting either the probability of deleterious
 heeltr, effects from irradiation under unspecified conditions, or the severity
 of such effects.  The RBE of radiation is clso not useful,  primarily because of
 the many factors that car. influence it.   Consequently, an additional quantity
 hes been defined.
     This quantity, a quality factor, Q, accounts for the different biological
effects tnat  result from the ways various types of radiation distribute eneroy.
Tne \elues  of  Q are a'e-'ined es e  -unction of the radiation's LET in water and
are based on  relevant values of RBE.   Table 1.4 shows the recommended values
                                     1.31

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DARCOM-P 385-1
              TABLE 1.4.   Relationship  of  LET  and Quality Factor

                          LET (keV/um)           Q
                               ^_3.5              1
                                7               2
                               23               5
                               £3               10
                              175               20

of Q as a function of LET in water.   It is possible  to  find  exact  quality
factors based on the LET by interpolating  the  values  given  in  the  table.
However, it is common practice to use the  recommended values for  different
types of radiation, es given in Table 1.5.
     The absorbed dose and the quality  factor  are  incorporated into a third
quantity, called the dose equivalent.  The dose  equivalent,  H, at a point in
tissue is given by the equation

              TABLE 1..S.  Recommended Values of  Q  for Different
                          Types of RadiationU)
                 	Radiation	      Q
                 X rays, gamma rays, and electrons           1
                 Neutrons, protons,  and singly charged
                 particles with a rest  mass greater
                 than 1 atomic mass  unit and with  an
                 unknown energy                            10
                 Alpha particles and multiply-charged
                 particles (and particles  of unknown
                 charge) with an unknown energy             20
                 (a) Based on Report No,  39 of the National
                     Council on Radiation Protection and
                     Measurements (NCRP 1971).
                                     1.32

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                                                               DARCOK-P  385-1
                H = DQN                                               (1.9)

where           H = dose equivalent (rem)
                D = absorbed dose (red)
                Q = quelity factor
                N = modifying factors.
The numerical value of N is generally considered equal  to 1.   The special  name
for the unit of dose equivalent is tne rem.   The SI  unit for  dose equivalent is
the si&vert  (Sv), which equals 1 J/kg.   If the absorbed dose  is given in units
of gray, then the dose equivalent is in  units of sievert.
     The dose equivalent is a valuable term because  the varying biological
effects of different types of radiation  are accounted for through the quality
factor, Q.  Therefore, tne effect of 1 rem (or O.Ol  Sv) of radiation is  nearly
the same for all types of radiation.  This equivalence permits the addition  of
dose equivalents when several radiations ere involved.
                 Section 1.5  BIOLOGICAL EFFECTS OF RADIATION

     Just as atoms are the basic building blocks of elements, cells are the
basic unit of the human body.  The body is composed of millions of cells, each
with a specific job to do to keep us alive and well.   When radiation transfers
energy to cells, primarily by the processes of excitation and 1onizat1on, it
can disturb the cells so they can no longer perform their original functions.
     Tne cells that make up the various tissues of the body do not have identi-
cal functions or appearances.  For example, the cells that make up nerve tissue
look and act differently from those that make up muscle tissue.  Each type of
cell may react differently to radiation.  Some cells  are more radiosensitive
than others (that is, susceptible to relation injury).  In the body, the most
radiosensitive cells are the blooc-producing and the  reproductive cells.
Muscle, nerve, and bone cells are the Usst radiosensitive.  Radiation has two
main types of effects on biological systems:  genetic effects and somatic
effects.
                                     1.33

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DARCOM-P 385-1
1.5.1  Genetic Effects
     Genetic effects are biological  effects of raciation  that  result  in  mi'te-
tions, or changes, in the genes of reproductive cells  and thet  are  e>pressed
in the descendants of the exposed individual.   Mutations  occur  in all  living
organisms.  They can be induced by agents such as  radiation  or  chemicals,  or
they can occur spontaneously, without any outside  alteration in the physical
Or chemical environment.  Genetic effects of radiation appear  as  birth defects
in the offspring c,f the irradiated individual  and  in succeeding generations,
as demonstrated in experiments involving thousands of  irradiated  animals.
These effects have not been observed in human populations, perhaps  because few
people have received the high doses thought to cause such effects.
1.5.2  Somatic Effects
     Somatic effects are biological effects of radiation  that  are expressed in
the exposed individual.  The somatic effects of radiation can  be divided into
prompt effects and delayed effects.
     A.   Prompt Effects.  Prompt effects are observed shortly after an indi-
vidual receives an acute rad i a t i o n d o s e, a very large dose received in a very
short time period.  Prompt effects are  generally associated with a  threshold;
that is,  if the radiation dose is below a certain level,  no effect is noticed,
but  if the dose exceeds that level, most people suffer an effect.  Prompt
effects tend to be short-term.  The short-term effects of acute exposure to
high levels of ionizing raciation are well known from observations of indi-
viduals exposed during  atomic warfare, medical treatments, or  industrial acci-
dents. These effects may include nausea, fatigue, blood disorders, intestinal
problems,  temporary loss of hair, skin  burns, and in extreme cases, death.
Table 1.6  shows the effect of an acute whole-body exposure  in  relation to dose,
and  Table  1.7 shows the effect of partial-body irradiation  in  selected organs.
Note that  whole-body irradiations are much more dangerous than partial-body
irradiations.  If radiation safety standards are met, there is no  reason for
any  individual to experience prompt radiation effects.
     B.   Delayed  E'fects.  Delayed effects can result from  an  acute  racieticn
dose and  are the major  effects of a chronic radiation dose.  A chronic  radia-
tion dose  is the continuous or repeated  subjection  of an  individual  to
                                      1.34

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                                                              DARCOM-r 365-1
     TABLE  1.6.  Dose-Effect Relationship for Acute Whole-Body Irradiation

            Acute Dose
               (rem'                          Nature of Effect
             5-23              Minimal dose oetectable by cnromosome
                               analysis or other specialized er,e";ysis
            25-125             Slight blood changes
            75-125             Minimal acute dose likely to produce
                               vomiting in about 10% of people so
                               exposed
           150-200             Temporary disability, blood changes
           300-500             Mean lethal dose
    TABLE 1.7.  Dose-Effect Relationship for Acute Partial-Body Irradiation
Acute Dose
(rad)
50
200
500
800
2000
2500
Organ
Testis
Ovary
Skin
Testis
Ovary
Liver
Skin
Effect in Relevant Oraans
Temporary sterility
Temporary amenorrhea, steril
Temporary reddening and loss
heir
Permanent sterility
Permanent menopause, steril i
Hepatitis
Temporary ulceration and


ity
of
ty


                                      permanent loss of hair

radiation at low dose rates over a long period of time.  Tne primary delayed
somatic effects are the development of cancer and, to a lesser extent, the
production of cataracts.   As opposed to prompt effects of radiation, delayed
effects are associated not with thresholds but with probabilities of occur-
rence:   as the radiation  dose increases, the likelihood of observing an effect
increases.  A relationship between radiation dose and cancer induction has been
shown from studies  of 1)  Japanese survivors of the atom bomb;  2) the Marshall
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DARCOM-P 385-1
Islanders, who were exposed to fallout from weapons  tests;  3)  uranium miners;
and 4) radiation therapy patients who received excessive  doses in the early
part of the century. These situations all  involved much  higher radiation doses
than those today's radiation workers can legally receive.
     C.  Relationship Between Exposure'and Delayed Effects.   Th^ exact rela-
tionship between chronic low-level exposure and delayed  effects is difficult to
establish for two reasons.  First, effects such as cancer can be caused not
only by radiation,but also by otht,- agents in the environment, such as ciga-
rette  smoke or chemical pollutants.  Second, long periods of time may elapse
between an exposure to radiation and the observation of  any effects.
     We do not yet know how radiation causes cancer.  However, most diseases
are caused by the interaction of a variety of factors, including general
physical condition, inherited traits, age, sex, and'exposure to outside agents.
It is  impossible to know whether a given cancer is caused by radiation or some
other  agent.  However, we do know that an increased incidence of cancer is
observed in groups of highly exposed people.  Although several studies have
been performed, there is no firm evidence that exposure  to radiation at cur-
rently accepted levels results in an increased incidence of cancer.
1.5.3  Environmental Dose and Occupational Dose Limits
     Individuals who work with radiation receive a  radiation dose from the
environment as well as from their workplace.  Table 1.8 shows the estimated
average individual dose in millirem from natural background radiation and other
sources.  The table indicates that the average individual in  the United States
receives a dose of about 200 mrem of radiation each year from sources that are
part of our natural and man-made environment.
     The standards of radiation dose suitable to the workplace are  set by
federal regulations.  Table 1.9 lists the dose standards for  various  parts of
the body.  These standards do not represent boundaries between safe  conditions
and harmful or lethal conditions.  Rather, they represent dose levels for which
regulators consider there is sufficiently small probability of radiation
effects,  because the likelihood of causing an effect increases  gradually witn
increasing dose, it is wise to keep the actual radiation dose as  low as is
reasonably achievable (ALARA).
                                     1.36

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   TABLE 1.8.
                                          DARCO>;-P  385-1
U.S. Generei-Population Dose Estimates (I978)'s;
                  Source
 Natural  background
 Medical
 Release  of radioactive material  by  mining,
   milling, etc.
 Nuclear  weapons  development (primarily
   fallout)
 f.1' clear  energy
 Consumer products
 TOTAL
                                 Average Indivicub'
                                   Dose (mrerc/yr'
                                      100
                                       90
                                        5 to 8
                                        0.28
                                        0.03
                                      200 mrem/yr
(a) Interagency Task Force on  the  Health  Effects  of Ionizing
     Radiation, 1979.
    TABLE 1.9.   Maximum Dose Equivalent  Per Calendar Quarter^
                     Oroan
  Whole  body;  head and trunk;  active  blood-forming
  organs;  lens of eyes;  gonads
  Hands  and  forearms;  feet  and ankles
  Skin of  whole  body
                                       Amount (rem)
                                           1.25
                                          18.75
                                           7.5C
  (a) AR  40-14.
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DARCOM-P 385-1
      Section 1.6.   PROPERTIES  OF  RADIOACTIVE MATERIALS IMPORTANT IN THE
                    DEVELOPMENT OF RADIATION PROTECTION PROCEDURES

     Several  properties  of radioactive  materials  play a key role in the devel-
opment of radiation protection  procedures.  These include the distinction
between external  and internal exposure,  and the properties of dispersibility,
chemical  toxicity,  radiotoxicity,  and  criticality.
1.6.1  External  Versus  Internal Exposure
     The extent  to which radiation causes  biological effects depends  in part
on whether the body is  exposed  externally  or  internally, and on  the types  of
radiation involved in the exposure.
     A.  External Exposure.  External  exposure  results  from exposure  to a
source of ionizing radiation outside the body.   Sources of external exposure
can be divided into two classes:  penetrating  and nonpenetrating radiations.
Penetrating radiations—gamma rays, x  rays,  and neutrons—have  sufficient
energy to pass through  the surface of  the  skin and interact with internal  body
tissues.  Nonpenetrating radiations--alpha particles  and  low-energy  beta
particles — interact only with the skin surface.   Therefore, from the  stand-
point of external exposure, penetrating radiation is  a  greater  hazard than
nonpenetrating radiation.
     The principles and procedures that minimize external  exposure,  and the
calculation of external dose, are discussed in Chapter  6.
     B.  Internal Exposure.  Radioactive materials can  be  taken into  the  body
by inoestion, inhalation, and absorption through pores  of  the  skin or through
breaks in the skin.  Once in the body, these  materials  may be  deposited in
vtrious organs and constitute a source of internal exposure.
     A stable isotope and a radioactive isotope of the  same  element have
identical chemical behavior in  the body.  The chemical  characteristics of an
isotope or nuclide determine the organ in which  it is deposited as well as the
rate at which it is excretec frorr, the  body.   If e radicnuclide has no stcMe
counterpart  in the body, it follows the metabolism and excretion pattern of
another element with similar chemical  properties.  For example, strontium is
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                                                              DARCOM-P  3S5-1
not normally found in large Quantities  within the body.   However,  the chemical
properties of strontium ere similar to  those  of calcium.   Thus,  strontium that
enters the body behaves much es calcium does  and may be  deposited  in the bone.
     The radiological hazard associated with  internal  exposure depends upon the
type of radiation emitted by a radionuclide,  the radiosensitivity  of the organ
in which it is deposited, and the physical  properties  of the  radionuclide
(e.g., its solubility and particle size).   Of the various types  of radiation,
alpha particles are usually considered  the  greatest  internal  hazard.
     The calculation of internal  dose is  discussed in  Chapter  5  along with two
principles that are important in making those calculations,  the  principles of
maximum permissible concentration (MPC) and the critical  organ.   Procedures
for minimizing internal exposure to radiation are also discussed ir. Chapter 5.
1.6.2  Dispersibility
     The physical form of a radioactive material and its intended  use influ-
ence how much it will scatter, or disperse.  For example, a  radioactive powder
has a greater chance of being scattered over a wide  area than  does e sealed
source.  Conditions of use under which  various forms of radioactive material
are nondispersible, of limited dispersibility, dispersible,  or highly dispers-
ible ere listed below.  Engineered safeguards and administrative controls for
each of these types of materials are discussed throughout the  manual.
     A.  Nondispsrsi'ble
          1.    nondestructive use of encapsulated or sealed  sources
          2.    storage of nonflammable, nonexplosive radioactive materials in
               sealed containers especially designed for such  storage.
     B.  Li mi ted Pi s pers i bi 1 i ty_ _
          1.    simple operations that can result only in fractional releases of
               material from a radiation area during credible  accidents
(a)  Criteria usec  zo  classify radionuclides in this category are subjective
    end thus depend  in  part upon experience end judgment.
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DARCOM-P 385-1
          2.   use of radioactive materials that  are  strongly  bound  in  a  solid
               matrix or biological system.
     C.  Dispersible
          1.   use of unsealed, noncombustible, nonexplosive  liquids or compact
               solids in standard chemical processes  or operations.
     D.  Highly or Readily Dispersible
          1.   use of radionuclides in hazardous  or complex chemical
               operations
          2.   use of radioactive powders, gases, vapors,  or  other aerosols
          3.   use of radioactive materials in combustible or explosive
               procedures
          4.   dry, dusty operations
          5.   high-temperature or high-pressure operations  that may increase
               the probability of producing radioactive aerosols
          6.   use of radioactive materials that can ignite  spontaneously.
 1.6.3   Chemical Toxicity
     Chemical  toxidty refers to the harm  that can be caused  by an element
 because of  its chemical nature.  Many elements are toxic and  can cause severe
 il'ii'iess 1f  ingested.  Examples of toxic elements Include arsenic, which damages
 blood  vessels; cadmium, which is a kidney  poison; mercury, which in large doses
 is  a kidney poison and in chronic situations affects the nervous system; and
 lead,  which also affects the nervous system.  A  radionuclide  may be hazardous
 both because  of its chemical nature and because  of the radiation it emits.
 Uranium, for  example, is a kidney poison and is  also radioactive (it has no
 stable Isotopes).  In the case of long-lived isotopes of uranium, it is the
 chemical rather than the radiation hazard  that limits the amount that may
 safely be ingested.  Other radioactive materials, such as plutonium, have
 negligible  chemical toxicity but are considered  hazardous because of the amount
 of  radiation  damage they an produce.  These materials are called radiotoxins.
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                                                              DARCOM-P 385-1
1.6.4  Radiotoxicity
     The term radiotoxicity indicates the relative radiological  hazard
associated with internally deposited radionucl ides.   Nuclides that ere highly
radiotoxic, such es those that emit alpha particles  or high-energy beta
particles, present the greatest relative health nazard when deposited inter-
nally.  The level of radiotoxicity strongly dictates the degree  of control
required in work with radioactive materials.  A listing of the relative radio-
toxicity of some radionuclides is given in Table 1.10.  Note that Group I
radionucl ides are the least radiotoxic and Group 8 the most radiotoxic.
1.6.5  Criticality
     Fission occurs when a heavy nucleus (with an atomic number, Z, of 90 or
more) absorbs a neutron and splits into two lighter nuclei, each with about
half the mass of the original nucleus.  Each fission can also result in the
emission of up to eight neutrons, with two and one-half neutrons being the
average.  A nuclide that is capable of undergoing fission is called a fission-
                                                                      "' ............... '
able, nuclide.  Examples of fissionable nuclides are    U and    U.  Some
        "'•         238
nuclides, such as    U, undergo fission only when 'they absorb a fast neutron.
                        235      23°
Other nuclides, such as    U and   "Pu, undergo fission when they absorb a
thermal or slow neutron, and are called fissile nuclides.  Materials that con-
tain such nuclides are fissile materials.   Natural  uranium, which is a combina-
tior
23Sf
tion of 235U and 238U,  is a fissile material.   Some  nuclides,  such as *35U and
   'Pu, also undergo spontaneous f1ssion;  that is, they can split without first
having been irradiated by neutrons.  Table 1.11 lists some of the more common
fissionable nuclides.
     After a fission,  the neutrons that are released have three possible fetes.
They may 1) completely escape from the fissile material, 2) be absorbed by
nonf1ss1le atoms, or 3) be captured by fissile atoms.  If they are captured by
fissile atoms, more fissions can occur and more neutrons may be released,  The
continuing process of  fission, release of neutrons, capture of neutrons, and
subsequent fission is  called a chain reaction.
     If the neutrons released by a fissioning atom cause, on the average, less
then one subsequent fission, then no chain reaction is possible and the reaction
is said to be subcritlcal.  When the neutrons released by each fission cause
                                     1.41

-------
DA.RCOK-P  385-1
                 TABLE  1.10.   Radiotoxicity of Various  Nudities
                                                                        (a)
Radionucl ioes
Grouped by
Relative
Radiotoxicitv
Crouo 1
14c
Grouo 1 1
Activity in Curies of Single
Innalation that Results in
IS-reir Oose(b)
to \.
6.
2.
"< !:
Crouo 1 1 1
-:
_Au
H /

132

141

S5
140
95

65
56

55


Ca

1

Ce

Sr
La
Nb

Zn
Co

Fe
7.
7.

2.

4.

7.

2.
4.
3.

2.
8.

3.
r-, tica .
15
88
84
17
23
25

59

50

06

00
20
60

60
40

00
x
X
X
X
x
x

X

X

X

X
X
X

X
X

X
10
10
10
10
10
10

10

10

10

10
10
10

10
10

10
Groan
-2
-2
-3
-4
-4
-4

-4

-4

-3
-4
-3

-4
-3

-4


5
2
6
S

4



4

2
2
2

1
1

1
to

.30
.30
.90
.30

.60



.20

.70
.60
.30

.50
.30

.30
Lunoic )
...
x
X
X
X

X

--.

X

X
X
X

X
X

X

10
10
10
10

10



10

10
10
10

10
10

10

-3
-3
-4
-4
-4



-4

-4
-4
-4

-4
-4

-4

Crouo IV
181

147
32
140
234

85
192

36

91

Hf

Pm
P
Ba
Ba

Kr
Ir

Cl

y
182.

45

89
137

60
144
126
15-

>a

Ca

Sr
Cs

Co
Ce
1
Eu

9.

8.
8.
1.
8.

' 6.
3.

2.

5.
1.

4.

4.
2.

2.
1 .
1 .
1.

94

90
70
40
50

90
20

70

00
10

30

00
60

60
40
40
30

X

X
X
X
X

X
X

X

X
X

X

X
X

X
X
X
X

10

10
10
10
10

10
10

10

10
1C

10

10
10

10
10
10
10
-5

-5
-5
-4
-5

-5
-4

-5

-5
-4

-5

-5
-9

-5
-5
-5
-5


1

^
2
6
7

5
6

5

7
5

2

8
2

1
1
7
1

.92

.30
.10
.60
.30

.80
.90

.30

.30
.00

.60

.50
.20

.50
.50
.30
.60

X

X
X
X
X

X
X

X

X
X

X

X
X

X
X
'
X

10

10
10
10
10

10
10

10

10
10

10

10
10

10
10
10
10
-4

-4
-4
-5
-5

-5
-5

-5

-5
-5

-4

-5
-5

-5
-5
-4
-5

                                                  Croup IV (contd)

                                                  131
                                                  170.
                                                   82
  .1
   Tm
   Br
                                                  Croup VI
233
210
227
 90
210

233
                                                     Po
                                                     Cm
   U 4 U Nat
   Th & Th Nat

Crouo VI I
                                                     Sm
147
144

244 Ra
   Cm

CroL'S VIII

243
241^
237^
227NP
230AC
                                                  236
                                                  239
                                                  240
   Pu
   Pu
  "Pu
1.20 x 10
3.80 x 10
                 7.47
                                                                          10
2.25
.-5
 -5
 -3
                                                                    2.30 x 10
                                                                    9.10 x 10
                          -6
                          -6
3.90 x 10
1.30 x 10
5.50 x 10
3.90 x 10
3.20 x 10
3.00 x 10
7.00 x 10
1.10 x 10
1.90 x 10
                                                                           10
 -6
 -6
[-7
 -7
 -7
 -7
 -7
 -6
 -7
 -9
7.70 x 10
7.70 x 10
4.90 x 10
1.10 x 10
.-S
 -S
                                                                             -6
                                                                      60  x  10
                                                                      60  x  10
                                                                      20  x  10
                                                                      00  x  10
                                                                      80  x  10
                                                                      50  x  10
2.20 x 10
2.00 x 10
2.00 x 10
 -9
 -9
 -9
 -9
 -9
 -9
 -9
7.30 x 10
7.30 x 10
    -4
    -5
                 1 .60 x 10
 .30
 ,00
"•::
x 10 _
                                                                                    4.60 x 10
                                                                                    1.30 x 10
  30 x 10
  60 x 10
  70 x 10
  60 x 10
3.00 x 10
2.60 x 10
6.90
7.30
    -5
   ;-7
    -7
    -7
    -7
    -7
    -8
  10"_
                                                                                           10
         1.50  x  10
         2.30  x  10
         -8
         -7
2.70 x 10
2.70 x 10
2.70 x 10
6.90 x 10
2.30 x 10
£.50 x 10
E.50 x 10
                         .-7
                          -7
                          -7
                          -6
                          -6
 £.50 x
 3.50 x
  10
  10
-8
-9
-8
 (a) Broasky 1965.
 (b) Insoluble materials.
 (c) 50-year cumulative  dose.
                                             1.42

-------
                                                               DARCOM-P 385-1
                      TABLE 1.11.  Fissionable Materials
          Capable of Chain Reaction
        with Fast and Thermal (Neutrons
233u
235U
239Pu
?^Pu
242A.
243Cm
2«5Cm
247Cm
249Cf
251r^
Capable of Chain Reaction
 with Fast Neutrons Onlv
          237
          241
          244
          238
          240
          242
          238,
                                                           Np
                                                           Am
                                                           Cm
                                                           Pu
                                                           Pu
                                                           Pu
one additional fission, then the reaction is  self-perpetuating and the chain
reaction is said to be critical.  Finally,  if the neutrons released from a
fissioning atom cause, on the average, more than one subsequent fission, the
reaction is said to be supercritical.   An unplanned supercritical  chain reac-
tion is called a criticality accident.
     Criticality accidents are extremely serious because very high levels of
gamma and neutron radiation can be produced.   That is,  lethal doses of radia-
tion can be received in a very short  time.   For this reason, special efforts
ere made to reduce the chances of a criticality accident to a very low level.
Particularly important is the design  of facilities.  "Safe-by-geometry" is the
best rule to remember in reducing the  probability of a  criticality accident.
     A.  The Double-Contingency Rule.   One  of the most  Generally accepted
approaches to preventing a criticality accident is the  double-contingency
rule.   This rule assumes that a sufficient  number of limits end controls
exists to ensure that, before a criticality is possible, at least two unlikely,
independent, and concurrent changes must occur in one or more of the condi-
tions  specified as essential to nuclear safety.  This rule calls for controls
which  ensure that no single mishap can lead to a criticality accident, regard-
less of the probability that that mishap might occur.
                                     1.43

-------
DARCOM-P 3S5-1
     B.   Factors That Affect Criticality.   Nine  physical factors  affect  the
likelihood that an accumulation of fissile  material  will sustain  a chain  reac-
tion.  Criticality safety programs take  account  of  these factors  and  employ
safeguards based on them to prevent criticality  accidents.
     (1)  Amount.  The amount of fissionable material  needed  to  support  a  chain
reaction is the critical  mess.   If the amount of fissionable  material  present
is small enough, criticality cannot occur  no natter what the  condition of  the
other eight physical  factors.  On the other hand, the  greater the amount  of
fissile material present, the more difficult it  is  to  avoid criticality.
Limiting the amount of material present  helps ensure a subcritical state.
Many safeguards are designed to limit the  total  amount of  fissile material
that can be assembled in one place.
     (2)  Geometry.  The size and shape  of fissile  materials  have an  important
effect on the probability of a  chain reaction occurring.   Decreasing  the dis-
tance that neutrons travel  within the fissile material decreases  the  chance
that the neutrons will interact to cause a subsequent fission.   For  this rea-
son, a thin slab of fissile material is  unlikely to support -Mssion  reactions,
but a sphere is most conducive to criticality.
     (3)  Density.  If the density of fissile materials is increased, the fis-
sile atoms are more tightly packed together.  This  packing reduces the chance
that a neutron will escape from the material; thus, the higher the density of
the material and the atoms in it, the higher the probability  that a  fissile
atom will capture a neutron in the material and undergo fission.
     (4)  Moderation.  The speed of a neutron affects its  chances of being
captured by a fissile atom.  The faster a  neutron travels, the less  likely it
is to be captured.  Thus, the fast neutrons produced by a  fission are not
likely to cause more fissions until they slow down.
     Fast neutrons are slowed down when  they collide with, but are not absorbed
by, the nuclei of atoms.   This  slowing-down process is called thermalization,
or moderation.  Moderation of a neutron  increases its chances of being capr
tured and causing a fission.  Graphite and hydrogen-containing materials such
as paraffin, oil, and water are good moderators.  Human tissues are 70* water
and thus are good moderators also.
                                     1.44

-------
                                                              DARCO.M-P 385-1

     (5)  Re-'lection.  Neutrons that escape from fissile materiel?  w.thout
causing additional fissions or being absorbed by atoms continue  to  move  away
from the materials unless they hit something in their path.   Anything placed
close to fissile material will tend to bounce (reflect) a few of the neutrons
back into the material and give the fissile atoms another chance to absorb
them.
     Materials that have a low atomic number are good reflectors; in fact,
many moderators are also very good reflectors.   Human tissues ere botn good
moderators and good reflectors.
                                                                        238
     (5)  Enrichment.  Naturally occurring uranium is mostly nonfissile     U
                         235
and less than 1% fissile    U.  If the world's entire supply of  natural
uranium ore were collected into a giant sphere and.covered with  drinking
water, it would not be critical.  However, uranium can be enriched.  Uranium
                                              235
is said to be enriched when the percentage of    U atoms has been increased
above the percentage found in natural uranium.   As the enrichment increases,
the number of fissile atoms that can capture neutrons and then undergo fission
increases, and fewer nonfissile atoms are available to capture neutrons  and
prevent the fissioning process.  Therefore, the greater the level of enrich-
ment, the easier it is for an accumulation of fissile material to attain
criticality.
     (?)  Interaction.  The escape of neutrons from one accumulation or "pile"
of fissile material, and their subsequent entrance into another  accumulation
that can cause more fissions, is called interaction.   Interaction can occur if
accumulations of fissile material  are close enough together.  For this reason,
accumulations of fissile material  must be stored far enough apart to prevent
interaction.  Keeping accumulations of fissile material at preestablished
distances apart is a commonly used criticality control technique.
     (8)  Type of Material.   Each type of fissile materiel has different
nuclear properties.   For example,  the amount of    U needed to support a
                                                    230
chain reaction is about 1 kg, whereas the amount of    Pu needed is only
about 1/2 kg.
                                     1.45

-------
DARCOM-P 385-1

     (9)  Nuclear Poisons.   Nuclear poisons  are  materials  tSet  absorb neutrons

without undergoing fission.  This absorption decreases  the number  of neutrors

available to cause e fission.  Examples of nuclear poisons include cadmium,

boron, and samarium.

     The nine factors mentioned above can interact and  make the problem of

determining safe bendling procedures for fissile materials very complex.

Because of this complexity, a criticality safety expert should  be  consulted

whenever questions arise concerning criticality  safety.
                                  REFERENCES
Brodsky, A.  1965.  "Determining Industrial Hygiene Requirements for Instal-
  lations Using Radioactive Materials."  Amer.'lnd. Hyc.  Ass.  J. 26:294-310.

International Commission on Radiation Units and Measurements (ICRU).  1980.
  Radiation Quantities and Units.  ICRU 33, Washington, D.C.

Interagency Task Force on Health Effects of Ionizing Radiation.  1979.  Report
  of the Interagency Task Force on the Health Effects of Ionizing Radiation.
  U.S.  Department of Health, Education and Welfare, Washington, D.C.

Lederer, C. M., and V. S. Shirley, eds.  1978.  Table o^ Isotopes.  John Wiley
  and Sons, New York.

National Council or Radiation Protection and Measurements (NCRP).   1971.   Basic
  Radiation Protection Criteria.  NCRP 39, Washington, D.C.

Radiological Health Handbook.   1970.  U.S. Department cf Health, Education  end
  Welfare, Bureau of Radiological Health.  Available from U.S.  Government
  Printing Office, Washington,  D.C.

U.S. Department of the Army and Defense Logistics  Agency.   Mecical  Services  -
  Control and Recording Procedures for Exposure to  Ionizing Radiation and
  Raoioactive MateriaTT  AR 40-14, DLAR 1000.2£,  Wasmngton,  D.C.
                                      1.46

-------
                                                               DARCOK-P 385-1
                                  APPENDIX A

                                  DECAY  SCHEMES
     Decay  schemes are c'iegrammatic  representations cf radioactive decay  path-
 ways.  The  chemical  symbol, mass number, and half-life of the parent nuclide
 eopear on the uppermost horizontal line.  Decay leading to an increase in the
 N/Z  ratio (alpha emission, positron  emission, and electron capture) is indi-
 ceted by a  bent arrow leading to the lower left; decay leading to a decrease in
 the  N/Z ratio (electron emission) is drawn with an errow leading to the lower
 right.  These arrows terminate on horizontal lines that represent zhe nuclear
 energy levels of the daughter nucleus.  If the daughter nucleus formed is in an
 excited state, then gamma rays are emitted until the ground, or unexcited,
 state is reached.  Gamma rays are represented by vertical lines that may  be
 either straight or wavy.  The maximum kinetic energy of the emitted particle or
 the  energy of the gamma ray is indicated near the appropriate arrow.  If more
 than one pathway may be followed, the fractional or percentage occurrence of
 each pathway is indicated.
     As mentioned in Section 1.2.2, radium-226 (which has a half-life of
 1600 years,  or 1.60 x 10  years) can undergo radioactive decay by five path-
ways.  The first and most common pathway consists of the emission of en alpha
particle that has 4.78 MeV of kinetic energy.  In this case, the daughter
nucleus,  redon-222,  is not in an excited state, so no gemma rey is emitted.
A decay scheme showing this  pathway is shown below:
                     226
                       Ra
1.60x10
                                                      a 4.78 MeV
                                                       94.45*
                                      0.0
                     222
                       Rn
                                      1.47

-------
DARCOM-P 385-1
This scheme  indicates  that of all the decays of 226Ra to 222Rn, 94.45%
proceed by the  emission of only a 4.78-MeV alpha particle.
     The next most  common pathway for the decay of    Ra is the emission of
an  alpha particle that has a kinetic energy of 4.60 MeV.  The dauahter nucleus
222
    Rn, is in an excited state, and a gamma ray of 0.186 MeV is emitted.  The
scheme for this pathway is shown below.
                       226
                         Ra
1.60xl03y
                                                       a 4.60 MeV
                                                        5.55%
                                           0.186
                                           0.0
                       222Rn

This scheme shows that, of all the decays of    Ra, 5.55% decay by this path-
way.  The numbers to the right of the horizontal lines represent the energy
level of the daughter nucleus, in MeV.  The straight vertical line between the
0.186 line and the 0.0 line represents a gamma ray that has an energy of
0.186 MeV (shown by the numbers above the gamma ray).
                              226
     A third pathway by which '"Ra decays 1s the emission of a 4.34-MeV
alpha particle.  The 222Rn dauahter nucleus 1s left 1n an excited state and
loses energy by the emission of two gamma rays, one that has 0.262 MeV and a
second that has 0.186 MeV of energy.   The two gamma rays are emitted 1n oulck
succession, the 0.262-MeV gamma first, followed by the 0.186-MeV gamma.  The
0.186-MeV level of 222Rn has a half-life of 0.32 nsec (3.2 x 10'10 sec).
This amount of time 1s not long enough for this energy level to be considered
             p50
an isomer of    Rn.   The decay scheme 1s shown on the next page.
                                    1,48

-------
                       226
                          'Ra
1.60xltr y
                                                                DARCO>!-P 385-1
                                                       a 4.34 MeV
                                                        0.0055^
                                            0.448
                  0.32ns
                                            0.186
                                            0.0
                        222,
                          Rn
                                  226r
      The fourth pathway by which '-''"Ra decays  is  the  emission  of an alpha
 particle that has 4.19 MeV of kinetic energy.  The  222Rn  daughter nucleus is
 in an excited state and releases its extra energy in  two  ways:   62% of the
 time, a sinn'e gamma ray with 0.601 MeV of energy is  emitted;  3B% of the time,
 two gamma rays are emitted, one following the  other in a  cascade.   The total
 energy of the two gamma rays (0.415 and 0.186  MeV)  is equal  to the energy of
 the single emitted gamma ray.  The scheme may  be  drawn as  follows:
                       226
                         'Ra
 i.60xio3:
                                   ^  ^              a^.19
                                VV            J    °'°°
                               ^  
-------
DARCOM-P 385-1
                0.32ns
•26Ra 1.60xl03y
" 0.186
r 0.0
~4
2*7 yi r\ ct
• / Ai w »'
0.00107.
0.00657c
5.5500%
04.45007:
— ^ 	 1
a
4 16 MsV
4 19 N\6V
4.34WeV
4.60MeV
4.78MeV
                 222,
                   Rn
     This diagram is one way of presenting the decay scheme information.
Notice that the gamma ray resulting from the de-excitation of the 0.185-MeV
nuclear energy level follows the 4.60-MeV alpha, the 4.3^-MeV alpha, the
4.19-MeV alpha, and the 4.16-MeV alpha.
                                    1.50

-------
                                                               DARCOM-P  385-1
     A decay  scheme  for  beta  negative emission, in the decay of scandium-46 to
titenium-46,  is  shown  belov/:
                               89.3d
                           0.004
                                    100*
                                    0.357MeV
                                     /      ^
                           1.457 McV \ \     N>
      2.010
                                              
-------
                                                              DARCO.M-P 385-1

                     CHAPTER 2.  RADIATION INSTRUMENTATION
2.1  BASIC CONCEPTS IK RADIATION DETECTION AND MEASUREMENT  .     .     .    2.5

     2.1.1  Characteristics of Instruments   	    2.5

     2.1.2  Source Characteristics 	    2.8

2.2  RADIATION PROTECTION INSTRUMENTS AND HOW THEY WORK     .     .     .    2.9

     2.2.1  Gas lonization Detectors	2.9

            A.  Pri.iciples of Operation	2.9

            B.  lonization Chambers     .     .     .     .     .     .     .    2.14

            C.  Proportional Counters   .     .     .     .     .     .     .    2.16

            D.  Geiger-Mueller Counters .......    2.19

     2.2.2  Scintillation Detectors     	    2.20

            A.  Principles of Operation	2.21

            B.  Inorganic Scintillators .......    2.22

            C.  Organic Scintillators   	    2.23

     2.2.3  Semiconductor Detectors     	    2.26

2.3  CALIBRATION OF INSTRUMENTS    	    2.29

     2.3.1  Calibration Sources    	    2.29

     2.3.2  Calibration Facilities 	    2.30

     2.3.3  Instrument Characteristics That Affect Calibration and
            Calibration Frequency  ........    2.31

2.4  FACTORS  THAT  AFFECT THE SELECTION AND USE  OF RADIATION-
     MONITORING INSTRUMENTS	2.33

     2.4.1  Detection  Versus Measurement     .     .     .     .     .     .    2.33

     2.4.2 Type of Radiation	2.33

     2.4.3  Radiation  Energy and  Instrument Energy'Dependence    .     .    2.34

     2.4.4 Nonuniform Fields .........    2.34
                                      2.1

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DARCOM-P 385-1




     2.4.5  Angular Dependence	_.     .     .     2.34



     2.4.6  Calibration	2.35



     2.4.7  Unwanted Response .......               2.35



2.5  TYPES OF RADIATION-MONITORING INSTRUMENTS     	     2.35



     2.5.1  Portaole Survey Meters ........     2.35



            A.  Portable Detection Instruments     	     2.36



            B.  Portable Measurement instruments   .     .     .     .     .     2.36



     2.5.2  Laboratory Counting Instruments  ......     2.37



     2.5.3  Air-Monitoring Equipment    .......     2.39



            A.  Air Samplers	2.39



            B.  Air Monitors	2.40



            C.  Principles of Operation .......     2.41



     2.5.4  Other Fixed Instruments     	     2.42



            A.  Remote Area Monitors    .     .     .     .     .    .    .     2.42



            B.  Continuous Air Monitors .......     2.42



2.6  PERSONNEL DOSIMETERS     	     2.43



     2.6.1  Photographic Film	2.43



            A.  Principles of Operation .......     2.44



            B.  Dosimeter Design    ........     2.4£



            C.  Effects of Environment  .     .     .     .     .     .    .    2.44



            D.  Processing Techniques   .     .     .     .     .     .    .    2.45



            E.  Interpretation and  Calibration    .....    2.45



     2.6.2  Nuclear Track Emulsions     .......    2.46



     2.6.3  Thermo!uminescence Dosimeters    ......    2.46



            A.  Principles of Operation .......    2.46



            B.  Advantages and Limitations   ......    2.47
                                      2.2

-------
            C.   Interpretation and Calibration

            D.   Practical Applications  	

2.7  STATISTICS AND ERROR DETERMINATION 	

     2.7.1  Systematic and Random Errors of Measurement

     2.7.2  Basic Statistical  Distributions for Radioactive Decay

2.8  RECORDS   	
                 t
REFERENCES  	
DARCOM-P 385-1

            2.48

            2.46

            2.49

            2.49

            2.49

            2.52

            2.52
                                    FIGURES


2.1  Simplified Version of a Chamber Used to Collect Ions

2.2  Relationship Between Applied Voltage and the Number of
     Electrons Collected on the Anode   	

2.3  Plateaus for Typical Proportional Counter

2.4  Diffused p-n Junction Detector     	

2.5  Frequency of Occurrence of Count Rates for a
     Long-Lived Sample   ........

2.6  Normal Distribution Function Showing Standard Deviations
     and Mean  	
            2.9


            2.12

            2.18

            2.28


            2.50


            2.51
                                    TABLES
2.1  Portable Survey Instruments
                         •»

2.2  Laboratory Counters .
             2.37

             2.38
                                      2.3

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                                                               DAKCO.M-P 385-1
                     CHAPTER 2.  RADIATION INSTRUMENTATION1
     Ionizing radiation cannc: be detected by unaided human senses;  instru-
mentation must be used to detect and measure it.   This chapter describes the
fundamental characteristics of radiation detection and measurement instruments
and their principles of operation, their application, and their limitations.
Included in the chapter are an introduction to measurement concepts;  a review
of instruments used in the field of radiction protection and how they work;
information on the calibration of instruments; factors that affect the selec-
tion of radiation-monitoring instruments; the types of monitoring instruments
and personnel dosimeters available for use; and a brief discussion of sta-
tistics and error determination.
      Section 2.1  BASIC CONCEPTS IK RADIATION DETECTION AND MEASUREMENT

     Numerous types of instruments are used for a wide variety of purposes in
the field of radiation protection.  Some instruments simply detect the pres-
ence of radiation; others give a quantitative measurement of the dose rate or
exposure rate produced by the radiation.
     Detection and measurement instruments have two basic components, a sensing
element and an indicating element.  The sensing element, called the detector.
responds to the radiation and through various means provides a measurable
signal to the indicating element.  Common types of indicating elements include
meters, recorders, counting sealers, and speakers.  Intermediate electronic
circuitry may be used to amplify the signal from the detector so that it can
be more readily observed in the indicating element.
2.1.1  Characteristics of Instruments
     Instruments can be characterized by how radiation Interacts with the
detector.   Several instruments depend for their operation on the ionizetion of
matter by radiation.   Other detection systems depend largely upon the excita-
tion of electrons rather than on ionizatlon.   Both ionization and excitation
                                      2.5

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DARCOM-P 385-1
result either directly or indirectly  in  the  formation of electrical charges
within the sensitive volume of the  detector  or  its associated circuitry.   If
an electric field is applied across the  sensitive volume of the detector,  the
electrical charges can be collected because  negative charges travel to  the
positive pole of the electric field and  positive charges travel to  the  nega-
tive pole.  The collection of the electrical  charges causes a build-up  of
charge that flows through an external  circuit.
     A second wey, of distinguishing types of instruments is by whether  the
flow of charge is recorded as a pulse or a current.  An instrument  that oper-
ates in the pulse mode records an output pulse  for each individual  interac-
tion between the detector and a particle or  photon of radiation.   An  instrument
that operates in the current mode records an average of many individual
interactions and subsequent pulse fluctuations.  An advantage of  the  pulse
mode is that, for many instruments, the  amplitude  (size or  height)  of each
individual pulse carries valuable information about the type and  energy of the
radiation that caused the pulse; in the  current mode, information on  individual
pulses, and thus on individual radiations, is lost.  Pulse  detectors  also  have
a greater sensitivity than detectors  that operate  in the current  mode;  that
is, they detect more of the incoming  radiation.  Because of these advantages,
the pulse mode is more commonly used  for radiation detection  instruments.
      A third distinction among instruments  is  how those that  operate in the
pulse mode record the pulses.  Rate meters record  a  pulse  rate,  with  readouts
in counts per minute (cpm), mR/hr, mrem/hr,  etc.   Integrating  instruments  have
a digital counting accessory that tallies the pulses for the duration of the
measurement, with readouts given in counts,  mR, mrad, etc.
     Counting devices that accept pulses may have  fixed or variable discrim-
inators.  The pulse amplitude must be of a certain  size to pass  the discrimi-
nator level end be counted; otherwise it is rejected.   If  the  discriminator
level can be varied, information can  be obtained about  the amplitude distri-
bution of the pulses, and therefore about the types  and energies of the
radiations.
     In nearly all detector systems,  a minimum amount  of time is required
between two interactions in order for them to be registered as two separate
                                      2.6

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                                                              DARCOM-P 385-1
pulses.  This interval  is called the dead time  of  the  system,   immediately -
after E pulse, the detector is insensitive to radiation  and  is  unable  to
respond to other ioniring events.   If an  ionizing  event  occurs  during  this
time, it does not produce a pulse.   The important  consequence  of  dead  time is
that a detector in a high-radiation field may indicate less  radiation  than is
actually present.  Counts that are  recorded can be corrected for  dead-time
losses, and many laboratory counters have a meter  that indicates  the percent-
age of time the counter is dead.
                 »
     The object of many applications of radiation  detectors  is  to identify the
energy distribution of  the incident radiation.   The ability  of  a  detection
system to distinguish between or separate two pulses of  slightly  different
sizes is called its energy resolution.   The resolution .capabilities of various
instruments are discussed later in  this chapter.
     If a detector counts every particle  or photon of  radiation that enters
its sensitive volume, it has a counting efficiency of  100%.  Practically
speaking, however, counting efficiencies  of 100% are rarely  achieved.   It is
always possible, especially with gamma  rays and neutrons,  that  some radiation
will pass through the detector without  interacting with  it.  In order  to  relate
the number of pulses counted to the actual number  of radiations incident  on
the detector, the detector's counting efficiency must  be calculated.   The
absolute efficiency is  calculated using Equation (2.1).
     = kr«i,,+^ *f*i^
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DAKCOM-P 385-1
2.1.2  Source Characteristics
     In the detection and measurement  of  radietion,  consideration  must  also  be
given to characteristics of the  radioactive  source.   The  emission  of  raciation
from a radioactive source is generally assumed  to  be isotropic;  that  is,  radia-
tion is emitted by the source  in all  directions with equal  intensity.   In oraer
for all of the radiation emitted by the source  to  be detected,  the source must
be completely enclosed within  the sensitive  volume of the detector.   This type
of counting arrangement is called 4-  geometry (because the solid angle  sub-
tended by the detector at the  source-  position is 4n steradiens).   Most  detec-
tion systems do not achieve 4n geometry because the source is  placed  outside
the detector and only a fraction of the emitted radiation is  directed toward
the sensitive volume.  The geometry factor is the  fraction of  the  source  sphere
that actually intercepts the detector.  It can  be  used to determine the actual
number of radiations being emitted by the source.
      Other source factors that  must  be considered are self-absorption, radia-
tion attenuation, and the inverse-square law.  When a radioactive source  pro-
duces radiation, there is a finite probability  that the radiation will  lose
its energy within the source itself.   This process, called self-absorption,
occurs most frequently with encapsulated alpha  and beta sources because the
energy of the particles is absorbed by the capsule material.   Radiation may
also lose its energy in the air between the source and the detector,  or in the
shielding of the detector before it reaches the sensitive volume, and this
attenuation must be considered when attempting to determine the activity of
the source.  Finally, assuming that the radioactive source is a point source
(very small compared to the distance to the detector) and that particles or
photons radiate outward from it, the number of radiations in a unit area fells
off with cistance.  The greater the distance between the source and the
detector, the fewer the radiations entering the sensitive volume of the
detector and therefore the lower the count rate.  A complete discussion of
this principle, celled the jnverse-souare law,  is presented in Chapter 6.
                                      2.8

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                                                              DARCOM-F  385-1
        Section 2.2  RADIATION PRQTECTIQf. INSTRUMENTS AND HOW THEY WORK

     Instruments used for radiation protection are of three general types:
ges ionization detectors, scintillation detectors, and semiconductor detectors.
The principles on which these detectors work and the types of detectors in
each group are described in this section.
2.2.1  Gas Ionization Detectors
     As radiation passes through a gas, it gives energy to orbital electrons,
causing ionization and excitation of the ges atoms through the mechanisms
described in Chapter 1.  Gas ionization detectors use the process of ionization
to detect the presence of radiation.
     A.  Principles of Operation.  A simplified diagram of a gas ionization
detector is shown in Figure 2.1.  The detector assembly usually consists of a
power supply and a closed,  electrically conductive cylinder or chamber that is
filled with gas.   The chamber walls are usually made of metal, which can be
                             RESISTOR-
                        JNSUIATOR
                                 POWER SUPPLY
                                                       PULSE
                                                      -CAPACITOR
                                            •= GROUND
                    \
                    ION CHAMBER
                     (CATHODE)
                                • COLLECTING ELECTRODE
                                      (ANODE)
       FIGURE  2.1.   Simplified Version of a Chamber Used to Collect  Ions
                                      2.S

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DARCQM-F 385-1
penetrated by photons and some high-energy  beta  particles.   Trie  chamber may
h>ave a "window" made of a material  such  as  mylar, which  can  be easily  pene-
trated by alpha particles and lower-energy  bete  particles.
     The positive and negative poles  of  the power supply are called electrodes.
A thin wire in the center of the chamber is connected  to the positive  electrode
of the power supply and is called the central  collecting electrode, or anode.
The chamber wall i: connected to the  negative  electrode  of  the  power supply  and
is called the cathode.  As radiation  passes througn  the  gas  that fills the
chamber, it gives energy to the orbital  electrons of the gas atoms and may
cause them to be removed from the originally neutral gas atoms.   This  ioniza-
tion process results  in the formation of a  free  electron (negc'.ive ion) and  a
positive gas atom (positive ion), which  together are called  an  ion pair.
Repeated interactions between radiation  and the  fill gas in  a closed chamber
gradually cause the degradation of the gas  until eventually the  detector loses
its effectiveness, and either the degraded  gas is  removed from the chamber and
replaced with new gas, or the entire  detector is replaced.
     The number of ion pairs created  in  a given  volume of the chamber's fill
gas depends on the type of gas used and  the type and energy of the radiation.
A dense gas has more  atoms for the radiation to  interact with than does a less
dense gas and thus leads to tht-creation of more ion pairs.   Alpha particles,
which are relatively  heavy and slow and  have a double positive charge, create
many ion pairs within a very short distance as they travel  through the fill
gas.  They typically  give up all of their energy to the gas within a few centi-
meters.  Beta particles, which are much  smaller  and faster than alpha particles,
do not interact as readily with the orbital electrons and thus create fewer ion
pairs.  Gamma rays and x rays, which  are uncharged and have negligible mass,
interact indirectly with the gas (see Chapter 1) and produce even  fewer ion
pairs.  If an alpha particle, a beta  particle, and a gamma ray with identical
energies passed through the same volume  of a fill ges, the alpha  particle would
create tens of thousands of ion pairs, the beta  particle a few hundred  ion
pairs, and the gamma  ray just a few ion  pairs per centimeter of gas.  The num-
ber of ion pairs created also depends on the energy of the  radiation.   On the
average, one ion pair is produced for every 30  to 35 eV of energy transferred
                                     2.10

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                                                               DARCO.M-P 385-1
to the gas.  Thus, a single 1-MeV radiation that loses all  of its energy in a
gas creates approximately 30,000 ion pairs; a 2-MeV particle creates 60,000
ion pairs.
     When voltage is applied across the chamber, the ion pairs produced in the
gas by the incident radiation move toward their respective  electrodes:  the
negatively charged electrons move rapidly to the positively charged anode, and
the positively charged ions, which are much heavier, move very slowly toward
the negatively charged chamber wall.  The electrons that collect on the anode
produce a build-up of cnarge.  The collected charge flows tnrough the external
circuit as a pulse or surge of current.  Each pulse represents the interaction
of one particle or photon of radiation with the gas.  The pulses flowing
through the external circuit of the instrument can be recorded in one of two
ways, depending on the type of electronic circuit used.   If a nonintegrating,
or differential, circuit is used, each individual pulse can be tallied, which
gives a record of the total number of ionizing radiations entering the chamber;
if an integrating circuit is used, the total current flow over a given period
of time can be measured.   The total current flow is proportional to the degree
of ionization in the chamber.
     The magnitude of the voltage applied to the electrodes is another factor
that affects the number of electrons collected on the anode and the resulting
charge.  Figure 2.2 shows the relationship between the applied voltage end the
pulse height in the circuit.  In this figure, six regions can be observed:
1) the recombination region, 2) the ionizetion chamber region, 3) the propo1"-
tional  region, 4) the limited-proportional region, 5) the Geiger region, and
6) the continuous-discharge region.
     (1)  Recombination Region.  In this region, the voltage across the elec-
trodes  is relatively low, and the force of attraction between the ions and the
electrodes is not great.   Therefore, most of the positive and negative ions
produced by the radiation are attracted to each other, rather than to the
electrodes,  and they recombine.  As the voltage applied to  the electrodes is
increased, fewer ions recombine.   Hcvever, no radiation detectors operate in
this  reoion.
                                     2.11

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DARCQM-P 385-1
                     SIMPLE
                    ION IZATION
                           GAS AMPLIFICATION
    CC
    o:
    UJ
    O
    O
  3S
  i±f o
  — uj
    <->
    on
    o
    ce.
    LU
    CO
\ IONIZATION
I  CHAMBER
\   REGION
                              3

                         PROPORTIONAL
                             REGION
LIMITED
PROPOR-
TIONAL
REGION
GEIGER
REGION
   6

 REGION OF
CONTINUOUS
 DISCHARGE
                                      VOLTAGE

       FIGURE 2.2.  Relationship  Between  Applied  Voltage  and the Number
                    of Electrons  Collected  on  the Anode

     (2)  lonization Chamber  Region.  At  a  certain voltage,  the force of
attraction between the ions and the  electrodes is-sufficient to cause all of
the electrons produced by the incident  radiation  to be collected on the anode.
Subsequent moderate increases in  the voltage do not create any further
increase in the electron current:  a saturation voltage  has  been reached.   (For
this reason, the  ionization chamber  region  is  also celled the saturation
region.)  The number of electrons  collected at the anode  is  a function of the
amount of ionization occurring in  the chamber.

     Figure 2.2 shows three curves in the ionization chamber region, one each
for alpha particles, bete particles, and  gamma rays (photons).  Because alpha
particles create  £ larger number  of  ion pairs per path length than the other
                                      2.12

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                                                              DARCCK-? 385-1
radiations do, more electrons are collected  on  the  anode  and  e  larger  pulse  is
produced in the external  circuit.  The pulse height for beta  particles,  wnich
create fewer ion pairs than alphas,  is slightly smaller,  and  the  pulse height
for gamma rays is the smallest.   Thus, in  the ionization  chamber  region,  the
different types of radiations can be distinguished  from each  other  because of
the different pulse heights produced in the  external  circuit.
     (3)  Proportional Region.   If the voltage  between the  anode  and the
cathode is increased, the number of  ion pairs collected is  larger than the
number of primary ion pairs (those initially formed by the  incident radiation).
At high voltages, the primary negative ions  (i.e.,  electrons)  are accelerated
toward the anode fast enough to  cause additional  ionization of the  gas,  creat-
ing secondary ion pairs.   The secondary electrons that are  then accelerated
toward the anode may also have enough energy to cause even  further ionization
of the gas.  This multiplication or  avalanche of electrons  moving toward the
anode is called gas amplification, and in  the proportional  region the  avalanche
is restricted to the vicinity of the primary ionizations.   The gas  amplifica-
tion factor, or multiplication factor, is  a.  measure of the  number of secondary
electrons produced by one primary electron.   Thus,  if one primary electron
causes 10,000 secondary electrons to be produced, the multiplication factor  is
10,000.  (In the ionization chamber  region,  the multiplication factor  is 1
because the relatively low voltage across  the electrodes  does not result in  an
avalanche, or multiplication effect.)  In  the proportional  region,  the total
number of ion pairs eventually formed is proportional to  the  number of primary
ion pairs formed by the incident radiation,  and the multiplication factor is
constant over small voltage ranges within  the region.  Detectors  operating  in
the proportional  region have multiplication  factors up  to 10  , depending on
                                                 4
the applied voltage, but  the typical factor  is  10 .  These  detectors,  like
those operating in the ionization chamber  region, can distinguish between
alpha, beta, and gamma radiations.
     (4)  Limited Proportional Region.  At the  upper range  of the proportional
region, the gas amplification factor is no longer constant  for a  given voltage
rer.ae but can change markedly with small changes in the  applied voltage.  This
region is celled the 1imited-proportional  region and, in  general, has  no useful
purpose for radiation measurement.
                                     2.13

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DARCO.M-P 385-1
     (5)   Geiger Region.   A further  increase  in voltage leads to the Geiger
region.   The gas amplification  in  tnis  region  is so extensive tne~. an ava-
lanche of electrons  spreads alor.g  the entire  length of the instrument's anode,
and all  pulses are the  same size,  regardless  of the type of radiation that
initiated the ionization.   Thus, a detector operated  in the Geiger region can-
not distinguish between the different types of radiation.  The  pulses in the
Geiger region are much  larger than those  in any of the previous  regions.   In
fact, the production of only one primary  ion  pair  results  in an  easily  measur-
able pulse (^-1 V).'
     As positive ions approach the cathode wall of the detector,  tney have  so
much energy (because of the high voltage  in the Geiger region)  that  they  attract
electrons from the wall and oecome neutral atoms.  During  this  process, a  low-
energy x ray is often emitted that can  cause  further  ionization.   If this  addi-
tional ionization were allowed to  proceed, the detector would  remain in a
continual state of discharge and would  not count  a second  pulse.   To terminate,
or quench, the perpetual ionization  in  the detector,  a small  amount  of  quench-
ing gas is added to the chamber.   The quenching gas  transfers  its electrons to
the positive ions, and the electrons and positive ions combine to create
neutral gas atoms.  Without its electrons, the quenching  gas  has a positive
charge; it migrates to the cathode and collects electrons  to  become neutra-
lized.  The energy produced in this  process  goes  into the  dissociation  of the
gas molecule rather than the production of an x ray.   Bromine, chlorine,
ethanol, and methane are typically used as quenching gases.
      (6)  Continuous-Discharge Region.   If the voltage is increased still
further, arcing occurs across the electrodes, and pulses are registered-
continuously even if no radiation is present.  Instruments operated in this
region can  be permanently  damaged in a short  time.
      The three  types of ionization instruments commonly used by radiation
protection  personnel--ionization  chambers, proportional  counters, and  Geiger-
Mueller counters—correspond to the three regions of the  pulse  height-voltage
curve in which  radiations  can be  detected.
      B.   Ionization Chambers.  Instruments designed  to operate  in the  ionize-
tion  chamber  region of  Figure 2.2 are  called  ionization chambers, or ion
chambers.   They can be  passive or active  detectors.
                                      2.14

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                                                              DAKCOM-P 385-1
      (1)  Passive Ion Chambers.  In a passive ion chamber,  e  voltage is  placed
across the electrodes in a process called charging.   The chamber is then
separated from the charger and placed in a radiation field.   The ions formed by
the  incident radiation neutralize the charge, and the subseouent drop in
voltage can be measured and correlated to the amount of radiation that was
present.  Two types of passive ion chambers are pocket ionization chambers  and
condenser chambers.
      Pocket ionizetlon chambers, also called pencil  dosimeters,  are integrat-
ing  instruments that record the total current flow,  or true charge, produced
by the radiation entering tne chamber.  These dosimeters have a  metal-coated
quartz fiber that is attached at one end to a rigid  metal  electrode and  sus-
pended in a small gas-filled chamber.  When a positive charge is placed  on  the
electrode, the charge is also transferred to the fiber, and because like
charges repel, the fiber moves away from the electrode.  When radiation
ionizes the fill gas in the chamber, the resulting negatively-charged elec-
trons combine with and neutralize some of the positive charges on the fiber
and  electrode (the fiber and electrode are said to discharge).  This results
in a  decrease in voltage between the two, and the fiber moves closer to  the
electrode.  How far it moves depends on the number of electrons  formed by the
radiation; thus, the distance between the electrode  and the fiber indicates
how much radiation the dosimeter was exposed to.
     Self-reading pencil dosimeters are equipped with a built-in microscope and
a scale that enables the wearer to read the exposure at any time.  When  the
dosimeter is fully charged, the fiber lies on the zero point on  the scale.   As
the fiber discharges in response to ionizing radiations, it moves along  the
scale.  Non-self-reading pencil dosimeters must be inserted into a specially
designed voltmeter to be read.   If a dosimeter is dropped or subjected to
other sudden motions, it may discharge and incorrectly indicate  a very high
exposure.
     Another type of passive ion chamber, the condenser chamber or condenser
R-meter,  is  used to  make highly accurate and precise measurements.  Condenser
chambers  are similar to non-self-reacing pocket ionization chambers but are
very carefully constructed and  have walls of uniform thickness so that the
                                     2.15

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DARCQM-P 385-i
energies of incident photons can be measured.   These  instruments  also  respond
to beta rays with energies higher than 1 MeV.   If  the  inside of a  condenser
chamber is coated with boron, it also responds  to  thermal  (low-energy)
neutrons.
     (2)  Active Ion Chambers.   Active ion chambers have  a  built-in  voltage
source.  The circuits in tnese chambers can be  nonintegrating,  registering 6
pulse for each particle or photon of radiation  that interacts with the  fill
gas, or integrating, measuring the total current produced by the  ionizations.
     The most popular use of active ionization  chambers  is  as portable  instru-
ments to survey for beta and gamma radiation.   These  instruments  come  in
various forms, shapes, and sizes, but the most  common  type  is the pistol-
sheped, portable rate meter known as the "Cutie Pie."  Most of  these survey
instruments are thin-window ionization chambers that,  have a removable  shield
over the window end of a cylindrical chamber.   When the  shield  is removed,  the
instrument responds to both beta and gamma radiations, but when the  shie",d  is
in place,  only the gamma rays can penetrate it  to  enter  the chamber.  There-
fore, to get a correct beta reading, it is necessary  to  take two  readings,  one
with the shield on and one with it off.  The shield-on reading  is then sub-
tracted from the shield-off reading to give the beta  reading.
     Active ion chambers can also be used to measure  alpha particles.   A
chamber for this purpose is usually designed so that  the alpha  source can be
placed inside the cylindrical chamber.  Because the chamber completely
surrounds  the source, which is emitting particles  uniformly in  all directions,
all of the alpha particles emitted from the source deposit their energy within
the chamber.  This type of counting system is an illustration  of 4ir  geometry
and results in a near-100» counting efficiency.
     C.  Proportional Counters.  A proportional counter  is a  gas  ionizstion
detector that is operated in the proportional  region  of  the pulse height-
voltage curve (see Figure 2.2).  The anode, or collecting electrode, is a loop
of very thin wire (approximately 0.025 mm) that is usually made of fine, clean
tungsten with minimal sur-'-ce irregularities.   The cathode, or outer sheath of
                                     2.16

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                                                              DARCOM-P 385-1
the cylindrical chamber, is either metallic  or metal-  or  carbon-coated glass.
Detectors operating in the proportional  region can  have eitner  nonintegrating
or integrating circuits.
     A mixture of 10* methane and 90C* argon,  known  as  P-10  gas,  is  commonly
used as the fill  gas in proportional  counters.  A mixture of  4*  isobutylene
and 96« helium can also be used.   These  gases provide  stable  operation and
high gas amplification.  Air is rarely used  as the  fill gas because oxygen
easily captures electrons before  they reach  the anode,  reducing  gas
                t
amplification.
     The proportional counters used today are either gas-flow or sealed.   In
gas-flow proportional counters, gas flows through the  counting  chamber at a
very low rate, removing the degraded gas and any contaminants.   Because of the
continual replacement of the fill gas, these detectors  have a long  life.
Sealed proportional counters have a finite life because the fill gas,  which is
sealed inside the counting chamber, degrades over time as incoming  radiations
interact with it.   However, the chamber  can  be emptied and  completely  refilled
with new counting gas.
     (1)  Gas-Flow Proportional Counters.  Before a gas-flow  proportional
counter is operated, residual air and contaminants  must be  removed  with a
brief, large flow of counting gas.  This process is called  purging.  The
chamber of a simple gas-flow proportional chamber is hemispherical  or  some-
times cylindrical.  The radiation source is  typically  positioned at the bottom
of a hemispherical chamber or in  the middle  of « cylindrical  chamber.   If the
source is suspended in the chamber, 4n geometry is  achieved.   If the source is
positioned at the bottom of the chamber, the device is referred to  as  a 2r
counter.
     Wlndowless gas-flow counters are used to count alpha and beta  particles.
Because alpha particles have a much higher specific ionization than beta
particles (they form many more ion pairs per path length  as they move  through
the fill gas), the large pulses of electronic charge that result from alpha
interactions with the fill  gas can be electronically distinguished  from the
smaller beta pulses by adjusting  the operating voltage,   if the count rate
versus the operating voltage is plotted, two plateaus  are observed  (see
                                     2.17

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DARCOM-P 385-1
Figure 2.3).  At low voltages, only the alpha particles produce pulses because
they a-e more energetic and more highly ionizing than the beta particles.
This portion of the curve is called the alphc plateau.  If tne applied voltage
is increased past the alpha plateau, the counting rate begins to increase
as gas amplification is caused by increasing numbers of bete particles.  After
a transition region, another plateau is reached that represents the pulse
created by alpna and beta particles together.  This plateau is often referred
to as the beta plateau.  Because beta particles vary widely in their energies,
the beta plateau i's not as flat as the alpha pUteau.
     Alpha particles on surfaces can be detected using a specially designed
gas-flow proportional counter.  The detector is flat and has a window made of
aluminized mylar.  The counting gas is frequently propane, which is attached to
the counter in small, interchangeable metal bottles.  This survey  instrument
is especially useful in areas where alpha surveys are required and gamma
radiation levels are high (50 to 500 mR/hr), because  it can discriminate
against the smaller pulses produced by gamma rays.
     (2)  Sealed Proportional Counters.  A specially  designed sealed propor-
tional counter can be used to detect and measure low-energy  (thermal)  neu-
trons.  Neutrons do not interact directly with the  orbital electrons of  the
                 r/A
           20,000 -
        o
        CJ
        <   10,000
        3
        o
        o
                                                    BETA PLATEAU
                                                  x
                          ALPHA PLATEAU

flL//
                     |/
800     1000    1200     1400     1600
             COUNTER VOLTAGE (VOLTS)
                                         180D
                                                                2000
            FIGURE 2.3.  Plateaus for Typical  Proportional  Counter
                                     2.18

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                                                               DARCOM-P 385-1
fill gas  (they are not directly ionizing radiation—see Chapter 1).   There-
fore, the detection and measurement of neutrons relies on the interaction of
the neutrons with some materiel capable of causing ionize:-ions.  The  boron
trifluoride (BF,) gas proportional counter is the most commonly used  instru-
               J
ment for  this purpose.  Low-energy neutrons interact with the boron to form
alpha particles, which can then ionize the gas.  The BF., counter can  also be
used to measure high-energy (fast) neutrons.   For this, the chamber is wrapped
in polyethylene, paraffin, or some other hydrogen-containing material that
slows down  (reduces the energy of) the incident neutrons.  These instruments,
often referred to as rem meters, have the advantage of being insensitive to
most other  types of radiation.  The small pulses produced by gemma rays can be
discriminated out electronically.
     D.   Geiger-Mueller Counters.   Geiger-Mueller (GM) counters are gas
ionization  detectors designed to operate in the Geiger region of the  curve in
Figure 2.2.  They can be used as pulse counters in the laboratory or  as
portable  survey instruments to detect alpha,  beta, and gamma radiation.
However,  they cannot be used to distinguish between the different types of
radiation because all of the pulses produced  in the Geiger region of  the pulse
height-voltage curve ere the same  size.
     The  detector itself is a stainless-steel tube that contains the  fill gas
(usually  argon) and the anode and  that may have an end or side window.  Pulses
are electronically transmitted to  a counter or a meter, and the readout is
generally given in cpm.  Some GM instruments  are designed to read out in mR/hr
to R/hr in  response to gamma rays  with energies between 60 keV and 1.5 MeV.
However,  these instruments should  not be used as dose rate or exposure rate
meters because they produce pulses of the same size regardless of the energy
of the phtons  causing the ionization.   True dose rate meters give a response
that is related to the energy of the photons.
     Wall  and  window thicknesses,  which are expressed in mg/cm ,  ' range
             2
from 30 mg/cm   (for counting gamma rays and high-energy beta particles) down
                   2
to 0.4 to  1.4  mg/cm  (for counting alpha perticles and low-energy bete
                    ?                                  ^
(a) Thickness (mg/cm )  = density of the material  (mg/cm ) x linear
    thickness (cm).
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DARCO>'j-P 385-1
particles).  One of the more popular GM survey meters  uses  a  tube (10 cm long
and 2 cm in diameter) encased in a stainless-steel  housing  that  contains a
window.  The window can be opened to admit beta particles and gamma rays, o>-
closed to admit only aemma rays.  Thus, the beta contribution to the radiation
field can be determined with this instrument.
     For monitoring alpha and beta radiation, a "pancake" GM  tube is used.  The
detector is a flat, round cylinder with a large window that is approximately
                          2
5 cm in diameter and 16 cm  in total areas.  The thickness  of the window is
                2 '
1.4 to 2.0 mg/cm .  The detector is sensitive to alpha radiation with energies
aoove 3 MeV and to beta radiation with energies above  40 keV.  In addition,
the detector has a shield (usually made of tungsten) over all surfaces except
at the window location, to reduce the  influence of gamma radiation.  To pro-
tect the thin window, a wire screen is sometimes provided.
     Portable GM survey meters can be  equipped with either  a  head set contain-
ing ear phones, or a speaker attached  to the instrument case.  Each time a
pulse is recorded in the counting circuit, a click is  heard.   These devices
are extremely useful in surveying for  radiation because their response is much
fester than the meter indication.  The audible circuit is separate from the
meter circuit and does not fail even if the device saturates  and the meter
indicates zero.
     Geiger-Mueller counters are probably the most widely used and versatile
instruments for detecting radiation.   They are inexpensive,  easy to operate,
sensitive, and reliable.  However, their use in or near very high  radiation
fields requires caution because most counters saturate in such a field.   The
incident radiation enters the sensitive volume of the tube at such a  rate  that
the tube is in a stcte of continuous discharge, and the count rate circuit
fails to function properly.  As a result, the meter begins to respond but  then
fells off and reads near zero rather than off the high end of the  scale.   A
person entering a very high radiation  area might not  realize it  because  the  GM
had failed.
2.2.2  Scinti'lction Detectors
     Shortly after x rays were discovered, researchers found that  certain
materials fluoresce, or emit visible light, when struck  by radiation.   These
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                                                              DARCOM-P 385-1
material: are  referred to es phosphors,  -or scinti"Meters.   Scintination
detectors were among the earliest instruments for detecting and  measuring
ionizing radiation and they are still widely used today.
     A.  Principles of Operation.  As radiation enters  and passes  through a
phosphor, it gives up its energy to electrons in the phosphor b\  both ioniza-
tion and excitation.  Excited electrons  move into defects, or gaps,  in the
atomic structure of the phosphor, called traps.  When the  electrons  escape
from the traps to return to lower energy levels, the excess energy is released
in the form of visible light.  This process is called scintillation, and  the
light fU'hes  produced are called scintillations.
     The light flashes generated in the  phosphor can be detected and related
to the Incident radiation by means of a  photomultiplier tube, which  is a  com-
bination of a  photocathode and an electron multiplier.   A  photocathode con-
verts fleshes of light (light photons) into electrons by the photoelectric
effect (see Chapter 1).   An electron multiplier multiplies the number of
electrons using a series of electrodes,  called dynodes, which are  positively
charged.  The electrons  from the photocathode are accelerated to the first
dynode by the application of enough voltage to cause multiple emission of
secondary electrons at the first dynode.  The secondary electrons  are then
accelerated to subsequent dynodes, resulting in further multiplications.   The
typical voltage between  each multiplying stage is 50 to 250 V, with  each
dynode having a more positive voltage than the preceding one.  After the  last
multiplying stage, the electrons are collected at the anode of the photomul-
tiplier tube and fed tc  ?.n external circuit in the form of a pulse.   Photo-
multiplier tubes typice.iy have a gain,  or multiplication, of 10 .  That  is,
the number of electrons  released by the  photocathode is multiplied a
million times by the time all of the electrons reach the last dynode.
     The output current  from the photomultiplier tube is then detected and
analyzed by the electronic circuit.  The extent of the electronic circuitry
depends upon the application of the system.  A simple circuit, used  simply to
detect radiation,  consists of a battery-operated power supply anc an amplifier
with a pulse shaper and  a rate meter. However, when the device is used for
analyzing the energies of the photons emitted by a radioactive materiel,  the
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DARCQM-P 385-1
circuit includes a pulse height analyzer,  a  sealer, and other equipment.  A
pulse height analyzer sorts the detector signals,  or  pulses, by size and
stores them in appropriate pulse height channels.  The size of a detector
signal, and thus the channel to wmch it is  assigned, corresponds  to the
energy of the incident photon.
     A single-channel analyzer can analyze only one channel at a time;  that
is, it can count the number of pulses within a  size limit  that is  manually  set
(using upper- and lower-level discriminators) on the  face  of the analyzer.
If, for example, the lower-level discriminator is  set to  reject pulses  below
50 kV and the upper-level discriminator is set to  reject  pulses above  60  kV,
only those pulses within the 50- to 60-kV range will  be counted.   By  starting
at the lower end of the scale and going upward, an operator can identify  which
cr.annels have the greatest number of counts, or peaks.   Each peak  corresponds
to photons of a specific energy, which in turn correspond to specific  radio-
nuclides.  This process is called spectrometry.
     A multichannel analyzer has up to several  hundred  or several  thousand
single-channel analyzers automatically sorting pulses into specific channels.
The data that is accumulated is displayed as a plot with  channel  number (or
photon energy) on the x axis versus the number of counts  in  a  specific channel
on the y axis.  This plot i.s called a spectrum.  Display  modes  include
oscilloscope screens, x-y plotters, and electric typewriters,  which type  out
channel numbers versus counts.  Because each radionuclide has  its  own distinct
spectrum, spectrometry can be used to identify unknown  radionuclides.
     B.  Inorganic Scintillators.  Inorganic scintillators are inorganic  (not
carbon-containing) salts that form regular crystalline  lattices.   These
lattices contain small amounts of impurities that activate the scintillation
process (that is, they cause the crystal to emit light  when it is exposed to
radiation).
     Crystals of the alkali  halides  (e.g., sodium iodide) are  the most widely
used class of scintillators.  Sodium iodide  (Nal) is  a dense material with
wrich gamma rays interact readily.   Crystals of this  material  are activated
for scintillation by the deliberate  inclusion  of a trace amount of thallium
(Tl).  These crystals, which can be  used to  detect gamma and x radiation, can
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                                                              DARCON-P  385-1
be produced  in a solid cylinder or shaped like a  well.   T"e well  shape  is
formed from  a crystal with a hole drilled part way  into  it; small  vials or
cylindrical  samples that are placed in the well  are,  in  effect,  surrounded by
the crystal, a configuration that results in the  detection of most of the
emitted radiation.
     Sodium  iodide crystals are very effective for  high-efficiency analysis of
gamma-ray spectra.  However, these crystals have  a  relatively poor energy
resolution;  that is, they cannot easily distinguish between, or separate,
photon peaks of slightly different sizes.  They are therefore cr.  limited use
in distinguishing between radionuclides that emit gamma  rays of very similar
energies.
     Zinc sulfide (ZnS), another inorganic salt,  is activated for scintilla-
tion by the  inclusion of silver (Ag) and is used to detect and measure  heavy
charged particles, such as alpha particles.  A zinc sulfide crystal must be
about 20 ym  thick in order to detect alpha particles.  If the material  is
thicker or thinner than this, its detection efficiency decreases.   In portable
alpha survey meters, the zinc sulfide can be applied to the back of a thin
window or sometimes painted right on the face of the photomultiplier itself.
     When large areas or large volumes of a scinfillator are needed, as in
whole-body counters, the use of inorganic crystals  involves high cost and con-
siderable handling problems because the crystals must be protected from thermal
and mechanical shock.  These problems can be minimized by the use of organic
scintillating materials.
     C.  Organic Scintilletors.  Organic scintillators contain carbon,  which
combines readily with hydrogen and oxygen.  These scintillators have a low
atomic number and a relatively low density, which makes them suitable for beta
counting and, in the case of liquid organics, for alpha counting  (the density
is too low for high-efficiency counting of gamma rays).   Organic scintillators
can take the form of solid crystals, liquids, or plastics because the scintil-
lation process arises from a transition in the energy level of a single
molecule, and the t.snsition does not depenc on the physical state of the
scintillator material.
                                     2.23

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 DARCOM-P 385-1
     (1)  Organic Crystals.  The two most  common  organic  crystalline  scintil-
lators are anthracene (C-,4H,Q) and stilbene  (Ci^H.,).   Anthracene  has  the
highest efficiency for light output of any organic  scintillator, but  both
materials are fragile and difficult to obtain  in  large  sizes.   They can  be
used to detect high-energy beta particles, but low-energy betas  are either
self-absorbed or absorbed by the surroundings  before  they can  interact with
the crystal.  To overcome this problem, liquid organics can  be used.
     (2)  Liquid Organic Sc .ntillators.  Liquid scintillators  are  made by
dissolving an organic scintillator material,  called the solute,  in an  organic
solvent.  The radioactive source, or sample,  is then  dissolved in  the
solution.  Because all the radiations emitted  by  the  sample  must pass  through
some portion of the scintillator solution, counting efficiencies can  approach
100%.  This method is particularly advantageous for counting low-energy  beta
emitters, such as  H and   C, and can also be  used  for  alpha emitters.
     The scintillator solution, which is often called a cocktail.  consists  of
the radioactive sample, the organic solvent,  a primary  scintillator solute
(primary fluor), and sometimes a secondary solute (secondary fluor) and  a
solubilizing agent (diluent).  The solvent,  which is  often toluene, xylene, or
dioxane, absorbs most of the energy of the beta particles through  particle
interactions (see Chapter 1) and transfers it  to  the  primary fluor.   The
primary fluor is made up of large organic  molecules,  such as p-terphenyl or
PRO (chemical name:  2,5-diphenyloxazole), that scintillate  after  they have
received the excitation energy from the solvent.   The concentration of the
primary fluor in the cocktail is usually about 1%.   The secondary  fluor
absorbs the light emitted by the primary fluor and  re-emits  it at  a somewhat
longer wavelength, which is closer to the  wavelength needed  for optimum  opera-
tion of the photomultiplier tube.  A diluent such as  a  hydrocarbon, ether,  or
alcohol may be added to the cocktail if the  radioactive sample does not  readily
dissolve in the solvent.
     Although diluents favorably change the  character of  tne solvent, they
also decrease the counting efficiency by interfering with the  transmission  of
light tc the photomultiplier, as does the  introduction  of the  radioactive
sample itself.   This interference, known as  quenching,  may limit the  amount of
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                                                              DARCO.M-P 385-1
a radioactive sample that can be effectively incorporated  into the solution.
Examples of diluents that are effective but that have  a  strong quenching
action are phenols, amines, aldehydes,  and nitro- and  iodo-coroounds (com-
pounds containing NCL or iodine),  as well  as colored  substances.   All  modern
instruments for liquid scintillation counting nave electronic  circuitry to
assist in estimating the degree of correction needed  to  account for
quenching.
     After a cocktail is prepared, it is enclosed in  a glass  or plastic vial.
Glass vials should have a low potassium content to reduce  the  background
counts produced by naturally occurring    K, which is  radioactive  (it emits
                                        40
beta particles).  To further reduce the   K background,  glass  vials should
be very thin (and should therefore be handled carefully).   Plastic vials  are
popular because plastic contains no potassium, and the vials  therefore have a
lower radioactive background than glass vials.  They  also  have a  slightly
higher efficiency for  H counting.  The disadvantages  of plastic  vials are
that they are permeable to toluene, a commonly used solvent;  therefore, count-
ing rooms or laboratories I'D which plastic vials are  used  should  be well  ven-
tilated.  Some plastics also swell with time, which may preclude  counting a
sample again at a later date.
     Vials containing the cocktail are  placed in a lightproof enclosure con-
taining one or more photomultiplier tubes.  Quenching  effects, and the fact
that this counting method typically involves low-energy radiations, may pro-
duce pulses that correspond to no more  than a few electrons in the photo-
multiplier tube.  Noise (pulses arising from sources  other than the
radioactive sample) may also interfere  with accurate  and reproducible counting
of the sample.   Significant sources of  noise include  photoelectrons that are
generated by heat production within the photocathode,  and  chemiluminescence,
or additional  scintillations caused by  chemical reactions  in  the cocktail.
     Sources of noise can produce extraneous photoelectrons that are included
in the pulse and are difficult to discriminate against when the primary pulse
(from the rad1c>£Ctive sample) is produced by only a few photoelectrons.  The
practical  counting efficiency of a liquid scintillation counter is determined
by its ability to distinguish between the primary pulse and the noise.
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DARCOM-P 385-1
     The counting interference  caused  by  noise from the photomultip!ier tube
can be eliminated by using two  photomultip! ier tubes placed on differert sides
of the scintillctor vial,  and counting only  those pulses that are observed at
the same time by both tubes.  Pulses arising from only one tube, which would
be noise, are not counted.
     Because of tne efficiency  and  uniform  geometry of liquid scintillation
counting, its most common  application  with  respect to alpha particles  is for
counting low-activ,ity environmental  samples.  The relatively high-energy
alphas have a much higher  light output than  the  low-energy betas, and  noise
interference is not a problem.   The energy  resolution, however,  is  poor
compared with the resolution  that can  be  achieved using the semiconductor
diode detectors discussed  below.
     (3)  Plastic Scintillators.   Plastic scintillators are solid organic
solutions that are sometimes  used for  beta  counting.  They can  be made much
larger than organic crystal scintillators and are easily handled  and  shaped.
A disadvantage that limits their use,  however,  is that  they have  much  lower
counting efficiencies than organic crystals  of  equal  size.
2.2.3  Semiconductor Detectors
     A semiconductor, or solid-state detector,  is a  solid  crystalline  material
that has an electrical conductivity between that of  insulators  (nonconducting)
and good conductors such as metals.  Tne electrical  conductivity of the
semiconductor changes, however, when it is exposed  to radiation,  and  the
degree of change is related to  the radiation exposure.   The  semiconductor
deiector operates on the same principle as the  gas  ionization  detector;  that
is, ionizations produced within the sensitive volume of the  detector cause a
voltage pulse within the detector, which is then amplified and counted on a
sealer system.  In the semiconductor detector,  a solid  replaces the fill gas
of the gas ionization detector, and the phenomenon  of gas  amplification (the
production of secondary ions) does not occur.  However, the  voltage pulse from
a gas-filled detector is smaller than the pulse from a  semiconductor detector'
because the solid material in a semiconductor produces  10 times as many
primary ion pairs as does  the gas in a gas-filled chamber.
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                                                              DARCOM-P 385-1
     The ator.s of semiconductor materials usually  have  four electrons  in--their
outermost shell (i.e., four valence electrons);  examples of these  materials
are germanium and silicon crystals.  In the production  of semiconductor
detectors, other elements are added to the semiconductor materials.  These
elements are called impurities because the semiconductor crystf.l  is  no  longtr
pure after they are added.   The introduction of  imparities such  as lithium,
aluminum, or boron, which have three valence electrons, produces  a total  of
seven valence electrons.   An atom with eight valence  electrons  is  very  stable.
A material with a configuration of seven valence electrons has  a  space,  or
hole; because it wants to accept one more electron,  it  is called  a positive  or
p-type material.  If an impurity with five valence electrons, such as  arsenic,
is added to the semiconductor material, the result is nine valence electrons,
or one more than the stable configuration of eight.   In this  case, the  result-
ing material wants to giv<  up its extra electron to  become stable  and  is
called a negative or n-type material, or an electron  donor.
     When n-type and p-type materials are combined,  the extra electrons in  the
n-type materials combine  with the holes in the p-type materials,  creating
electron-hole pairs and forming an electrical  potential across  the junction.
This small potential difference is then enhanced by  applying  an  external
electric field to oppose  the natural motion of the electrons  and holes.   This
"reverse bias" is applied by connecting the positive  pole of  a  battery to the
n side and the negative pole to the p side.  The depletion layer that  is  thus
set up is the sensitive volume of the detector (see  Figure 2.4).   When a
charged particle (alpha or  beta particle) loses  its  energy within this
depletion region, electrons are released and are attracted to the positive
electrode.  This movement produces e current pulse that can  then be amplified
and electronically measured with considerable accuracy.  The  diffused  p-n
junction detector is not  useful for detecting photons because the depletion
layer is only a few millimeters deep.
     The germanium-lithium  detector, or GeLi detector (pronounced "jelly"),
and the silicon-lithium detector, or SiLi detector (pronounced  "silly") are
two examples of semiconductor detectors that operate on the  same principle as
diffused junction detectors but that have a much larger sensitive volume,
                                     2.27

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DARCOM-? 385-1
                               CHARGED PARTICLES
                                ENTER FROM THIS
                                     SIDE
                                        \
                        n-TYPE (NEGATIVE) REGION
         DEPLETION-
          LAYER
K^NXX'iyX-.'vvXvXvXv:-
^mmmtmmz
                            p-TYPE (POSITIVE) SILICON
                                                 METAL ELECTRODE

                  FIGURE 2.4.  Diffused p-n Junction Detector

which makes them suitable for gamma counting.  Lithium is drifted into a p-type
germanium or silicon crystal by heating the crystal  and applying a reverse
bias across it.  A wide layer, called the intrinsic or compensated layer, is
formed where the lithium, which denotes one valence electron, exactly compen-
sates the p-type material.  This is the sensitive volume of the detector, and
thicknesses of more than 1 cm can be achieved.  GeLi detectors must be kept
cold using liquid nitrogen (the detectors are designed to hold this coolant)
because the lithium tends to "redrift" if the crystal is allowed to warm up to
room temperature.  Si Li detectors can be operated at room temperature but they
have a relatively low counting efficiency compared with GeLi detectors because
of their lower density.
     Semiconductor detectors of the GeLi and Si Li type are most frequently
used for gamma-ray spectroscopy.  They have the ability to differentiate, with
a high degree of resolution, among various energy peaks.  Semiconductor
detectors have a lower counting efficiency than sodium iodide crystals.
However, their energy resolution is far better than that of sodium iodide
detectors because of the lone sequence of events that takes place in  the
sodium iodide detector to convert the radiation to light and then to  an
electrical signal.  Semi conductors dezectors are relatively expensive, and
because of their fragile nature and design, they cannot be decontaminated.
                                     2.28

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                                                              DARCQM-P 385-1
                    Section 2.3  CALIBRATION  OF  INSTRUMENTS

     The performance and accuracy of radiatior detection  and  measurement
instruments depend on the design characteristics  of  each  instrument  and on
proper calibration and reliability checks  made during  its use.   Calibration  is
the evaluation of an instrument's response to the type  end energy of rtdiation
it was designed to detect or measure, as well  as  to  any other radiation that
may be present and contribute to the radiation reading.   Calibration also
involves examination of the instrument's electrical  and mechanical  integrity.
The AN/UDM-2 calibrator, which is intended to calibrate tactical  instruments,
should not be used to calibrate instruments used  for radiation safety.
     The extent of a raciological calibration operation at an installation
depends largely on the requirements of the radiation protection organization.
The funds available to a radiation protection office may  limit the availabil-
ity of facilities, calibration sources, and technical  staff  for radiological
calibrations.  If, for any of these reasons,  an  office  is unable to  provide  a
proper calibration program, the RPO should seek  outside assistance from
another command or from a commercial calibration  service, rather than per-
mitting the quality of the calibration services  provided  to  be compromised.
2.3.1  Calibration Sources
     The foundation of a good calibration  program is the  use of standard
radiation sources that have well-defined properties  and are  traceable to  the
National Bureau of Standards (NBS).  Such  sources can be  obtained in three
ways:
 1.  They can be purchased from a vendor.
 2.  In certain cases, the installation's  own sources (e.g., small neutron
     sources) can be shipped to NBS for direct calibration.   Because of  the
     time, cost, and complication in transportation, this procedure is not
     frequently used.
 3.  An intercomparison transfe- standard  ar. be obtained by sending an   ionize-
     tion chamber to NBS for direct calibration  with their primary standard.
     The NBS "certifies" the calibration and accuracy of the instrument  as e
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DARCOM-r 385-1
     "secondary standard."   The  chamber,  which  is referred to as directly
     traceable to NBS,  is  then  used  to  calibrate radiation sources at  its home
     facility, and the  calibrated  sources are used to calibrate the  facility's
     instruments.  Sources  and  instruments calibrated against a secondary
     standard are assigned  an allowable error that is larger than  that  of the
     secondary stanacrd.
     The types of radioactive sources  used to calibrate  instruments  and
dosimeters  vary depending  upon  the needs  of the radiation protection program.
To ensure that the proper  sources  are  being selected, one of the following
standards documents should  be referred  to when  calibration facilities  are
being designed and when calibration  frequencies and  the  types and  strengths  of
raaionuclides suitable  for the  instrument calibration process are  being
determined:   American National  Standards  Institute  (ANSI) Standards  N323-1978,
N42.3-1949,  and N12.11-1978.
     As pert of a routine  quality  assurance program  (see Chapter  15),  the
activity of sources should  be checked  periodically.   Verifying  the activity  of
a source that will be used  as a  radiation standard  requires  absolute counting
methods and the use of  accurate detectors with  known counting efficiencies.
Sources that emit alpha and beta particles can  be  verified by placing the
source in a  gas-flow proportional  counter, thus providing 4ir geometry for  the
counting.  A well-type  ionization  chamber (in which  the  source  is  completely
surrounded  by the detector) is  frequently used  for  standardizing  short-lived
gamma-ray sources.
2.3.2  Calibration Facilities
     Radiation calibration  facilities  should be located  where  the  radiation
background  is low, the  radiation field is well  known,  and conditions are
stable.  Facilities should  be constructed of a  material  that minimizes scatter
and should  be large enough  to allow for good geometry when  calibrating instru-
ments that  measure photons  and  neutrons.   General  criteria  for  facility design
are discussed in Chapter 8.
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                                                             DARCO.M-P  3E5-1
2.5.3  Instrument Characteristics That Affect  Calibration and Calibration
Frequency
     Under certain conditions, the ability of  health  physics  instrumentation
to measure radiation accurately is linrted by  the  equipment and  it?  operating
characteristics.  Some of these concitions create  a  relatively small  error
while others could, if not recognized, put the radiation protection  staff and
radiation workers in jeopardy.  For example, as discussed earlier, a  GM
detector saturates and reads zero in a high-radiatior; field.  As  another
example, a standard ionization chamber o~ten produces a fo'se reading when
used around a source with a three-phase alternating  current  (e.g., a  three-
phase x-ray machine).  An ionization chamber that  is  compensated  for radio
frequency must be used to avoid this problem.
     The size of a source and the distance between'the source and the instru-
ment also affect measurement accuracy.  If the source is not  a point source
and the distance between the source and the detector varies,  corrections for
source size and source-to-detector distance need to  be developed  and used.
Curves illustrating these corrections are supplied by some  instrument manu-
facturers upon request.  If they are not available,  they can  be  generated by  a
qualified health physicist.  An effective calibration program should include
the assignment of proper correction factors for each instrument  type used in
the radiation protection program.  The correction  factors  should  be  based on
the range of sizes of radiation fields and the source-to-detector configura-
tions commonly used for each instrument type.
     One of the primary factors affecting the  accuracy of  any measurement
(either in calibrations or in field use) is the position of the  source
relative to the position of the sensitive volume of  the detector, that is,
whether the entire sensitive volume is being  irradiated.   It  it  is not, then
geometry correction factors must be applied to the instrument readings.  Part
of the contribution to geometry errors is the  difference  in the  radiation
field during actual  use and during calibration.  Exposure  rate  instruments are
usually calibrated in a radiation field of nearly  unifonr,  intensity.  However,
in many actual  field situations, these detectors are used  in  nonuniform fields
(i.e., close to e source) or are not entirely  exposed.  In either of these
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DARCOM-P 385-1
actual-use situations, the response of the instrument can be low by a factor
of 50.  Under normal conditions, underestimate,-, factors  of 10 and above may
occur.
     Limitations associated with the ability of an instrument to accurately
measure both high and low radiation energies are known as the energy depen-
dence of the instrument.  Energy dependence can be caused by many factors.   If
high-energy radiation causes photoelectrons to be emitted from the detector
wall and the instrument reads them, then the total instrument reading is high.
If low-energy radiation is absorbed by the detector wall, then the instrument
reads low.  The energy dependence of an instrument can be evaluated by expos-
ing it to identical exposure rates from NBS-traceable sources that emit dif-
ferent photon energies.  An instrument correction factor for a given energy
can be calculated by dividing the measured exposure rate by the true exposure
rate.  Curves of correction factors versus radiation energy are usually avail-
able from the instrument manufacturer.
     If the measurements made with an instrument vary significantly when the
instrument's position is rotated through a radiation field, the instrument is
considered to have angular dependence.  Angular dependence may cause serious
discrepancies in instrument readings, particularly if the instrument is not
properly positioned in the radiation field.  If angular dependence appears to
be a problem for an instrument, the instrument should be calibrated at  15°
increments in a full 360° plane perpendicular to the source.
     During the calibration process, portaole survey meters should be tested
to ensure that they respond only to the type of radiation they are designed to
detect.  That is, alpha or neutron monitors should be verified to be insensi-
tive to photon radiation.  Similarly, photon monitors such as ionization
chambers should be insensitive to other forms of penetrating  radiation  such
as neutrons.  Also, scintillation detectors should be closely checked with a
high-intensity light source to verify the absence of light leaks  that could
produce a false count.
     The frequency and extent of routine instrument calibrations  are governed
by many factors, including the rate at which components  in each  instrument age
or become damaged.  The ANSI standards listed earlier in this section describe
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                                                              DARCOM-P  385-1
 the process used  in establishing calibrations frequencies.   They else describe
 procedures for  simple constancy checks to be used between calibrations.
      Section  2.4   FACTORS THAT AFFECT THE SELECTION AKD USE Or RADIAT10N-
                   K'jHITORlNG INSTRUMENTS

      Individuals who are selecting instruments for radiation monitoring should
 know  the  purpose for which the instrument will be used, the degree of accuracy
 needed, the type of radiation to be detected or measured, the energy of the
 radiation, the  source form (whether solid, liquid, or gaseous), and the inten-
 sity  and  uniformity of the field to be measured.  Knowledge of these parameters
 and of the limitations of various types of radiation detection and measurement
 devices will  ensure the selection of the best Instrument for each application.
 Each  facility should have on hand a 11st of available radiation survey Instru-
 ments, Including the types of Instruments available and, for each type, the
 number available,  the radiation 1t detects, Us sensitivity and range, the
 thickness of  any windows, and the general use it was designed for,  This list-
 Ing,  together with the calibration date on each Instrument, can assist in the
 selection of  the best available Instrument for each situation.
      Several  of the factors that should be considered 1n the selection and use
 of radiation  monitoring Instruments are discussed briefly below,
 2.4.1  Detection Versus Measurement
      The purpose for which an Instrument will  be used and the accuracy required
 dictate which Instrument should bt selected,   An Instrument designed only for
 detection should not be used to measure radiation dose rate or exposure rate,
 2,4,2  Type of Radiation
     A principle factor 1n the selection of an Instrument 1$ the type of radia-
 tion to be detected or measured.   For example, a specially designed GK counter
 can detect alpha,  beta,  and gamma radiation,  but a portable alpha counter that
 1s property callbrtted should not measure gamma radiation.  A standard ion
 chamber measures'beta  and  gamma  radiation but  does not detect neutrons.  A rem
meter detects  neutrons but does  not detect external  alpha particles.  If an
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DARCOM-P 385-1
instrument is sensitive to several  types  of  radiation, either mechanical
devices (shields or filters)  or electronic discriminators  can be  used  to  dis-
tinguish between the various  types  of radiation.
2.4.3  Radiation Energy and Instrument Energy Dependence
     The instrument selected  must be capable of measuring  the radiation in
question.   Most instruments are designed  to  respond  to a wide energy spectrum
(e.g., 150 eV to 3 MeV).   However,  a GM counter or an  ionization  chamber can-
not, monitor a substance such  as tritium;  the weak beta  radiation  (18.6 keV)
emitted by tritium requires measurement by liquid scintillation methods or
special windowless counters.
     The most reliable method of determining whether an  instrument operates
accurately in the energy range of a specific radionuclide  is to attempt to
calibrate it against the radionuclide.  Because each instrument will respond
differently, it is useful to provide calibration curves,  especially for beta
calibration.
2.4.4  Nonuniform Fields
     The quantification of radiation exposure rates  from nonuniform fields may
require special calculations and the use of correction factors.  Nonuniform
fields can be expected when measuring 1) dose rates  at the surfaces of mate-
rials, 2) plane circular sources that are smaller than the diameter of the
detection chamber, 3) surface-contaminated cylinders such as rods,  pipes, and
cables, and  4)  radiation beams smaller than the diameter of the  detection
chamber.  Correction factors for these special conditions may  range from 1 to
over 100 depending upon the condition, the type and energy of  the  radiation,
and the particular instrument being used.  Special studies and consultation
with experienced health physicists may be needed.
2.£.5  Angular  Dependence
     If the  direction from which radiation arrives at an  instrument differs
significantly from the directions used in the  calibration field, correction
may be necessary.   Instrument response may be  extremely directional for  some
instruments  and radiations; for others,  directional effects may  be relatively
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                                                              DARCOK-P 385-1
insignificant.  Radiation protection personnel  should  be  alert  to the  poten-
tial for directional  response and should provide  corrections  if necessary.
2.4.6  Calibration
     The selection of an instrument should be based  on the  instrument's  demon-
strated capabilities, including its ability to be calibrated.   Before  any
instrument is placed in field service,  a thorough calibration and operational
check should be performed, including verifying that  batteries are fully
charged.
2.4.7  Unwanted Respor.se
     A portable survey instrument's response to stimuli other than the radia-
tion it is supposed to measure constitutes what is called unwanted response.
Instruments may respond to hr-2t, light, radio frequency radiation, and mechan-
ical shock.  When used near operating equipment,  particularly vehicles with
generators or alternators, survey instruments may respond to induced electrical
fields.  In some instances, components  of an instrument (other than the detector
itself) may respond to radiation, causing measurement errors.  This response
is called extracameral sensitivity.
            Section 2.5  TYPES OF RAD IATION-MONITORING INSTRUMENTS

     Radiation-monitoring instruments are generally classed in one of four
areas, depending upon their particular application:  1) portable survey
meters; 2) laboratory counting instruments; 3) air-monitoring equipment; and
4) other fixed instruments.  The uses of these four classes are discussed
below.
2.5.1  Portable Survey Meters
     Portable survey meters are instruments small and light enough to carry
from place to place.  Some are used for detecting radiation and radioactive
materials, and ethers for que titctivily measuring radiation 'ieve"is.  In both
cases, some degree of accuracy and precision must be sacrificed to provide the
light weight, small size, and ruggedness necessary for portable instruments.
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DARCQM-? 385-1
For measurements of very low levels  of  activity, such as many measurements  cf
environmental  samples,  or for measurements  requiring a high level of  accuracy,
laboratory conditions  and laboratory counting  equipment should be used.
     A.   Portable Detection Instruments.   Portable  survey meters for  detecting
radiation or radioactive materials  (e.g.,  SM counters) should be selected based
upon the type, energy,  and intensity of the radiation to be encountered.  Most
portable detection instruments are  count  rate  instruments.  They frequently
incorporate a  meter display and an  aural  output, using earphones or a speaker
or both.  For  surveys  of areas, equipment,  or  personnel, the aural output
should be used if it is available because  the  aural circuitry of these instru-
ments responds more rapidly to radiation  increases  than does the meter circuitry.
Small radioactive spots or beams can be more readily detected by sound than by
observing the  meter movement.  In addition, the  aural circuitry does  not  fail
if the device  saturates and the meter indicates  zero.
     Even though portable survey instruments are relatively small and rugged,
they must be handled and used carefully to prevent  damaging them while still
effectively detecting  radioactivity.  Most instrument detectors or  probes  have
a very thin window or  covering over the sensitive detector area or  the probe.
Puncturing this window may cause an implosion  in some detectors  (GM  tubes)  or
light leaks that lead  to erratic response  in others (scintillation  detectors).
For this reason, most  detectors have a  screen  or grid protector over  the
window.   This  screen helps protect the  window, but  it also reduces  the
sensitive window area.
     B.   Portable Measurement Instruments.   Portable  survey  instruments for
measuring exposure or  exposure rate are generally  small,  portable  ionization
chambers.  Like portable detection  instruments,  portable  measurement  instru-
ments are selected based on the type, energy,  and  intensity of  the  radiation
to be measured, and the degree of accuracy needed.   The  technical  specifica-
tions of an instrument should be reviewed  to determine whether  it  is  appro-
priate for a particular use.  In addition,  the methods and  radioisotopes  used
to calibrate the instrument, the calibration curves,  and  the  necessary correc-
tion factors all affect the suitability of an  instrument.  Table  2.1  sum-
marizes  the kinds of portable survey instruments available for  both detection
and measurement of radiation.

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                                                              DARCOM-P 385-1
                    rABLE 2.1.  Portable Survey Instruments
Instrument
Air proportional
counter with probe
Gas-flow proportional
Range of
Counting Rate
0 to 100,000 dpm
over 100 cm2
0 to 100,000 dpm
Radiations
Detected
Q, photons
a, photons
Typical
Surfaces,
clothing
Surfaces,
Uses
hands,
hands,
counter with probe .     over 100 cm2
Portable scintillation
counter with probe
Portable count rate
meter (thin-window
GK counter)
Portable count rate
meter (BF3 tube)
0 to 100,000 dpm
over 100 cm2
e,
0 to 1,000,000 cpm  £,  y
                    (o-sensitive
                    with  appropriate
                    detector probe)
0 to 500,000 cpm    Neutron.
clothing
Surfaces, hands,
clothing
Surfaces, hands
clothing
               Area, beams
2.5.2  Laboratory Counting Instruments
     Field assessments of radioactive contamination are generally qualitative
rather than quantitative, and even when portable measurement instruments are
used, they cannot measure levels of radioactivity as low as  the levels labora-
tory counters can measure,  To precisely quantify levels of  activity, labora-
tory conditions and laboratory counting Instruments are required,  Laboratory
counters may Include GM tube detectors 1n heavily shielded chambers with
sealer readouts, scintillation counters, proportional  counters, semiconductor
detectors, and multichannel  spectrometers with computer analysis capabilities.
The counter selected for a specific application depends on the type, energy,
and level of radiation to be measured, and on the accuracy and precision
required.  Certain laboratory counting Instruments (e.g., Nal crystals) can be
used to determine the particular radionuclides in a sample as well as to
measure the activity of each radionucllde.
     Table 2.2 lists some of the available laboratory counters and their
sensitivities, as documented in Report 57 of the National Council on Radiation
Protection and Measurements  (NCRP 1S78).  Most samples analyzed as part of
radiation protection programs contain very small amounts of  activity.  The
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DARCOM-P 385-1
                      TABLE  2.2.'  Laboratory Counters
                Instrument
          GM counter
          Gas-flow proportional
          counter

          Gamma scintillation
          counter
                 Well
                 Probe

          Liquid scintillation
          counter

          Alpha scintillation
          counter

          Semiconductor
                                                    (a)
Radiations    Sample Sensitivity
 Detected     	(uCi)
   6
   Y
                                       a
                                       Y
10
10

10
  -4
  -2
5 x210


ID'5


5 x 10


<1 dpm
5 x 10
                          -5
                          -4
                           r
                          °
          (a) NCRP 1978.


counting instruments used should  therefore  be highly sensitive,  and the effect
of natural  background radiation levels  on the detectors  should be kept as low
as possible.   Facilities  used for laboratory counting should be  located in
areas of low background.   Room or detector  shielding may be required to reduce
instrument background levels.

     Extra precautions should be  taken  to assure that laboratory counters
are not contaminated by the samples  being counted.   Because these instruments
are highly sensitive to radiation, very small amounts of contamination bias
their counting results.  Frequent verification of background counting levels

is necessary.  In counters that have reusable sample holders, or planchets,
the empty sample holders  should be counted  periodically to ensure that they
have not become contaminated.
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                                                              DARCOM-P 3&5-1
2.5.3  Air-Monitoring Equipment
     Instruments used to monitor gaseous  or paniculate radioactivity  in air
should be highly sensitive because  the  amount 01" activity to be detected or
measured is usually small.  The  type  of equipment  used depends upon  the type,
energy, and half-life of the radiation  to be detected, whether it  is in
qaseous or participate form, and whether  sampling  or monitoring is to  be done.
     Air sampling and air monitoring  are  both performed to  determine tne pres-
ence or amount of radioactivity  in  air.   An air  sampler either collects the
air (for sampling radioactive gases)  or pulls the  air through e filter (for
sampling radioactive particulates in  air).  In either case, the sample is
removed for later analysis.   An  air monitor, on  the other hand, analyzes the
air in question as it is collected.
     A.  Air Samplers.  Air sampling  is performed  in the following circum-
stances:  when the probability for  airborne contamination is low;  as part of a
long-term environmental  program; where  a  high level of background  radiation  or
excessive contamination  prohibits air monitoring;  when the  consequences of
airborne contamination are known not  to be of immediate concern to the per-
sonnel in the area; as a check on the monitoring program; where great  sensi-
tivity for radionuclide  identification  is required; and where  surrounding
conditions (e.g., potentially explosive atmospheres) do not allow  the  use of
monitoring equipment.  The advantage  of an air-sampling system  is  that the
sample can be taken to an area of low background radiation, where  it can be
evaluated or held for the decay  of  natural radioactivity,  if desired,  and
where various sample-processing  steps can be performed and  sophisticated
equipment can be used to analyze the  sample.
     A general-purpose air-sampling system consists of a  collector (filter  or
sorbent), collector holder, flow-measuring device, flow  rate controller, and
air mover.   Some gas-sampling systems use evacuated flasks, cold  condensate
traps, or specially treated traps (e.g.,  activated charcoal for sampling  radon
gas).   Most sampling systems have the advantage  of being  small  and portable.
     In some areas, small oattery-operated samplers  (lapel  samplers) can  be
carried or worn by individuals to provide an integrated  sample of the
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DARCO.M-P 385-1
contaminants in the individual's  breathing  zone  (the air directly surrounding
the face).   Fixed samplers  can also  be  selectively  located to provide long-
term integrated samples,  which are useful  in  establishing the average concen-
tration of contaminants near probable  points  of  release and throughout the
work area.
     Grab samples are usually high-volume  samples collected over a short time
(i.e., from 2 to 20 minutes) and  used  for  determining  the level of particulate
contamination in air.  A portable air  suction pump  containing a filter paper
holder is located at the point of interest, and  a large volume of air (2 to
100 m ) is drawn through the filter.   The  filter is then removed to a count-
ing room or laboratory for  rapid  analysis.  Low-volume air samplers are used  in
environmental programs because they  can be  operated continuously for weeks or
months at a time.  When analyzed, the  filters from  these samplers indicate the
total release from a specific site over a  given  period.
     B.  Air Monitors.  Air monitoring is  performed when the sampling results
are needed immediately; when a real-time monitor is required to indicate the
need for immediate evacuation of  a work area; to provide a continuous reading
for trend analysis; to monitor releases to the environment  (as  in stack monitor-
ing); and to measure immersion doses from  gaseous releases.
     An air-monitoring system is  basically the same as an air-sampling system
except that an appropriate  counter  (e.g.,  a proportional counter) or other
evaluation instrument is placed near the collecting medium  (filter paper or
sample chamber).  Air monitors are often equipped with strip-chart recorders,
air activity meters (which  indicate, for example, cpm  per liter of air), check
sources, and visual and audible alarms. The  advantage of an air-monitoring
system is its continuous  and immediate indication of activity  levels.
     Most air monitors cannot detect low levels  of  radioactivity; therefore,
these monitors are most useful where the potential  for large radioactive
releases is highest.  For example, an  alpha air monitor  is  relatively  ineffec-
tive for measuring airborne depleted uranium  (DU).  By the  time an alpha
monitor detected DU and sounded an alarm,  the airborne activity would  be
several times above acceptable limits.
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                                                             DARCOM-P  385-1
     Some airoorne activity, such  as  low-energy  bete  ^articles  from  tritium,
can be measured using a Kanne chamber,  an  ionization  chamber through which the
air flows.  The beta particles are drawn  into  the  chamber, where  the ionizo-
tion must occur if it is to be detected.   However,  these  chambers  are also
sensitive to higher-energy background radiation, and  some compensation  for
background is normally required.
     C.  Principles of Operation.   Air  sampling  and monitoring  involve  collect-
ing a sample of a'ir or a material  removed  from the  air  and determining  by
analysis what the contaminant is  (if  that  is not already  known)  and  the quan-
tity of it.  Accurate determination of  the activity in  a  sample  requires
accurate measurement of the volume of air  sampled.  For gaseous  samples, this
may be as simple as knowing the volume  of  the  chamber in  the sampler used.
However, a system for sampling particulates requires  accurate measurement of
1) the rate at which air flows through  a  filter  medium  and 2) the time  over
which the sample is taken.  The system must have an air mover capable of mov-
ing the air at the rate desired,  a method  of ensuring that the  air flow is
constant for the sampling period,  and calibration  of  the  air sampler.
     Many variables must be considered  in  establishing  a  quantitative air sam-
pler.  The type of filter paper or sorbent medium  should  be selected to effec-
tively remove from the air the contaminants of interest.  The collection
efficiency of the medium should be established,  taking  into account  the size
of the particles collected and the air velocity  during  collection.  Isokinetic
sampling of ducts and effluent stacks should be  used.  This means that  the
opening of the sampling device should be  set perpendicular  to the direction  of
air flow, and the sample flow rate should  be adjusted so  that the linear eir
speed into the sampler is the same as that of  the  approaching air si-earn.
Anisokinetic conditions may cause  an  over- or  underestimation  of particulate
air concentrations in the air stream.  In  addition, the representativeness  of
the sample at the collecting point may be  affected by materials becoming
deposited on the sampling lines or passages, a condition  called plateout.
Attention must be given to limiting the length of  a sample  line, the degree of
curvature of bends in the line, and the temperature gradients  between  the  air
being sampled and the line.
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 DARCQM-F 385-i
     Instrumentation used to measure the  activity of the collected sample is
selected based on several factors:   whether  the  instrument  is to be used as
part of a continuous monitor or whether it  is  in a counting  room or labora-
tory; the type and energy of the radiation  being detected;  and the sensitivity
required.  Geiger-Mueller counters,  gas proportional counters, scintillation
counters, or semiconductor counters  can be  used  to measure  the activity of air
samples.
2.5.4  Other Fixed Instruments
     In addition to the radiation detection  and  measurement  instruments pre-
viously discussed, special-purpose instruments can be  used.   These instruments
include remote area monitors and continuous  air  monitors.
     A.  Remote Area Monitors.  Remote area  monitors (RAMs)  are usually GM
detectors or ionization chambers used to moni'tor direct  exposures to  photon
radiation.   These monitors are usually permanently positioned and have visual
or audible alarms or both.  They are often  connected to  other RAMs in a
network, with the results displayed  in a central control  room.  These monitors
usually have a variable alarm setting so that  the alert  level can be  altered.
     In addition to the alarm function, RAMs may incorporate a continuous
recorder so that a historical record of radiation levels is provided  and
radiological conditions and trends can be followed and evaluated.
     B.  Continuous Air Monitors.  Continuous  air monitors  (CAMs) are similar
to remote area monitors in function, but they  always monitor the  radioactivity
concentrations in sir continuously.   This type of air  monitor can be  fixed  in
place, with sample lines to the instrument  from  the area being monitored,  or it
can be semiportable (usually a relatively heavy  cart on  wheels)  and  can  be
moved to the area to be monitored.  Depending  upon the type of  radiation  to be
measured and whether it is in gaseous or particulate form,  CAMS  may  use  GM,
gas proportional, semiconductor, or ionization chamber detectors.   The  com-
plete CAM unit includes an air mover, air flow controls, the appropriate
electronics for the detector being used, an alarm, and usually  a  recorder.
Tnose fixed in pUce may also be wired for a meter  readout, a strip  chart
recording,  and an alarm at some remote or central  location.
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                                                              DARCOM-P 385-1
     Tne factors th=t effect other air-sampling  and  air-monitoring  systems
also affect CAMs.  in addition, CAM units can be affected  by  changes  in
ambient radiation levels, the fluctuations of unregulated  power,  and  contami-
nation from outside the area being sampled.
     To avoid a long-term buildup of radioactivity and  dust on  filter media,
fixed CAMs require frequent filter changes.   Other CAM  units  use  a  moving
filter tape.  An advantage of the moving-filter  CAM  is  the capability of pro-
viding a delayeo counting sequence to allow for  the  decay  of  natural  back-
ground radioactivity.  Instruments of this type  can  be  provided with  duplicate
detectors, one instantaneous and one delayed, and electronic  circuitry to
allow background compensation and alarm functions for both instantaneous
releases and long-term buildups of radioactivity.
                      Section 2.6. * PERSONNEL DOSIMETERS

     A radiation dosimeter, loosely defined, is any instrument or system
capable of measuring radiation dose.  Dosimeters are typically used to provide
a quantitative estimation of the radiation dose actually received by personnel.
Their response should be reproducible, precise, and accurate,  and the instru-
ments should be able to measure all  ionizing radiations encountered by per-
sonnel.  They should be simple and convenient to use, small,  easy to handle,
and low in cost.  Because personnel  dosimeters record only the dose they have
received, it is extremely important that personnel  be trained in their proper
use.  One type of dosimeter, the pocket ionization  chamber or pencil dosim-
eter, was already discussed in Section 2.2.1.  Three other types—photographic
film, nuclear track emulsions, and thermoluminescence dosimeters—are dis-
cussed below.
2.6.1  Photographic Film
     Photographic film is measurably darkened by radiation, and can therefore
provioe c useful estimation of personnel exposure.   The response of photo-
graphic film depends on the type, energy, and amount of the radiation reaching
the film.
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DARCQM-f 385-1
     A.   Principles of Operation.   The sensitivity of film is defined as the
amount of darkening produced by a  specified  radiation exposure.   Photographic
films, or emulsions, consist of a  layer of tiny  silver halide crystals
embedded in a gelatin matrix.   The emulsion  is spread across a thin  sheet of
plastic  or glass plates.   The  thickness of the emulsion can range  from  10 to
2000 urn, depending on the sensitivity desired.   The  thicker the  emulsion, the
greater the sensitivity of the film.
     When ionizing radiation travels  through photographic emulsions, the
radiation imparts a small amount of energy to the silver halide  crystals,
causing  some of the silver ions to be reduced to free atomic silver.  These
silver atoms form traps capable of capturing electrons, which can  in turn
reduce more silver ions and create a  microscopic aggregate of silver atoms.
These silver aggregates are frequently referred  to as latent image centers.
Chemical treatment of the film causes the latent image centers to be reduced
to metallic silver, which appears  to  the eye as  a blackening of  the film.  The
degree or density of darkening can then be related to radiation  exposure.
     B.    Dosimeter Design.  Photographic films  are  incorporated into the so-
called film badge.  The modern film badge is designed so that  radiation can
reach the film either directly through an open window, or through filters.
The filters are disks made of metals, such as lead,  tin, copper, cadmium,
silver,  or aluminum, and are used to  distinguish between different types  and
energies of radiation.  For example,  thin filters of a  low-atomic-number
(low-Z)  material, such as aluminum, can be used  to distinguish  between  gamma
rays and high-energy beta particles.   Other  metallic filters  can help  identify
the contribution of different components of  the  gamma-ray  spectrum. Most film
wrappers stop beta particles with an  energy  less than  about  150  keV.  Thus,
film cannot be used to monitor radiation exposures from  low-energy beta emit-
             •?      14
ters such as "H and   C.
     C.   Effects of Environment.  Photographic  film  degrades  with age.   Under
normal conditions, dosimeter films usually last  for  several  months before they
begin to deteriorate.  However, the latent image centers  and the overall
response of the film can be adversely affected  by  environmental  conditions.
The latent image fades if the film is subjected  to  high  temperatures,  high
humidity, or oxygen.  Of all these influences,  relative  humidity is the

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                                                              DARCOM-P 385-1
dominant factor.  Film packets shoulc not be used or handled  by  unqualified
personnel.   Films should be kept in their lightproof packages to reduce the
possibility  of  light leaking in, which could ruin the film.
     D.  Processing Techniques.  The process used for developing film dosim-
eters  is basically the same as that used for de\eloping  medical  x-ray films.
Specifically, a film is placed first in a developer solution  and then in a
fixer, which stops the development process by dissolving the  unused silver
                 t
halide crystals.  How long the :"ilm is left in the developer  solution, the
amount of agitation of the solution, and the temperature and  age of the
solution all affect the first step o:' the process.   How  long  the film is left
in the fixer affects the quality and permanence of the image  on  the film.
When the film is removed from the fixer (after approximately  10  minutes), it
is washed and then dried at room temperature.
     E.  Interpretation and Calibration.  Once the film  has been processed, it
is read and  interpreted.   To reduce the probability of error  in  the reading of
the film, unexposed control films are processed along with the exposed films.
Unexposed films produce a density or darkening during processing known as the
base fog.  By processing control (unexposed) dosimeters  along with the exposed
dosimeters,  it is possible to subtract the degree of darkening of the base fog
from the degree of darkening on the exposed dosimeters.
     The processed film is analyzed using a densitometer, a device that mea-
sures the degree of film darkening.  Interpretation of the densitometer read-
ing is then related to exposure, depending on the density value  under each of
the filters in the badge.   Doses should be interpreted only by personnel who
are highly skilled in evaluating photographic film.  Even with properly
designed filters and film badge holders, the accuracy of photographic film is
limited because its response is dependent on the radiation energy and the
inherent variability in films.   In mixed radiation fields (fields that include
both high- and low-energy radiation), low energies can result in errors of
=50« to ±200«.   However,  with properly designed film badges and  properly
controlled usege,  photographic  films can achieve an accuracy  of  ±25* in most
personnel  dosimetry situations.
                                     2.45

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UARCCLV ,
      Photograph", fHm dosimeters are not. absolute devices and therefore must
be calibrated against a known source in order to relate the film density to
the exposure delivered.  The calibration of dosimeters should be performed
under carefully controlled laboratory situations using sources traceable to
NBS.
2.6.2 Nuclear Track Emulsions
      Standard photographic film badges are not designed to respond to neu-
trons.  However, nuclear track emulsion (NTA) film, which is thicker than
standard photographic film, can be used to monitor for neutrons.  The neutrons
reaching the NTA film interact in a variety of ways with the atoms in the
emulsion, charged particles are produced, and the charged particles in turn
interact with the silver atoms of the NTA film to form tracks that are visible
after the.film is developed.  The tracks can be counted and related to neutron
dose.
      Nuclear track emulsions are even more sensitive to latent image fading
than  are the standard films.  Therefore, the wearing interval for NTA film
dosimeters normally does not exceed 2 months, and 2 weeks is the preferred
wearing time in a high-humidity climate.  Fading can be reduced and the wear-
ing time increased if the NTA film is sealed into a moisture-proof package in
a nitrogen atmosphere.
2.6.3 Thermoluminescence Dosimeters
      Some crystals emit light when they are heated after exposure to ionizing
radiation; this process, .known as thermoluminescence, is similar to the
scintillation process described earlier and is the basis for another type of
personnel dosimeter.
      A.  Principles of Operation.  The crystals most commonly used in thermo-
luminescence dosimeters today include lithium fluoride (LiF), calcium fluoride
(CaF^), calcium sulfate (CaSO,), and lithium borate (LipB.O-,).  When one of
these crystals is exposed to ionizing radiation, many of the free electrons
within the crystal become excited and are caught in imperfections of the
crystal, or traps.  The exposed crystal can be stored at room temperature for
long  periods without a significant number of the electrons escaping from the
                                     2.46

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                                                             DARCCLM-P 385-1
traps.  However,  when the crystal  is  heated  to  higher temperatures, the  trapped
electrons escape  and lose their excess  energy by  the emission of visible  light
(thermoluminescence).  Because  the amount  of light  released from a heated
crystal is proportional  to the  energy or radiation  dose absorbed within  the
crystal, the phenomenon of thermoluminescence can be used  in radiation dosim-
etry.  A dosimeter that uses this  f.henomenon is called a thermoluminescence
dosimeter (TLD).
     A TLD reader', which has a  controlled  heating element,  is used to determine
how much light is emitted during the  heating of a dosimeter crystal.  The  light
intensity is plotted as a function of temperature,  and the  resulting graph is
called a glow curve.  The glow  curve  normally has several  peaks at various
temperatures.  The area under any  peak can be used  es a measure of the dose
received by the TLD.
     When a TLD has been irradiated and read on a TLD reader, it  can be
annealed and reused.  Annealing is a  slow heating process  that  completely
empties the traps and restores  the crystal to  its original  state.  After the
crystal has been  allowed to cool,  it  is ready to  be reused.
     B.  Advantages and Limitations.   The TLD has a wide  dose-response  range
(I mrad to 10  rad) and a very  low energy dependence.  The most popular  TLD
material, LiF, has an effective atomic number very  close  to that  of  human
muscle tissue.  Thus, it is considered by most  users to  respond much  es  tissue
would and is frequently considered "tissue equivalent."
     Other advantages of TLDs are  that they are very small, quite rugged, and
essentially unaffected by environmental variables.   Because TLDs  show very
limited fading (unlike film dosimeters), the wearing interval  for the  TLD can
be a year or longer.  The advantage of the longer wearing period is  a  reduc-
tion in the error produced by numerous processings  throughout the year.   The
reported accuracy of most TLDs  under controlled laboratory conditions  is =1%.
An accuracy of ±10% is fairly easily achieved in  the field.
     Thermoluminescence dosimeters are essentially  unaffected by their orienta-
tion in the radiation field and by the rate of  exposure.   However, tne badge
                                     2.47

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DARCOM-P 385-1
or device that is designed to hold the thermoluminescent material  may adversely
affect the accuracy of the dosimeter.   Therefore,  proper badge  design is  essen-
tial in the correct use of TLDs.
     A major limitation of TLDs is thet,  after they  have been  processed,  their
exposure information is erased; film,  on  the other hand, retains  the information
as a permanent record.
     C.  Interpretation and Calibration.   Interpreting  the  results of a glow
curve produced from an irradiated TLD  requires establis-ing a  relationship
between the glow curve and a known exposure level.  The best procedure is to
obtain a large batch of dosimeters with well-matched responses  and to run a
calibration curve over the exposure range of interest,  using a  known radiation
field.
     The use of properly calibrated dosimeters is  critical  to  a good health
physics program.   An installation that has a small radiation protection staff
should procure the services provided by the Army or  a commercial  calibration
company.  Calibration companies should maintain their traceability to NBS
through a periodic direct intercomparison.
     D.  Practical Applications.   Thermo!uminescence dosimeters can be used in
any situation where film -dosimeters are currently being used.   They are pre-
ferred to film for extremity dosimeters (e.g., ring  and wrist  badges), for
personnel monitoring where radiation energies are below about  100 keV, and for
environmental monitoring.  However, TLDs  do not provide a  permanent record of
exposure, as film dosimeters do.
     Unlike film dosimeters, TLDs can  also be used to measure  the neutron
radiation to which an individual  is exposed.  Thermo!uminescent materials are
more sensitive to thermal (slow)  neutrons than to fast neutrons.   Thermal neu-
trons interact with a TLD as they pass through it to the  wearer.   Some thermal
neutrons may be reflected back to the  TLD from the irradiated  individual and
may interact with the dosimeter then also.  Fast neutrons,  on  the other hand,
do not interact with the TLD as they pass through it.  These fast neutrons
interact with the hydrogen in the wearer's tissues,  where they lose their
energy (become thermal).  Many are then reflected back toward  and interact
                                     2.48

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                                                              DARCOJ-5-P 385-1
with the dosimeter.   The reflected thermal  neutrons are called  albedo  neutrons.
Correct interpretation of albedo dosimetry  requires that the  radiation source,
the dosimeter, and the irradiated individual  be  in line and that  the original
energy of the reutrons be known.  The  neutron energy or a  description  of  the
radiation source should be given to the  dosimetry service  interpreting the
response.
                Section 2.7  STATISTICS  AND  ERROR  DETERMINATION

     The spontaneous emission of radiation by  nuclear  processes  occurs  randomly
in time, and all  measurement and detection instruments must  respond to  these
statistically random events.  This  means that  the  interpretation of instrument
response must take into ac:ount the random nature  of radioactive decay.   We
tend to assume thet a measurement is an  absolute  indication  of the activity of
the source.   However, this is usually not the  case.  It  is more  likely  that
only a fraction of the radiation car, be  detected.  This  error must be cor-
rected, us ing statistics and geometry correction factors.
2.7.1  Systematic and Random Errors of Measurement
     The errors associated with radiation measurements can be divided into two
types:  systematic and random.  Systematic errors  are  created in the measure-
ment process or in the interpretation of measurement data.   They are frequently
caused by faults  within the electronic systems of instruments.  For example,
low batteries or faulty electronic  components  could  bias measurements,  and the
results would be  considered to contain a systematic  error.   The primary source
of random errors  is radioactive decay.
2.7.2  Basic Statistical  Distributions for Radioactive Decay
     If a long-lived radionuclide of low activity was  counted many times, and
if a plot was made showing the number of times a  given count rate occurred
versus the count  rate, the plot would be similar  to  the one  shown in Fig-
ure 2.5. . This curve is called a normal, or  Gaussian,  distribution and repre-
sents the distribution of count rate values  obtained in successive counts.
                                     2.49

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 DARCOM-P 385-1
                  or
                  <->
                  •<
                  UJ (_>
                  S o
                  00
                    o
                    O
                                     COUNT RATE
               FIGURE 2.5.   Frequency of Occurrence of Count Rates
                            for a Long-Lived Sample

     The normal distribution curve can be  described mathematically by calculat-
ing the mean and the standard deviation of all the count rates used  to prepare
the curve.  The mean, or the arithmetic average of the count rates,  describes
where on \.he curve the greatest number of  counts occurs.   It is calculated by
summing all of the count rates and dividing by the number  of counts  taken.
Written in mathematical terms, the equation appears as follows:
                                                                       (2.3)
ni
where           n = the mean of the count rates
                N = the number of times the sample was counted
               n. = the value of the ith count rate
            N
            y n. = the sum of all the count rates.
The individual measurements taken in any radiation survey are distributed
about this sample mean.
     The standard deviation (o), e measure of variability, describes  the width
of the curve and is a useful indication of how extensively the  count  rates
                                     2.50

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                                                              DARCOM-P 365-1
vary from the average value.  The square of the standard Deviation is celled
the variance and can be approximated using the expression
                      1
                            N
                           L  (n - n.)'
                                          (2.O
where          c  - the variance
                o = the standard deviation
                N = the number of times the sample was counted
                 t
                n = the mean of the count rates
               n. = the value of the ith count rate
      N-         2
      £(n - n.)  = the sum of all  the squared deviations from the mean.

Wnen only a few measurements have been taken (fewer than 20), a best estimate
of the standard deviation can be derived as follows:  o = i/n".  When more than
20 measurements have been taken, the previous method for calculating the vari-
ance and the standard Deviation should be used.
     Figure 2.6 shows a plot of the normal distribution curve with several
features indicated.  In a normal distribution, 68.3* of all counts are within
            o
            O
                      -3o
-20
-o
t2c
+30
          FIGURE 2.6.   Normal Distribution Function Showing Standard
                       Deviations and Mean
                                     2.51

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DARCOM-P 385-1
±1 standard deviation of the mean value,  95.5* within  ±2  standara  deviations
of the mean, and 99.72 within r3 standard deviations of the mean.   For  example,
if a sample is counted 100 times, the mean value obtained is  1000  cpm,  and  the
standard deviation is 100, then we can say, with a  68.3%  chance of being  cor-
rect, that the mean count rate is between 900 and 1,100.   Thus, the specifica-
tion of activity is a "probabilistic event"; that is,  we  specify with a certain
statistical accuracy that the mean activity lies within a range of values.
     For statistical purposes, when the results of  a series of measurements
are recorded, both the mean and the standard deviation should be specified.
                             Section 2.8  RECORDS

     Records are needed to verify the availability and use of appropriate
radiation detection and measure instruments, the adequacy of their calibration
and maintenance, the proper interpretation and use of the resulting data, and
compliance with regulatory requirements.  A complete discussion of instrument
recordkeeping procedures is presented in Chapter 13.
                                  REFERENCES
American National Standards Institute (ANSI).  1949.  Standard Test Procedure
  for Geiger-Mueller Counters.   ANSI N42.3, New York.
American National Standards Institute (ANSI).  1978.  Criteria for Testing
  Personnel Do si.Tie try Performance.  ANSI N13.11-1978, New York.
American National Standards Institute (ANSI).  1978.  Radiation Protection
  Instrumentation Test and Calibration.   ANSI N323-1978, New York.
National Council on Radiation Protection and Measurements (NCRP).  1978.
  A Handbook of Radioactivity Measurements Procedures.  NCRP 57,
  Washington, D.C.
                                     2.52

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                                                              DARCO.M-P 3 £5-1
                   CHAPTER 3.   RADIATION1 PROTECTION  PROGRAM

3.1  REGULATION?	3.3
     3.1.1  Department of the  Army	3.3
     3.1.2  Federal	•   ....     3.4
     3.1.3  International     .........     3.5
3.2  RADIATION PROTECTION STANDARDS     	     3.5
     3.2.1  Radiation Exposure Standards     	     3.5
            A.  Occupational Exposure   .......     3.6
            B.  Occupational Exposure to Women    	     3.6
            C.  Occasional Exposure     	     3.7
            D.  Exposure of Minors      .     .    .     .     .     .     .     3.7
            E.  Emergency Exposure .     .     .    .     .     .     .     .     3.7
            F.  Nonoccupational Exposure     	     3.8
            G.  Alternate Exposure Standards      	     3.8
     3.2.2  Administration Limits and Action Levels     ....     3.8
     3.2.3  The ALARA Philosophy	3.9
3.3  ELEMENTS OF A RADIATION PROTECTION PROGRAM   	     3.9
     3.3.1  Licenses, Authorizations, and Permits .....     3.10
     3.3.2  ALARA Program	3.10
            A.  Management Commitment   	     3.10
            B.  Assignment of ALARA Responsibility and Authority .    .     3.11
     3.3.3  Surveillance and Monitoring Programs  .....     3.13
     3.3.£  Radio!ooiccl Desic-     	     3.13
     3.3.5  Radioactive-Materiel Control and Waste Management    .    .    3.14
                                      3.1

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DARCOM-P 385-1



     3.3.6  Emergency Planning     .          ......     3.14



     3.3.7  Personnel Selection,  Qualification, and Training     .     .     3.15



     3.3.8  Recordkeeping     .........     3.15



     3.3.9  Quality Assurance Program   .......     3.15



3.4  ADMINISTRATION OF THE RADIATION  PROTECTION PROGRAM     .     .     .     3.16



     3.4.1  Ionizing Radiation Control  Committee  .....     3.17



     3.4.2  Radiation Protection  Officer     	     3.18



REFERENCES	3.19



APPENDIX A - REVIEW OF PROPOSALS  FOR  RADIATION USE     ....     3.21








                                '    TABLES





3.1  Regulations Applicable to Army Activities    	     3.4



3.2  Radiation Protection Standards     	     3.6
                                      3.2

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                                                              DARCQM-P 385-1

                   CHAPTER 3.  RADIATION PROTECTION PROGRAM
     The objectives of a radiation protection program are to reduce exposures
to a level as low as is reasonably achievable within the occupational  dose-
equivalent limits set by the federal  government and the Department of  the Army
(DA) and to minimize the potential for accidental  exposures.  The components
of an effective radiation protection  program are common to all installations
where radioactive materials are used  or stored.  However, the magnitude and
complexity of the program may vary from one installation to another.  This
chapter describes briefly the principles and practices that should be  con-
sidered in the establishment of a radiation protection program.  These prac-
tices are covered in greater detail in later chapters of this manual.
                           Section 3.1  REGULATIONS

     A variety of government branches and international agencies have formu-
lated regulations governing the procurement, use, storage, transportation, and
disposal of radioactive materials and sources.  The National Council on Radia-
tion Protection and Measurements (NCRP) and the International Commission on
Radiological Protection (ICRP), whose members are professionals in health
physics or related fields of research, provide recommendations that serve as
the basis for most Army and other government agency requirements.  Knowledge
of and compliance with all applicable regulations are essential factors in the
administration of every radiation protection program.  Agencies thet may have
jurisdiction over specific radiological situations are discussed briefly below,
end the applicability of their regulations is summarized in Table 3.1.
3.1.1  Department of the Army
     All Army installations that produce, procure, receive, store, use, ship,
or dispose of radioactive materials or sources ere required to have a radia-
tion protection program.  Specific requirements for ionizing radiation protec-
tion programs can be found in AR 385-11, AR 40-14, DARCOK-R 385-25, and
AR 700-64.
                                      3.3

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DARCQM-P 385-1

             TABLE 3.1.  Regulations Applicable to Army Activities
                 Activity                        Applicable Reaulations
     Day-to-day operations                  10 CFR 20,  10 CFR 19
                                            AR 40-14, AR 700-64, AR 385-11
                                            DARCOM-R 385-25
     Use of radiation-producing machines    21 CFR 1000-1050
     (such as x-ray machines)
     Transport of'radioactive materials     49 CFR
     Shipment through mails                 39 CFR
     International shipments                Inter-Governmental  Maritime
                                            Consultative Organization and
                                            International Air Transport
                                            Association

3.1.2  Federal
     The U.S. Nuclear Regulatory Commission (NRC) regulates the production,
distribution, use, and disposal of source, byproduct, and special nuclear
materials.  The use of radioactive materials and radiation sources within the
work environment not governed by the NRC is regulated by the U.S. Occupational
Safety and Health Administration (OSHA).  The requirements of NRC are described
in Title 10 of the U.S. Code of Federal Regulations, Parts 19 and 20 (10 CFR
19 and 20).  Army regulations require that civilian and military personnel
within the United States and overseas be provided radiation protection that is
at least equal to that required by 10 CFR 19 and 20.
     The U.S. Department of Health and Human Services (HHS) conducts a radia-
tion control program for electronic products.  The program includes the devel-
opment of performance standards to protect the public health from ionizing and
nonionizing radiation in electronic products.  This department also regulates
and sets standards for the use of radioactive materials and radiation sources
in foods, drugs, cosmetics, and medical devices, as set forth in 21 CFR.
     The U.S. Department of Transportation (DOT) regulates the packaging end
transportation of radioactive materials shipped in interstate commerce by air,
rail, highway, and water.  The U.S. Postal Service regulates shipment via the
                                      3.4

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                                                              DARCQM-I 385-1
U.S. mail.  The regulations of their agencies are  presented  in  Titles 49 and 39
of the Code of Federal Regulations,,  respectively.
     The U.S. Environmental Protection Aoency (EPA)  provides  federal  guidance
on radiation protection.   The EPA also develops  standards  governing the release
of radioactive materials  and radiation sources to  the  environment (40 CFR).
3.1.2  International
     An agency of, the United Nations, the International  Atomic  Energy Agency
(IAEA), provides overall  safety guidance for the international  shipment of
radioactive materials.  The Inter-Governmental Maritime  Consultative Organiza-
tion (JMCO) and the International Air Transport  Association  (IATA) provide
regulations for the international shipment of radioactive  materials.  The
specific application and  enforcement of the regulations  is the  responsibility
of each nation through which material is transported.   Normally, a shipment
that complies with the regulations of the nation of origin complies by agree-
ment with the regulations of the nations through which the shipment is routed.
                  Section 3.2  RADIATION PROTECTION STANDARDS

     Dose-equivalent limits for controlling occupational and nonoccupational
exposure to ionizing radiation and radioactive materials have been established
by DA (AR 40-14).  These limits are based on the recommendations of NCRP and
ICRP.  Both organizations emphasize that dose-equivalent limits are upper
limits for planned exposures and that every effort must be made to keep expo-
sures below these limits and to avoid unnecessary radiation exposure.  This
principle is strongly emphasized in federal regulations as the As Low As is
Reasonably Achievable (ALARA) philosophy.
3.2.1  Radiation Exposure Standards
     Standards established by the Army fall into several categories:  occupa-
tional exposures, occupational exposures to women, occasional exposures, expo-
sure of minors, emergency exposures, nonoccupational exposures, and alternate
exposure standards.   These cateqories are described below.
                                      3.5

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DARCO.M-P 385-1
     A.  Occupational Exposure.  Occupational  radiation exposure  standards  are
presented in Table 3.2.  Occupational  exposure to  ionizing  radiation  is  that
resulting from military or civilian duties  that directly  support  the  use of
radioactive materials or equipment capable  of  producing ionizing  radiation.
Occupational exposure does not include exposure to naturally  occurring  ionizing
radiation or exposure received as a result  of  medical  or  dentcl diagnosis or
treatment.   An occupational^ exposed  individual,  or radiation worker,  is one
whose work  is performed in a radiation area or a controlled area  (see
Chapter 8)  and who might be exposed to more than 5% of the  limits given  in
Table 3.2.
     B.  Occupational Exposure to Women.   Special  radiation exposure  controls
are established for the protection of  unborn children.  The NCRP  recommends
that during the entire gestation period,  the maximum'dose equivalent  to  the
                 TABLE 3.2.  Radiation Protection Standards
                                                           (e)
                  Orqan
                                               Occupational  Dose-Equivalent
                                              	Limit,  rem	
 a. Whole body, head and trunk, active
    blood-forming organs, gonads, lens
    of the eye
 b. Skin of the whole body (other than
    hands, wrists, feet, or ankles)
    forearms, cornea of the eye, bone
 c. Hands and wrists, or feet and
    ankles
 d. Forearms
 e. Thyroid, other organs, tissues, and
    organ system
 Individuals under the age 18, and
 occasionally exposed individuals
 Individuals between ages 18 and 19
 (^hole-body limit)
Calendar Quarter
      1.25


      7.50


      18.75

      10
       5
Calendar Year
      30


      75

      30
      15
      10% of the values listed
      above
       1.25
 (e) AR 40-14.
                                      3.6

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                                                             DARCOM-P 365-1
embryo-fetus from the occupational  exposure of the expectant mother  should  not
exceed 0.5 rem (NCRP Report Nos.  39 (1971) and 53 (1977)).  A woman  staff mem-
ber is responsible for advising her employer  that she is  pregnant.   Special
consideration, such as a  change in  work  assignment, may be necessary to  ensure
that her occupational exposure does not  exceed recommended limits  and is kept
as low as is reasonably achievable.  The installation commander  and  the  Radia-
tion Protection Officer (RPQ) should determine appropriate actions and policies.
     C.  Occasional Exposure.  An occasionally exposed  individual  is one whose
duties do not normally involve exposure  to  ionizing radiation or radioactive
material, but who may have a reason to enter  a restricted area on  a  nonroutine
basis.  Examples ere repair personnel and messengers.   When such individuals
enter a restricted area,  they shall not  be  exposed to a whole-body dose  equiv-
alent of more than 1) 2 mrem in any 1 hour, 2) lOO.mrem in any 7 consecutive
days, 3) 500 mrem in any 1 calendar year, or  4)  5% of the values for other
areas of the body detailed in Table 3.2.
     D.  Exposure of Minors.  A minor is any  person under 18 years of age.   For
a minor, the accumulated dose equivalent of radiation shall not  exceed 10%  of
any of the values listed in Table 3.2.  Persons  over the  age of  18,  but  who
have not reached their 19th birthday, may be  occupationally exposed  to ionizing
radiation if they do not receive a  dose  equivalent of more  than  1.25 rem to the
whole body in any calendar quarter.
     E.  Emergency Exposure.  Radiation exposure standards  in  emergency situa-
tions vary according to the severity of the emergency.  When  entry into a
hazardous area is necessary to search for and remove seriously  injured persons,
or to prevent conditions that may injure a  number of people,  the accumulated
whole-body dose of each individual  entering the  area should  not  exceed 100 rad,
and the accumulated dose to the hands and forearms  should not  exceed 300 rad.
In a less severe situation, when it is desirable to  enter a  hazardous area to
protect property, minimize the release of effluent,  or  control  fires, the
accumulated whole-body dose of each individual  entering the  area should not
exceed 25 rad, and the dose to the  hands and  forearms  should  not exceed
100 rad.  Individuals who incur such radiation expos-ures  during  an emergency
                                      3.7

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DARCQM-P 385-1
should not be allowed to do so more than once  in  a  lifetime.   The  record  of
such exposure becomes part of the person's  health record  or civilian  employee
medical file.
     F.  Nonoccupational Exposure.   Sources of ionizing  radiation  must  be used
in such a way that 1) the accumulated dose  equivalent  to  the  whole body for  an
individual  person in the general  public does not  exceed  0.5 rem in any  1  calen-
dar year, and 2) the average accumulated dose  equivalent  for  a suitable sample
of the exposed population or for the whole  exposed  population does not  exceed
0.170 rem/year from all  sources of radiation (excluding  medical  and nature!
background radiation).
     G.  Alternate Exposure Standard?.  Radiation exposures standards that are
less restrictive than those described above may be  used  in special circum-
stances, but only when approved by the Surgeon General  of the United States  or
the director of the Defense Logistics Agency,  as  appropriate.  Proposals  for
alternate radiation exposure standards must contain a  complete justification
and must specify the procedures by which the standards will be implemented.
Less restrictive standards will not be considered for  1)  persons under 19 years
of age, 2)  women known to be pregnant, 3) occasionally exposed persons, and
4) nonoccupational exposure of the general  public.
3.2.2  Administrative Limits and Action Levels
     Administrative limits and action levels are frequently set to help main-
tain occupational exposures within established limits.  Administrative limits
are radiation exposure limits established by the administrator of a. radiation
protection program, for example,  80S or less of the occupational exposure stan-
dard.  An administrative limit is basically a  control  point:   as an individ-
ual's exposure approaches this level, the individual is carefully monitored so
that the exposure does not exceed the limit unless specific management approval
1s obtained.  Thus, individual exposures are kept lower, and  the possibility of
exceeding permissible exposure limits is reduced.
     Action levels are dose-equivalent limits  that, when reached or exceeded by
an Individual, require formal investigation into the cause of exposure.  The
RPO should establish investigative procedures.  An investigation should  lead to
                                      3.8

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                                                              DARCQM-P 385-1
the identification of portions of the radiation protection program that need to
be improved.  Action levels, also called investigative levels,  are established
by radiation protection management.
3.2.3  The ALARA Philosophy
     Even though current occupational exposure limits keep the  risk of injury
to personnel very low, it is prudent to avoid unneccessary exposure to radic-
tion.  The operating philosophy of every radiation protection program should be
to reduce occupational exposures as far below specified limits  as is reasonably
achievable,  Tnis philosophy, emphasized in federal  regulations and referred to
as ALARA (As Low As is Reasonably Achievable), means that each  work procedure
that will result in a radiation dose should be subject to scrutiny and that
methods to reduce the dose should be identified.   The methods that involve the
least cost and result in the greatest reduction of dose should  be considered
and implemented wherever possible.  References in the bibliography discuss
ALARA and ALARA programs in greater detail than is possible here.
     It is not desirable to maintain the dose equivalent of a radiation worker
at a small fraction of the applicable limit if this practice requires that a
larger number of people be exposed.  Therefore, in addition to  maintaining
occupational exposure to individuals as far below limits as is  reasonably
achievable, the goal of ALARA is to keep the sum of the doses received by all
exposed individuals (radiation workers, other personnel, and the general public)
at the lowest practicable level.  The sum of the dose equivalents received by
all exposed individuals is called the collective dose equivalent.
            Section 3.3  ELEMENTS OF A RADIATION PROTECTION PROGRAM

     An effective radiation protection program includes licensing, an ALARA
program, surveillance and monitoring programs, proper design of facilities in
which radiation sources are used, control of radioactive materials and waste
disposal, emergency planning, adequate mining of personnel, the maintaining
of reliable and complete records, and a quality assurance program.
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DARCQM-P 385-1
3.3.1  Licenses, Authorizations,  and Permits
     Whenever radioactive materials  or radiation  sources are produced,  procured,
used, stored, transported, or disposed of at  DA facilities, an NRC  license  and/
or DA approval is required.   Procedures for obtaining  the  necessary documents
are contained in AR 385-11.   Non-Army agencies, including  civilian  contractors,
are required to obtain a DA radiation permit  to possess, use, store,  or dispose
of radiation sources on an Army installation.
3.3.2  ALARA Program
     The establishment and management of all  radiation protection programs
within Army facilities should be  guided by the ALARA philosophy.   Each  radia-
tion protection program should therefore include  a  formal  ALARA  program. An
effective ALARA program requires  management commitment and the assignment of
ALARA responsibility to an individual or committee, as discussed below.  Pro-
cedures for maintaining exposures ALARA are described throughout this manual.
Particular attention should be directed to Chapters 5 and  6, which described
the control of internal and external exposure.
     A.  Management Commitment.  Management commitment to  the  safe and correct
use of radiation and radioactive  materials is probably the single most impor-
tant characteristic of a good radiation protection  program.  Upper management,
specifically the base commander,  sets the tone for  the safety  program.   The
commander must indicate by word and action that safety is  important.  Simply
displaying safety slogans and posters, holding safety contests,  and establish-
ing safety committees have little effect unless individual staff members
believe that safety is important  to their supervisors.
     The commitment made by management to minimize  exposures  should result  in
clearly defined responsibilities  for radiation protection  and  an environment  in
which the radiation protection staff can do its job properly.   This commitment
should be made evident in the following areas:
     (1)  Personnel Awareness of  Management Commitment.  The  ALARA principle
should appear in policy statements, instructions  to personnel, and similar
documents.  Staff members should  be familiar enough with this  commitment to
explain what management policy is, what is meant  by keeping exposure to
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                                                              DARCOM-P 385-1
radution "as low es Is reasonably achievable,"  why it  is  recommended,  and how
they have been advised to implement it on their  jobs.   They must  understand the
importance of the philosophy.
     (2)  Radiation ^rotection Personnel.  Management  should ensure t-.at there
is a well-supervised radiation protection staff  with well-defined responsi-
bilities.  The RPO should be quc'";fied to handle any potential  problems at the
installation.
     (3)  Training.  Management should ensure that personnel receive sufficient
training.  Section 19.12 of 10 CFR 19 requires that personnel be  instructed in
radiation protection.  They should understand how radiation protection relates
to their jobs and should be tested on this understanding at least once each
year.  Radiation workers should have opportunities to  discuss radiation safety
with the radiation protection staff whenever the need  arises.  The training
program in radiation protection should be reviewed by  management  at least once
every two years.
                                                   *
     (4)  FacilTty Modifications.   Modifications in operating and maintenance
procedures and in plant equipment  and facilities should be made if they will
substantially reduce exposures at  a reasonable cost.  Management  should encour-
age the staff to suggest improvements and modifications and should implement
them where practicable.
     (5)  Audit Programs.  A formal audit should be conducted periodically to
determine how exposures might be reduced.  The audit should include reviews of
operating procedures and exposure  records, inspections, and consultations with
the radiation protection staff.
     B.   Assignment of ALARA Responsibility and Authority.  The base commander
should formally assign ALARA responsibility to an  individual such as the  RPO or
to a group of individuals such as  the Ionizing Radiation Control  Committee
(IRCC).   The RPO should have sufficient authority  to prevent unsafe practices
and to communicate promptly with an appropriate level  of management about
halting  unsafe operations.   This authority should  be specified in written
policy statements.   The members of the IRCC are chosen for  their knowledge  of
radiation safety principles, engineering, and design,  knowledge that is useful
in evaluating the safety of projects involving radioactive  materials.
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     -M-P 385-1
     Operating procedures  related to  radiation  safety  snou1d be  reviewed  and
e^oroved by radiar.on protection  personnel.   The  RPO end/or the  IRCC  should be
responsible for conduction surveillance programs  and investigations  to  ensure
that occupational  exposures are es  far below the  specified limits  as  is
reasonably achievable.   All of these  individuals  should  constantly be seeking
new and better ways to perform all  radiation jobs with less exposure.   There
ere several aspects of this responsibility.
     (1)  Monitoring of Exposures.  The RPO  and the  radiation  protection  staff
should know the origins of radiation  exposures  by location, operation,  and  job
category and should be aware of trends in exposures.   They should  be  able to
describe which locations,  operations, and jobs  are associated  with the  highest
exposures and why exposures are increasing or decreasing.  Where standing
operating procedures are used, exposures received should be recorded on the
written procedures.
     (2)  Investigation of Unusual  Exposures.  When  unusual exposures have
occurred, the radiation protection staff should direct and participate  in an
investigation of the circumstances to determine the  causes and take steps to
reduce the likelihood of similar  future occurrences.   For each such occurrence,
the RPO should be able to demonstrate that an investigation was carried out,
that conclusions we're reached as  a result of the investigation, and that appro-
priate corrective actions were taken.
     (3)  Review of Operating Procedures.  The RPO and the  radiation protection
staff should periodically review  operating procedures  that may affect radiation
safety.  They should survey plant operations to identify situations in which
exposures can be reduced,  and should  implement any changes  that are needed.
The RPO should repeatedly emphasize that work performance that results in per-
sonnel meeting dose-equivelent limits is not acceptable  when  it is practical
to reduce the dose to a lower level.   Procedures should  be established for
receiving and evaluating staff members' suggestions relating  to radiation pro-
tection and dose reduction, and the staff shquld be aware of these procedures.
     (4)  Provision of Equipment  anc1  Supplies.  The RPO should be responsible
for ensuring that equipment and supplies appropriate for radiation protection
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                                                              DAJRCOM-P 385-1
v:ork tre available, ere maintained in good working order,  and are used prop-
erly.  Written procedures for the use of the equipment should be available and
followed.
3.3.3  Surveillance and Monitoring Programs
     Another component of a radiation protection program is  surveillance end
monitoring, which help keep radiation exposures to personnel  and the public
ALARA and within applicable dose-equivalent limits.   Routine  survey programs,
used to assess the radiological status of a facility, are  discussed in
Chapter 4 of this manual.  Procedures for monitoring personnel  are described
in Chapters 5 and 6.
3.3.4  Radiological Design
     The terms facility design, radiological design-, and radiological engi-
neering are often used interchangeably, although their meanings are different.
Facility design refers to a plan for a building or installation as a whole, and
thus includes nonradiological  as well as radiological design  features.  Radio-
logical design refers to the specific set of features required in a facility
because of the planned presence of radioactive source or radiation-generating
machines.  Radiological engineering refers to the actual construction of a
facility in which radioactive  materials will be stored or used.  (The term can
also be used in a broader context to include design.)  Design implies the devel-
opment of an idea as  opposed to the actual construction and  operation of a
facility.
     Proper facility  design is an effective approach to reducing occupational
exposures.   Well-designed facilities provide a greater degree of safety than
can be obtained by dependence  on administrative rules and procedures alone.
Although design can never eliminate the possibility of accidental radiation
exposure or contamination, it  can reduce the probability and magnitude of such
accidents.   A qualified expert should therefore participate  in the planning and
design of new facilities and of modifications to existing facilities.  Topics
that should be considered in radiological design are discussed in Chapter 8.
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DARCQM-P 385-1
3.3,5  Radioactive-Mater-al  Control  and Waste  Management
     Proper control  of radioactive materials  is  necessary to ensure  that  per-
sonnel and the general puolic are protected from unnecessary exposure  to  radia-
tion.  Such control  extends  to all aspects  of  radioactive-materials  handling,
including procurement, use,  storage, shipment, and waste disposal.
     The RPO should  review ail procurement  and transfer requests  for radiation-
producing sources and devices and should monitor and  inventory  radioactive
materials when they  are received to ensure  that  they  have not been  damaged in
transit or caused contamination of personnel  and facilities.  Radiation sources
may then be transferred to authorized users in the organization or  stored in
specially designated facilities until needed.  Later  transfer of radioactive
materials may require special procedures to assure proper controls,  and care
should be taken to ensure that the person or organization receiving the mate-
rials is licensed and authorized to received and use  them.
     An inventory should be  maintained to ensure that the RPO can at any time
determine the identity, quantity, and location of all  radioactive materials.
The location, safe condition, and use of radioactive  materials  should be con-
firmed by a periodic audit-and by routine surveys performed by  the  RPO.
     The RPO should  review the disposal of  all radioactive  materials.   They
should be disposed of by transfer in suitably prepared containers to authorized
locations for radioactive waste disposal.  Transportation  is  discussed more
fully in Chapter 9,  radioactive-waste disposal in Chapter  10, and inventory
record systems in Chapter 13.
3.3.6  Emergency Planning
     Every facility  in which radioactive material,  radiation-generating
devices, or radiation sources are produced, used, or  stored should  have an
emergency plan.  The emergency plan may be  simple or  complex, depending upon
the facility,  in all cases, however, it should  be  documented,  reviewed peri-
odically, and tested at least yearly.
     An emergency plan is created through evaluation  of the accident potential
of a facility.  The  emergency actions necessary  to  reduce  the consequences  of
potential accidents, and the individuals responsible  for those  actions, are
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                                                             DARCOM-P  385-1
then determined.   Coordination with outside  emergency  fc-ces  (public  informa-
tion officials, hospitals, and police,  fire,  and  health  Departments)  is  also
planned.  When an emergency plan has been established, realistic exercises  in
which key staff memoers participate should be held  to  test  the  adequacy of
emergency preparedness.  These exercises should  include  tests of evacuation
procedures, the use of emergency equipment,  and  those  rescue  and first  aid
techniques in which staff members may play a role.   Periodic  testing  of
emergency equipment and instrumentation is also  necessary.  Procedures  for
developing a plan are described in Chapter 11.
3.3.7  Personnel  Selection, Qualification, and Training
     Adequate training is fundamental to a radiation protection program.  Appro-
priate training should be extended to the radiation protection  staff, installa-
tion management,  and radiation workers.  Training programs  are  discussed in
Chapter 12.
3.3.8  Recordkeeping
     Documentation is needed as evidence to support the  reliability end effec-
tiveness of a radiation protection program.   Records should be  complete and
should reveal the patterns of radiation exposure at the  facility.   Data on all
operating and working conditions should also oe  available.  The records that
should be considered for retention are described in Chapter 13.
3.3.9  Quality Assurance Program
     A quality assurance program is a means of verifying that each part of a
radiation protection program is being carried out adequately  and that the total
program meets its purpose.  A quality assurance  program should  be developed for
any facilities or locations where the following  take place:
 1.  radioactive  material is received,  used, stored, or prepared for disposal
 2.  radiation-generating machines are operated
 3.  personnel dosimeters are evaluated
 4.  radiation detection or measuring equipment  is procured,  received,
     repaired, calibrated, or used
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DARCOM-P 385-1

 5.  facilities or equipment that will  be  used  for these activities are
     designed, constructed, or modified.

Quality assurance is discussed in Chapter  14.
        Section 3.4  ADMINISTRATION OF THE  RADIATION  PROTECTION  PROGRAM

     The success of a radiation protection  program is  dependent  on  firm  manage-
ment commitment to the program and on the availability of  individuals  who  have
a thorough understanding of radiation protection  principles.   Within  the DA,
the overall responsibility for the radiation protection program  rests  with the
commander, director, or chief of the installation or  activity.   The management
and administration of the radiation protection program is  delegated to desig-
nated personnel such as the RPO or the IRCC.  The IRCC is  an  advisory body that
assists the commander in establishing local  rules and procedures for  the pro-
curement, storage, and safe use of radiation sources.   The- committee  consists
of the commander, the RPO, the safety officer, and the medical  officer (if
ave11able)--or representatives of these Individuals—together with  a  repre-
sentative of employee groups, and others knowledgeable in  radiation protection.
The RPO is generally responsible for the Implementation of the radiation pro-
tection program,  This Individual must be  technically qualified through educa-
tion, training, and professional experience.
     The assignment of responsibility must  be accompanied  by  accountability and
authority,  Authority granted to the radiation protection  staff should be broad
and fully supported by upper management,  Specific authorities should include
the following:
 1.  approve plans for the construction or  modification of facilities in which
     radioactive materials will be used or  stored, or in which radiation-
     generating machines will be located
 2.  issue and approve standing operating  procedures  or job safety analyses
     (this implies review and approval of operating plans  and procedures before
     their implementation)
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                                                              DARCOM-P 385-1
 3.   determine  operational  protective measure? to ensure that exposures ere
     kept ALARA
 4.   train and  assess  the  qualification  of  radiation workers
 5.   plan for and establish equipment and procedures for monitoring and control
     of personnel  exposures.
     Authorities  should  be delineated between the IRCC and the  RPO.   One way
this can be done  is  outlined  below.
2.4.1  Ionizing Radiation  Control  Committee
     The duties of the IRCC can  include  the following:
 1.   Review proposals  for  the use  of  ionizing radiation sources and recommend
     protective measures to the  commander (AR 40-14).
 2.   Prescribe  any special  conditions and requirements that may be  needed  (such
     as physical  examinations, additional training, designation of  limited
     areas or locations  of use,  disposal methods, etc.),
 3.   Prepare and  disseminate  information on radiation safety  for use  by and
     guidance of  personnel.
 4,   Pass judgment on  the  adequacy of safety measures and  health protection  for
     safeguarding personnel.
 5.   Keep a record of  actions taken in approving  the use of radioisotopes,  and
     of other transactions, communications, and reports  involved in the work of
     the committee.
 6.   Provide policy  direction to the  RPO, based upon state and federal regula-
     tions and  licenses, for  the use  of  ionizing  radiation at the
     installation.
 7.   Approve or disapprove all applications from  prospective  users  of radioac-
     tive materials  and  from  prospective operators  of sources of ionizing
     radiation.
 8.   Approve or disapprove all applications for laboratories  in which radioac-
     tive materials  would  be  used  or  in  which sources of  ionizing radiation
     would be operated.
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DARCQM-P 3S5-1
 9.  Review plans for all  new buildings  or for  modifications  to existing  build-
     ings in which radioactive materials or other  sources  of  ionizing  radiation
     would be used.
10.  Suspend any operation that,  in  the  opinion of the  IRCC,  represents a
     serious radiation hazard or  violates applicable  regulations.
3.4.2  Radiation Protection Officer
     The RPO's responsibilities can  include the following:
 1.  Ensure compliance with current  directives  for radiation  protection.
 2.  Provide consultation  on the  hazards associated with  radiation and the
     effectiveness of measures to control these hazards.
 3.  Supervise the radiation protection  program and advise on the  control of
     hazards to health and safety.
 4.  Coordinate, the day-to-day administration and  development of  the radiation
     protection program.
 5.  Disseminate information on radiation safety and  health physics.
 6.  Review all proposals  for radiation  usage and  recommend to the IRCC
     approval or disapproval of all  applications from prospective users of
     radioactive materials and from  prospective operators of sources of
     ionizing radiation.  Detailed information on  such reviews is given in
     Appendix A.
 7.  Inspect facilities and equipment on behalf of the IRCC.
 8.  Review plans for all  new radioisotope and radiation facilities.
 9.  Obtain all necessary  licenses and registrations  pertaining to radioactive
     materials and sources of ionizing radiation for  the installation or
     activity.
10.  Develop procedures for the purchase and transfer of radioactive materials.
11.  Develop procedures for the disposal of solid and liquid radioactive
     wastes.
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                                                              DARCO.M-P 385-1

12.  Maintain required records,  including  the  following:   personnel  dosimetry,

     radioactive waste disposal,  radioisotope  inventory,  instrument  calibra-

     tion, and leak tests on sealed  sources.

13.  Provide radiation surveys and monitoring  of  all  radioisotope  and  radiation

     facilities.

14.  Offer brief courses on radiation safety for  users  and prospective users  of

     radioactive materials and ionizing  radiation.

15.  Suspend any operation that,  in  the  opinion of  the  RPO,  represents a  ser-

     ious radiation hazard or violates applicable regulations.   The  operation

     suspended will be reviewed  by  the IRCC.
                                  REFERENCES


National Council on Radiation Protection and Measurements (NCRP).  1971.   Basic
  Radiation Protection Criteria.   NCRP 39,  Washington, D.C.

National Council on Radiation Protection and Measurements (NCRP).  1977.
  Review of NCRP Radiation Dose Limit for Embryo and Fetus in Occupetionslly
  Exposed Women.  NCRP 55, Washington, D.C.

U.S. Code of Federal  Regulations.   1982.  Title 10, Part 19,  "Notices,
  Instructions and Reports to Workers; Inspections."  U.S. Government Printing
  Office, Washington, D.C.

U.S. Code of Federal  Regulations.   1982.  Title 10, Part 20, "Standards for
  Protection Against Radiation."   U.S. Government Printing Office,
  Washington, D.C.

U.S. Code of Federal  Regulations.   1982.  Title 10, Part 40, "Domestic Licens-
  ing of Source Material."  U.S.  Government Printing Office, Washington, D.C.

U.S. Code of Federal  Regulations.   1982.  Title 21, "Food and Drugs."  U.S.
  Government Printing Office, Washington, D.C.

U.S. Code of Federal  Regulations.   1982.  Title 39, "Postal  Service."  U.S.
  Government Printing Office, Washington, D.C.

U.S. Code of Feceral  Regulations.   1922.  Title 40, "Protection  cf Environ-
  ment."  U.S. Government Printing Office,  Washington, D.C.

U.S. Code of Federal  Regulations.   1982.  Title 49, "Transportation."  U.S.
  Government Printing Office, Washington, D.C.
                                     3.19

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DARCaK>-P 385-1

U.S. Department of Defense, Defense Supply Agency.  Radioactive Commodities in
  the POD Supply Systems.  DSAM 4145.8, AR 700-64, Washington, D.C.

U.S. Department of the Army, Headquarters.  Safety - Ionizing Radiation Pro-
  tection (Licensing, Control, Transportation, Disposal, and Radiation Safety).
  AR 385-11, Washington, D.C.

U.S. Department of the Army, Headquarters, Army Materiel Cormiand.  Safety -
  Radiation Protection.  DARCON-R 385-25, Washington, D.C.

U.S. Department of the Army and Defense Logistics Agency.  Medical Services -
  Control and Recording Procedures for Exposure to Ionizing Radiation and
  Radioactive Materials.  AR 40-14, DLAR  1000.28, Washinoton, D.C.
                                     3.20

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                                                              DARCOM-P 385-1
                                  APPENDIX A
                     REVIEW OF PROPOSALS FOR RADIATION  USE
     The RPO is responsible for reviewing project  plans,  personnel,  and facil-
ities before work with radioactive material  or radiation-producing devices  is
begun.  Standing operating procedures are then prepared by the RPO and the
IRCC, and records are maintained as the work proceeds.

A.I  PRC'ECT EVALUATION
     The project should be evaluated based on the  following criteria:
     A.  License.  The RPO should check the site or'facility license to ensure
that the radioactive material proposed for use can be brought onsite and that
the proposed chemical and physical form and the proposed uses of the material
are allowed by the license.  If the license does not show the proposed uses, an
amendment to the license must be requested.   For assistance in preparing an
amendment to the license, or in interpreting the license to determine whether
an amendment is necessary, contact DARCOM headquarters.
     Six months or more may elapse before a requested license amendment is
authorized.  Project leaders should be made aware  of the possibility of delay;
they can then inform the RPO of the needs of the project early enough in the
planning process so that the license amendment will be approved at about the
same time as the project is scheduled to begin.
     B.  Radionuclide.  The RPO should assess the  radionuclide to be used,
considering whether an alternate, less hazardous radionuclide could be used
              33              32
(for example,   P rather than   P) and whether a radionuclide is necessary
at all or whether other methods of achieving the same purpose are available.
     C.  Quantity.   The amount of radioactive material used for the project
should be the minimum possible.  If possible, the  stock quantity of radioactive
material shculo be divided into small eliquots.
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DARCQM-P 385-1
     The quantity of radionuc"! ide proposed for use  should  be  compared  with  the
amount for which the site or facility is licensed.   This comparison  should  take
into account both the quantity to be used during  the project  and  the total
onsite inventory for that radionuclide.   The inventory  of  concern includes  both
quantities in laboratories and waste quantities that are stored  and  waiting for
shipment.
     D.  Chemical and Physical Form.  If the material  proposed  for use is
volatile,  the need for a volatile form should be  assessed.   Chemical methods
for reducing the volatility of the chemical compound may be available; for
example, raising the pH of an iodine solution reduces the  amount of  iodine
released into the atmosphere.  Concentrated solutions of alpha-emitting
                       244
radionucndes, such as    Cm, may present difficulties.  Dilute  solutions are
less likely to cause volatilization.
     E.  Work Procedures.  The RPO should consider  whether there are standard
procedures for doing the proposed work;  whether the proposed work follows the
established procedures; whether the procedures can  be improved,  for example, by
reducing the work time; and what types of protective apparel  should be worn.

A.2  PERSONNEL CONSIDERATIONS
     Personnel considerations in the assessment of  a project include:
     A.  Pregnant or Potentially Pregnant Women.  The DA recommends in AR 40-14
that, during the entire gestation period, the maximum dose equivalent to an
embryo-fetus from occupational exposure of the expectant mother should not
exceed 0.5 rem.  Because pregnancy may not be confirmed for two or more months
after conception, women staff members should be made aware of this  recommenda-
tion and should be encouraged to tell the RPC when  they are contemplating
pregnancy or as soon as pregnancy is suspected.
     B.  Minors.  Individuals under 18 years of age shall  not be exposed to
more than 10% of the occupational dose limits.
     C.  Education and Training.  Personnel assigned to work on  projects
involving the use of radioactive materials or radiation-generating  sources
should be educated as to the hazards associated with radiation and  trained  in
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                                                              DARCOM-P 385-1
the specific skills required for their job.  Their attendance at education end
training sessions should be documented by attendance rolls,  and  the RPO should
administer tests that cover the material following the training  sessions.  The
tests should show whether the material was understood and indicate areas  of
training that require increased emphasis.
     D.  Personnel Monitoring.  The .RPO should ensure that personnel who  will
work with radioactive materials are provided with appropriate monitoring
devices.  Monitoring devices such as film badges shall be worn by all personnel
who receive, or may be expected to receive, a radiation dose higher than  5« of
the applicable standard tc the whole body or skin.  In practice, whole body
badges are usually issued ^o all individuals who work with x- or gamma-ray
sources or with beta emitters that have a maximum energy of 1.0 MeV.  Film
badges should also be worn by individuals who work around particle accelerators
and neutron sources.
     Extremity monitors should be worn by individuals who may receive an
extremity dose higher than 5% of the applicable standard.

A.3  EVALUATION OF FACILITIES
     The facility or work area in which the project will be carried out  should
be evaluated to ensure that radioactive materials can be used safely.  The U.S.
Environmental Hygiene Agency and DARCOM headquarters should be contacted for
assistance.  The information to be considered includes:
     A.  Shielding.  The amount of shielding required depends on the radio-
nuclide to be used (or the operating energy of the radiation-producing machine),
the quantity of radioactive materiel to be present (or the operating time of
the machine), and the proposed use of adjacent areas.  If shielding already
exists, the RPO should assess whether it win -be sufficient, how much addi-
tional  shielding will  be required, and whether the building can support  the
required shielding.
     B.  Equipment.  The working area should have appropriate equipment, which
may include hoods, glove boxes, and air filter systems.
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DARCOM-P 385-1
A.4  STANDING OPERATING PROCEDURES
     After the project has been analyzed, a standing  operating procedure (SOP)
is prepared by the IRCC and the RPO.  The SOP is  a  summary of the safety
findings and a listing of the procedures that must  be followed during the
course of the project.  The SOP should include the  following items:
 1.  type of protective apparel required, if any
 2.  posting requirements
 3.  radiation-monitoring devices required
 4.  personnel dosimeters required
 5.  bioassay types and frequency
 6.  recordkeeping requirements
 7.  reiteration of applicable administrative guidelines
 8,  any special procedures that may be required.  '

A.5  RECORDKEEPING
     The purpose of recordkeeping is to help the  RPO  1) document the radiation
doses received by personnel and 2) assess trends  in the rate at which doses are
being received over time.  Recordkeeping also allows  the RPO to compare the
doses received by staff members who are working on  similar projects and in this
way to learn which techniques result in the lowest  doses to workers.  It can
also make possible intercomparlsons of doses received during similar projects
at different facilities.
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                                                              DARCOM-P 385-1
                    CHAPTER 4.   RADIATION  SURVEY  PROGRAMS

4.1  SURVEY REQUIREMENTS 	     4.6
     4.1.1   Facilities  That Require  a  Survey  Program    ....     4.6
     4.1.2   Areas Within a  Facility  That Require  a  Survey  Program      .     4,6
            A.   Radiation Areas	4.6
            B.   Controlled  Areas   	     4.7
            C.   Uncontrolled Areas  	     4.7
     4.1.3   Frequency of Surveys	4.7
4.2  ROUTINE SURVEY PROCEDURES     ....'....     4.8
     4.2.1   Preparation	4.8
            A.   Gathering Information    	     4.8
            B,   Diagramming the  Facility	4.10
            C,   Preparing an Inspection List	4.10
            D,   Obtaining Equipment  and Material   .    .     .     ,     .     4.12
     4.2.2   Inspection  and  Measurement  .......     4.12
            A.   Inspection	4.12
            B.   Measurements  .	4.13
     4.2.3   Evaluation  and  Recommendations	4.13
     4.2.4   Survey Records  and Reports  	     4.13
4.3  SPECIFIC MONITORING PROCEDURES      	     4.14
     4.3.1   Measurements of Radiation  Fields  	     4.15
     4.3.2   Measurements of Contamination     	     4.15
            A.   Direct  Measurements      .     .      .         ,     .     .     4.15
            B.   Indirect Measurements    	     4.16
                                      4.1

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DARCQM-P 385-1



            C.  Action Levels ana Reporting   .     .     .     .     .     .     4.17



     4.3.3  Leak Testing Sealed Sources	          4.17



            A.  Direct Leak Testing	4.18



            B.  Indirect Leak Testing (Container  Interior)   .     .     .     4.18



            C.  Indirect Leak Testing (Container  Exterior)   .     .     .     4.19



     4.3.4  Personnel  Monitoring   ......               4.19
                  t


            A.  Contamination      .     .     .     .     .     .     .     .     4.20



            B.  Internal Exposure  	     4.20



            C.  External Exposure  .     .     .     .     .     .     .     .     4.21



     4.3.5  Air Monitoring	     .     .     .     4.21



     4.3.6  Tritium Monitoring     ........     4.21



            A.  Water	4.22



            B.  Urine	4.22



            C.  Surfaces	4.23



            D.  Air	     .     .     4.23



4.4  NONMEDICAL X-RAY INSTALLATIONS     	     4.23



     4.4.2  Classification of Nonmedical X-Ray Installations     .     .     4.24



            A.  Protective Installation 	     4.24



            B.  Enclosed Installation   	     4.24



            C.  Unattended Installation 	     4.24



            D.  Open Installation	4.25



     4.4.2  Engineered Safeguards  ........     4.25



            A.  Protective Installation      	     4.25



            B.  Enclosed Installation   	     4.26



            C.  Unattended Installation      	     4.26



            D.  Open Installation	4.27
                                      4.2

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                                                              DARCOM-? 385-1

     4,4.3  Administrative Controls     .......    4.2?

            A.  Training	4.27

            B.  Standing Operating Procedures     .     .     .     .      .    4.27

            C.  Operation and Maintenance Logs    .....    4.28

            D.  Radiation Area Requirements  	    4.28

     4.4.4  Survey of Nonmedical X-Ray Installations   ....    4.28

            A.  Frequency     .     .     .    .    .     .     ,     .      .    4.28

            B.  Procedure	4.29

            C.  Radiation Survey Report      	    4.29

4.5  ENVIRONMENTAL SURVEY PROGRAMS      .    .    .     .     .     .      .    4.30

REFERENCES	4.31

APPENDIX A - MAXIMUM PERMISSIBLE CONTAMINATION LEVELS  .     .     .      .    4.33
             FOR INANIMATE OBJECTS
                                      4.3

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                                                                   *: P 3E5-i
                    :HAPTER 4.  RADIATION SURVEY  PROGRAMS
     Routine survey programs are used to evaluate  actual  or potential  radia-
tion hazards at facilities where radiation sources are  usec.   Surveying and
monitoring are ways of maintaining radiation exposure  to  personnel  end the
environment at a level that is as low as is reasonably  achievable (ALARA)
within applicable dose-equivalent limits.
     The terms "radiation survey" end "radiation monitoring,"  although fre-
quently used interchangeably, are not synonymous.   A radiation survey is an
evaluation, under specific conditions, of the radiation hazard associated  with
the production, use, or storage of radioactive materials  or other sources  of
radiation.  Radiation surveys are conducted both in the working environment
and in the environment surrounding a facility.  Radiation monitoring, an activ-
ity frequently performed during a survey, is the measurement of radiation
fields or radioactive contamination using fixed or portable instruments.
Radioactive contamination can be defined as any radioactive material  that  has
escaped from its intended location or container, or as  the deposition of radio-
active material in any place where it is not desired,  and particularly in  any
place where its presence might be harmful.  Radioactive contamination can  be
any combination of alpha-, beta-, gamma-, or neutron-emitting radionuclides.
Radiation surveys and radiation monitoring are usually performed by the Radia-
tion Protection Officer (RPO) or a member of the radiation protection staff.
     Survey requirements and procedures for facilities where radiation sources
or radioactive materials are produced, used, or stored are discussed -in this
chapter.   Specific radiation monitoring procedures are also described, as are
special requirements for facilities that house nonmedicel x-ray units.  The
objectives and development of environmental survey programs are discussed
briefly at the end of the chapter.
                                      4.5

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DARCOM-P 383-
                       Section 4.1  SURVEY REQUIREMENTS

     Radiation surveys are recommended or required  for  certain  types  of  facil-
ities and for specific areas within those facilities.   The  frequency  of  surveys
varies depending on the facility, area, and other factors.
4.1.1  Facilities That Require a Survey Program
     A routine survey program should be considered  for  any  facility where the
radiation level may be higher than the natural  background  level.   A survey
program is required for facilities that contain the following  specific sources:
 1.  radioactive solids that exceed 1 uCi in activity,  that have  a specific
     radioactivity exceeding 0.002 uCi/g, or that emit  radiation  at a dose
     rate of 0.1 mrad/hr or more at contact
 2.  materials controlled by the Nuclear Regulatory Commission  (NRC), in quan-
     tities that exceed those listed in Title 10 of the U.S.  Code of  Federal
     Regulations. Section 30.71, Schedule B (10 CFR 30.71)
 3.  machines that produce radiation, for example,  x-ray devices, accelerators,
     and electron microscopes
 4.  radioactive gases or liquids in concentrations that exceed the values
     listed in 10 CFR 20, Appendix B, Table II
 5.  items activated in nuclear reactors, by accelerators,  or by nuclear
     weapons.
4.1.2  Arees Within a Facility That Require a Survey Program
     Facilities are generally divided into a series of sequential areas
according to the radiation hazard in each area.  The designations of these
areas helps control personnel exposure to radiation.  The areas used are:
1) radiation areas, 2) controlled areas, and 3) uncontrolled areas.  Each of
these areas should be surveyed by a member of the radiation protection staff.
The areas are described briefly below and more  fully in Chapter 8, "Selection
of Radiation Facilities."
     A.   Radiation Areas.  Radiation areas include three subclassifications:
radiation areas, high-radiation areas, and airborne-radioactivity areas.
                                      4.6

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                                                               DARCOM-P 383-1
     n radiation area is defined in 10 CFR 20 as  any  area  accessible to per-
sonnel in which radiation levels could result in  a  mejor portion of the body
receiving a dose in excess of 5 mrem in any 1 hour  or 100  mrem in any 5 con-
secutive deys.  Practically, this would be any area in which the dose-
equivalent rate is greater than 2 mrem/hr but less  than 100 mrem/hr.  A nigh-
radietion area is any area accessible to personnel  in which radiation levels
could result in a major portion of the body receiving a dose in excess of
100 mrem in any 1 hour.  An airborne-radioactivity  area is any area, enclo-
sure, or operating area in which airborne radioactivity exceeds the concentra-
tions specified in 10 CFR 20, Appendix B, Table 1,  Column  1 or in which the
concentration of airborne radionuclides, averaged over the number of hours an
individual works, will exceed 25% of the amounts  specified in 10 CFR 20,
Appendix B, Table 1, Column 1.
     B.  Controlled Areas.  Controlled areas are  areas controlled for the
purpose of protecting personnel from exposure to  radiation.  Normally, they
are areas adjacent to radiation areas.  They are  usually free of contamina-
tion, but they could become contaminated because  of accidental spreads or
releases from the radiation area or because radionuclides and contaminated
equipment may be transported through them.
     C.  Uncontrolled Areas.  Uncontrolled areas  are areas where direct radi-
ation exposure is not necessary or anticipated in the performance of a job.
These areas include "cold" laboratories (those containing no activity), offices,
lunchrooms, conference rooms, and reception areas.   Access to these areas does
not need to be restricted for radiological reasons.
4.1.3  Frequency of Surveys
     Radiation areas, high-radiation areas, and airborne-radioactivity areas
should be surveyed at least once each month.  Permanent storage areas may be
exempted from monthly surveys at the discretion of the  Ionizing Radiation
Control Committee (IRCC).  However, the time* between surveys of storage areas
may not exceed 12 months.  Controlled areas should be surveyed on a  routine
basis.
     The frequency of surveys should increase if changes  in conditions or pro-
cedures could increase the possibility of personnel exposure.  Daily  surveys
                                      4.7

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DARCOM-P 385-1
 or continuous monitoring may be required if conditions  ere highly  variable  or
 unpredictable, if unsealed radioactive materials  are  be.ng handled directly,
 or if a radiation accident has occurred.
      Surveys should be conducted before an operation  involving  radiation
 sources is begun and before changes in an existing  operation are approved.   A
 survey is also required at the termination of a  project involving  the  use  of
 radiation, to verify that no contamination exists and that  radiation  sources
 and radioactive materials have been properly stored or disposed of.
      All sealed sources in quantities larger than the quantities  listed in
 10 CFR 30.71, Schedule B, must be leak tested at least every 6  months, unless
 specifically exempted by a DA authorization or an NRC license.   Alpha  sources
 in quantities larger than those listed must be tested every  3 months,  unless
 otherwise exempted.
                     Section 4.2  ROUTINE SURVEY PROCEDURES

      An effective routine survey program includes the following steps:
  1) preparation, 2) inspection and measurement, 3) evaluation and recommenda-
  tions, and 4) completion of records and reports.  These steps are described in
  detail below.  Special considerations for the survey of facilities containing
  nonmedical x-ray devices are considered in Section 4.4.4.
  4.2.1  Preparation
      It is essential that adequate preparation be made before any routine sur-
  vey is conducted.  The member of the radiation protection staff who is conduct-
  ing the survey must be thoroughly familiar with the sources of radiation and
  the nature and purpose of the work performed in the facility.  The steps for
  complete preparation are:  1) gathering information, 2) diagramming the instal-
  lation, 3) preparing an inspection list, and 4) obtaining necessary equipment
  and material.
      A.  Gathering Information.  Preparation for a survey should begin with the
  gathering of information about the radiation sources present, their intended
  use, and the physical safeguards and written procedural controls used to
                                       4.8

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                                                              DARCO.M-P 385-1
minimize personnel  exposure to radiation.   This  information  can  be  obtained  by
talking to personnel  and by examining plans,  drawings,  recorrs,  and written
procedures.  A file containing f  1  information  pertinent  to  a  particular
facility or work area should be maintained.   Examples of  the types  of  informa-
tion to be obtained and filed are:
 1.  the types and  numbers of sources used (e.g.,  sealed  sources, unsealed
     sources, or radiation-generating devices)
 2.  the types and  energies of radiation produced  by the  sources, together  with
     any information  about absorbers or moderators usec to alter the  initial
     energy spectre
 3.  the geometry,  size, and position of radiation fields
 4.  the direction  of beams produced by radiation-generating devices
 S.  the chemical composition and physical form of radioactive materials
 6.  the expected type(s) of radiation and/or contamination (e.g.,  alpha, beta,
     gamma, neutron)
 7.  the potential  for release or dispersion  of radioactive material
 8.  the procedures and the nature of the facilities used for  the storage,
     handling, transportation, and disposal  of radiation  sources and  radioac-
     tive material
 9.  the design and construction of devices for containing unsealed radioactive
     materials and  sources (e.g., hoods or glove boxes)
10.  the design of  ventilation and exhaust systems
11.  the design of  interlock, alarm, and emergency shutdown systems
12.  the nature of  fixed monitoring equipment used in the facility
13.  the locations  inside and outside the facility that are occupied by per-
     sonnel, and whether persons potentially exposed there are classified as
     occupationally or nonoccupationally exposed
14.  protective barriers used for exposure control
15.  standing operating procedures  (SOPs)
16.  previous survey  records

                                      4.9

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DARCOM-P 365-1
  17.  emergency plans
  18.  the training and experience of personnel  working with  the  radiation
       sources.
       E.  Diagramming the Facility.   The second step  in  preparing  for  a  survey
  is to make a diagram of the facility showing the  location of  raaiation  creas,
  controlled areas, and uncontrolled  areas.   The relative position  of  sources,
  work areas, waste storage areas, and disposal  areas  within  radiation  areas
  should also be shown.  Such a diagram can  be useful  in  identifying locations
  where radiation measurements should be made.  The location  of the following
  items should be included on the diagram when appropriate:
   1.  radiation sources, radiation-generating devices, and radioactive materials
   2.  the direction of beams produced by radiation-generating  devices
   3.  radiation areas, controlled areas, and uncontrolled areas
   4.  protective barriers (e.g., ropes, shielding)
   5.  interlocks, alarms, emergency  shutdown systems, and warning  signs
   6.  equipment, such as hoods and glove boxes, used  to  contain  unsealed
       radioactive sources and materials
   7,  waste storage and disposal areas
   8.  ventilation and exhaust systems
   9.  monitoring equipment.
       C.  Preparing an Inspection List.  After reviewing all the information
  related to the facility, the radiation protection staff member conducting the
  survey should list all the items to be inspected during the survey.   The
  inspection should include a review of the  adequacy of procedural  controls and
  physical safeguards used to control personnel  exposure, and verification  that
  all radiation protection procedures are being complied  with.   A review of the
  lists above can be useful in preparing the inspection list.  Examples of  items
  that could be included are:
   1.  the presence, location, use, and physical integrity of each radiation
       source
                                       4.10

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                                                              DARCOM-P 365-1
 2.  the means of identifying  -.sen  radiation source (e.g., serial number,
     type, activity,  size,  roor, location)
 3.  the presence and adequacy of  required  protective barriers (e.g.,  ropes,
     shielding)
 4.  the possibility  of inadvertent movement or  removal of shields
 S.  the possibility  of change in  the  orientation of beams produced  by
     radiation-generating devices,  or  of any change in the position  of
     sources
 6.  the availability,  condition,  and  use of safety and special-handling
     equipment (e.g., portable shields, remote-control devices,  hoods, pro-
     tective clothing,  showers)
 7.  the possibility  of the introduction of radioactive materials  into the
     facility's effluent stream because of  improper air flow  or  water
     drainage
 8.  the adequacy of  facilities and procedures for retaining  and/or  disposing
     of radioactive waste
 9.  the facility's design, including  traffic  flow, any restriction  of access
     or exits, ventilation, the type of surface  finish, the location and  type
     of water outlets,  and  the accessibility of  shutoff valves or  switches  for
     air conditioning,  electricity, water,  gas,  etc.
10.  the presence,  correct  functioning, and use  of protective devices (e.g.,
     interlocks,  warnings devices,  evacuation  alarms,  ventilation  failure
     alarms, emergency  shutoff switches)
11.  the possibility  of bypassing  protective devices without  adequate warning
12.  the posting  of radiation  areas
13.  the correct  labeling of radioactive materials and radiation sources
14.  the adequacy of  and compliance with procedures  for controlling  personnel
     radiation exposure and for controlling the  spread of contamination during
     the handling,  storage, transportation, and  disposal  of  •, adioactive sources
                                     4.11

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DARCOM-P 3E5-1
 15.  the availability, adeauacy, and correct functioning,  calibration,  and use
      of survey and monitoring equipment
 16.  the adequacy of and compliance with routine survey and monitoring
      procedures
 17.  the existence, adequacy, and eisplay of emergency plans and the familiar-
      ity of personnel with these plans
 18.  the status of personnel rsdiation protection training.
      D.  Obtaining Equipment and Materiel.  After evaluating what type  of
 radiation and/or contamination  (alphu, beta, gamma, neutron) can be expected,
 the surveyor should decide what radiation detection and measurement equipment
 is needed.  The information in  Chapter 2 of this manual, "Radiation Instru-
 mentation," is useful for this  determination.  Other miscellaneous equipment
 and materials may be needed, for example, clipboards, survey report forms,
 smears, protective clothing, shoe covers, and disposable plastic gloves.
 4.2.2  Inspection and Measurement
      When adequate preparation  has been made, the inspection can be started
 and measurements made.  The radiation protection staff member who is respon-
 sible  for conducting the inspection and making radiation measurements should
 be aware of the controls needed to ensure that his/her own radiation exposure
 is kept ALARA.  Personnel dosimeters, protective clothing, and  respiratory
 equipment should be used when appropriate, and the surveyor should ensure that
 radiation generators, source-shielding mechanisms, or source-handling equip-
 ment cannot be operated except  under his/her control during the survey.
      A.  Inspection.  The inspection of a facility is conducted to:  1) pro-
 vide firsthand knowledge of the installation, personnel, surroundings,  radia-
 tion sources, and equipment; 2) assess where radiation measurements should  be
 made;  and 3) assess the presence and effectiveness of each physical safeguard
 and the extent of compliance with procedural controls used for  radiation  pro-
 tection.  The checklist prepared prior to the start of the survey should  be
 usefu". in identifying the items to be inspected.  The surveyor  should be  alert
 for any deviation from written  plans and procedures.
                                      4.12

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                                                               DAKCOK-P 3£5-1
      B.  Measurements.  The places identified for measurements during the
 facility inspection  should be monitored for contamination, end measurements of
 the  radiation  field  produced by sources should then be made.  Specie monitor-
 ing  procedures  are described later in this chapter,
 4.2.3  Evaluation and  Recommendations
      When  all  inspections and measurements have been made, the results should
 be evaluated to determine the overall radiological status of the facility.
 The  evaluation  should  include a determination of any significant levels of
 contamination  and any  significant dose rates produced by sources, and the iden-
 tification  of  any deficiencies it. the radiation protection program.  Recommen-
 dations for corrective action should be made so that dose equivalents ere kept
 ALARA.  Such recommendations may include changes in:
  1.   operational factors (e.g., time spent by personnel in radiation areas,
      equipment  use time, or methods of operation)
  2.   shielding  (e.g.,  size, thickness, type of material, or location)
  3,   manipulative equipment, particularly relating to the equipment's speed of
      operation  and the distance of personnel from sources
  4.   procedural  controls, particularly those that eliminate unnecessary per-
      sonnel exposure or contamination
•  5.   personnel  protection or warning devices
  6.   survey and monitoring procedures
  7.   personnel  monitoring and survey equipment
  8.   plans  of action for accidents or emergencies
  9.   personnel  training.
      A resurvey may  be needed after corrective action is  taken,  to  ensure  that
 the  changes made are effective.
 4.2.£  Survey Records end Reports
      Recorcs of radiation surveys are needed for assessing  tne effectiveness  of
 the  radiation protection program.  They nay also be useful  in  interpreting  the
                                     4.13

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    M-P 385-1
results of personnel  mon1 v^v-y.   Survey reports  should  contain  the  following
information:
 1.  date and time of the  survey
 2.  general location of the survey (building and room)
 3.  specific locations and objects where radiation  measurements were  made
 4.  purpose of the survey (e.g., leak test of sealed  source;  routine  survey
                i
     for contamination on  floors  and other surfaces; or  survey to establish
     dose rates to personnel)
 5.  identification (type  and serial number) of the  radiation  detection
     instruments used to perform  the survey
 6.  measurement results and conditions observed  (e.g.,  dose  rates and
     contamination levels)
 7.  conclusions and recommendations
 8.  identification of the individual performing  the survey.
A facility diagram may be  attached directly to the report and  used to note the
dose rates and contamination levels observed during  the  survey.
     More information on records  of surveying and monitoring  activities can be
found in Chapter 13, "Recordkeeping."  The degree of detail  included in survey
records must be sufficient to make them meaningful after the  passage of several
years.   Records should be  kept for at least 5 years.
                 Section 4.3  SPECIFIC MONITORING PROCEDURES

     Procedures for measuring radiation fields and contamination, for leak
testing sealed sources, and for personnel monitoring, air monitoring, and
tritium monitoring are described below.  Information on the instrumentation for
these procedures is given in Chapter 2, "Radiation Instrumentation."
                                     4.14

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                                                              DAKCOM-P 385-1
4.3.1  Measurements  of Radiation  Fields
     Measurements of radiation fields — the  areas around  sources  thet  receive
radiation from the sources—are made  to  provide a  basis  for estimating  per-
sonnel exposure and  for determining the  effectiveness  of procedures used  for
radiation protection.   The number of  measurements  to be  made depends  on how
much people move about within a given  field and how much the field varies in
space and time.  If  the radiation field  is  fixed,  as in  many x-'-ey installa-
tions, few measurements are required.  However, if the radiation pattern  is
variable, such as during the removal  of  a  source from  a  shielded container,
more measurements are required.  In the  extreme case,  it may be  necessary to
continuously monitor work in progress.   The intensity  of tne radiatirr,  should
be measured using dose rate ''nstruments  in  locations occupied  by personnel.
The measurements should be recorded on a deta  sheet  or on a floor plan  corre-
sponding to the area monitored and should  be compared  with specified  limits.
     Procedures for  calculating external exposure  are  discussed  in Chapter 6.
It may be useful, when planning the control of an  individual's occupational
expDsure, to compare short-term measurements in a  radiation field with
estimates of the dose equivalent  that would be received by an  individual  who
worked in that field for extended periods  of time.  For example, if  the
maximum dose-equivalent rate for  a particular radiation field  is 10  mrem/hr,
and if an individual worked in that field  for 5 hours  each week, the expected
dose-equivalent rate would be:  10 mrem/hr x 5 hr/wk = 50 mrem/wk.   The
results of this type of conversion can be  compared direct!} with applicable
administrative or regulatory limits.
4.3.2  Measurements  of Contamination
     Familiarity with the work performed in e radiation area  is essential for
determining what type of surface  contamination is  most likely to be  present,
where it is likely to be, and whether it is likely to be fixed or removable.
Fixed,_or nonremovable, contamination contributes  to external  exposure.  Remov-
able contamination can enter the  body and contribute to internal exposure.
     Because removable contamination  can be spread end  presents an internal
hazard,  the member of the radiation protection staff who  is measuring the
contamination must be careful to  avoid both exposure to himself and the  spread
                                     4.15

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DARCOM-P 385-1
 of contamination.  The surveyor snoulo wear adequate protective clothing dur-
 ing.the survey, taking care to avoid contamination of hands, clothing, and
 radiation detection  instruments.  When only gamma radiation is present, the
 detection instrument can be entirely covered by a thin plastic materiel for
 contamination control.  The sensitive areas of the detector must no* be covered
 when  alpha  radiation is present.  Shoe covers, gloves, instruments, and other
 equipment used during an extensive survey should be monitored periodically
 during the  survey,.   As soon as the entire survey has been completed, protec-
 tive  clothing should be removed and surveyed for contamination, together with
 the instruments and  equipment used.
       Direct measurements using portable instruments can be used to determine
 the total amount of  fixed  and removable contamination present.  An indirect
 measurement technique is used to detect removable contamination.  These two
 techniques  are described below.
       A.  Pirect_Measurements.  Any area within a facility where there  may  be
 contamination should be systematically monitored with a sensitive detection
 instrument.  During  the measurement, the probe should be held close to (within
 0.6 cm of)  the surface.  To prevent instrument contamination and damage, the
 probe must  not contact the surface.  The probe should be moved  slowly  over the
 surface to  allow the Instrument time to respond.  Instrument readings  should
 be recorded on a data sheet or on a floor plan of the area  being monitored.
       B.  Indirect Measurements.  A smear taken from a surface that may be  con-
 taminated can be used to monitor for removable contamination.   A smear test  is
 considered  an indirect measurement of contamination.
       To perform a smear test, a floor plan of the facility  to be monitored is
 needed, as  well as small pieces of paper, such as filter paper  discs,  to  be
                                                     2
 used  as smears.  A smear 1s taken by wiping a 100-cm  portion of the  surface
 to be monitored.  The Items or areas from which  smears are  taken are  identified
 on the floor plan.   The smear should be removed  from  the facility  being mon-
 itored and  counted according to specified laboratory  procedures.
       Care should be  taken  to avoid touching either  the surface  being  monitored
 or the contaminated  side of the smear, and to keep  the probe  from  touching the
 smear.  Cros:-contamination of the smears can be avoided by placing  each  smear
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                                                               DARCOM-P 3E5-I
 in  an  individual  envelope  immediately after the smear is taken.  Smears shoulc
 be  treated  as  radiation sources and handled according to radiological safety
 procedures.
     C.   Action  Levels and Reporting.  The results of monitoring for both fixed
 and removable  contamination should be comparec with the contamination limits
 given  in  Appendix A.  The  actions to be taken if the levels found exceed the
 limits are  also  identified in the table.
 4.3.3  Leak Testing Sealed Sources
     The  instruments and supplies needed for leak testing sealed sources are
 1)  a remote-handling tool, 2) sheets of paper with impermeable backing (or
 sheets of ordinary paper and sheets of polyethylene film), 3) discs of filter
 paper  that  have  a high wet strength (for making smears), 4) envelopes, 5) rods
 of  wood,  plexiglass, aluminum, or some other material, 6) adhesive tape, and
 7)  a radiation detection instrument.
     Before a  leak test is begun, a data sheet should be started that includes
 a description  of the source, the type of leak test to be performed, the date
 of  the leak test, and the name of the person performing the test,  Space should
 be  left on the data sheet so that the results of the leak test in yd and any
 action taken as a result of the test can be recorded later.
     Leak testing should be planned so that the surveyor'.* exposure is kept to
 a mimimum.  The dose rates at given distances from the source should be cal-
 culated so that shielding needs, the length of the remote-handling tool needed,
 and the time allowable near the source can be determined.  A rule of thumb is
 to  plan an operation so that the person performing a test or a series of tests
 does not receive a whole-body dose in excess of 5 mrem.  "Dry runs" can be
 performed if desired.
     It is always a  safe procedure to assume that a source is leaking and to
assess  the physical  provisions and operations that would be needed to deal
with a  contamination incident.  Knowing the construction of the source is
 important so that leak testing does not d-maoe the source.  Protective rubber
gloves  should  be worn  during the test.
                                     4.17

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ARCON-f 365-1
     A.  Direct Leek Testing.  This method is applicable to seeled sources that
are not in a container, or that are in a container but are not fastened in it,
and that can be handled safely with available equipment anc facilities.  The
total  whole-body dose received during the test should not exceed 5 mrem.   This
procedure must be performed in a hood or glove box rather than on an open bench
top to prevent possible contamination of the work area.
     A sheet of impermeable paper (or paper backed with a polyethylene sheet)
should be placed on the working surface and taped down if necessary, to prevent
contamination of the working surface if the source is leaking.  A clean filter
paper  disc should be marked to indicate the particular source being leak
tested.  If the source contains water-soluble radioactive material, the filter
paper  smear should be dampened with distilled water.
     When a contained source has been removed from its shielded container,
using  the appropriate remote-handling tool and observing applicable radio-
logical safety procedures, all of its surfaces should be carefully wiped.  The
source should then immediately be replaced in its container.  Dry smears (or
wet smears that have been allowed to dry) should be checked with an instrument
that monitors low levels of alpha or beta-gamma radiation, as appropriate.
Readings should be taken with the open window of the probe near the smear but
not touching it.  If contamination is detected, the source is likely to be
leaking, and precautionary measures should be taken to avoid unnecessary
exposure of personnel until the situation has been fully evaluated.  The smear
should be counted according to specified laboratory procedures in order to
obtain quantitative results.
     B.  Indirect Leak Testing (Container Interior).  This method is applicable
to sealed sources that are not in a container, or that are in a container but
are not fastened in it, and that have activity levels that prevent  safe direct
leak testing with existing equipment and facilities.  The test or series of
tests  should be planned so that the radiation protection staff member  perform-
ing it does not receive a whole-body dose in excess of 5 mrem.
     For this test, a contained source is removed from its normal shielded
container and transferred to an alternate shielded container or temporary
shielding set up specifically for this purpose.  An appropriate monitoring
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                                                               DARCOM-P  385-1
instrument should be used to ensure that the  source  in  the  temporary  housing
is adequately shielded.  In addition,  instruments  for monitoring  low-range
beta-gamma r-r alpha radiation should be used  to  monitor eccessib'ie  surfaces of
the empty container.  Any positive readings  should be recorded, and if con-
tamination is detected, further precautionary measures  snould  be  taken before
the leak test is continued.
     For this test, smears of the inside surfaces  of the empty source con-
tainer are taken, particularly of areas normally in  direct  contact  with  the
source. The smearing device should consist of a  rod  (of wood,  plexiglass,
aluminum, or other material) long enough to  reach  the area  to  be  wiped,  with  a
filter paper smear attached to one end.  If  the  source  contains water-soluble
radioactive material, the filter paper should be moistened  with distilled
water.  The wet or dry smear should be rubbed on the inside surfaces  of  the
empty container, especially on the surfaces  that most closely  contact the
source.  Dry smears, or dried wet smears, should be  checked with  a  low-range
beta-gamma or alpha-monitoring instrument, the readings taken  with  the open
window of the probe near the smear.  If contamination is detected,  steps should
be taken to prevent unnecessary exposure of  personnel until the situation  has
been fully evaluated.  The smear should be counted according to specified
laboratory techniques in order to obtain quantitative measurements.
     C.  Indirect Leak Testing (Container Exterior). This  method is applic-
able to sealed sources that are fastened in  a container.  It is also applic-
able to other sealed sources that cannot be  leak tested safely with existing
facilities and equipment.
     The portions of the shielded container or device where contamination would.
be expected to appear if the sealed source were leaking should be smeared using
the rod-and-smear device described above.  All applicable radiological safety
procedures should be observed, and the smear should be  counted in the same
manner as used for the interior indirect leak test.
4.3.4  Personnel Monitoring
     Personnel are monitored to determine whether contamination  is present on
them and to measure internal and external exposure.   Personnel monitoring
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DAKCOM-P 3S5-1
 serves two purposes:  1} to assure that all  exposures  are maintained  ALARA,
 and 2) to identify any unsuspected source of exposure  so that  prompt  correc-
 tive action can be taken.
      A.  Contamination.  Personnel must be monitored for contamination be*'ore
 leaving any area in which radioactive materials  or sources  are user or scored.
 If an individual is contaminated, follow-up surveys must be nade to determine
 the source of contamination and to detect any contamination that may neve been
 spread by the individual.  Prompt corrective action must be taken to eliminate
 the source of contamination.
      A sensitive detection instrument should be  used to monitor personnel.
 Skin and clothing should be carefully monitored, with  an emphasis on the head,
 hands, and feet.  Any point that shows visible signs of contact, such as dirt,
 grease, or liquid stains, should be monitored.  In addition, any surface known
 to have come in contact with equipment or contaminated surfaces should be
 monitored.
      The probe of the instrument should be held  close  to  the individual's skin
 or clothing but must not be allowed to contact it.  The probe should be moved
 slowly to allow time for the instrument to respond.
      B.  Internal Exposure.  The principal objective  of internal personnel
 monitoring is to determine whether radionuclides have  entered the body.  The
 routine determination of internal contamination  is necessary only in facilities
 where unsealed radioactive materials may become  airborne.   Internal personnel
 monitoring should also be considered whenever a  routine survey  indicates
 significant levels of contamination.
      Internal dose is determined using two indirect methods:  1) radiochemiccl
 analysis, which is the measurement of radioactivity in urine, feces, blood,
 secretions, and body tissues; and 2) in-vivo  (or whole-body) counting, which
 is the measurement of radiation emitted from  the body, using an external
 detector.  These procedures are highly specialized.  More  information on their
 use end on the control of internal exposures  is  provided in Chapter 5,
 "Internal Exposure."
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                                                             DAKCOK-P  385-1
     C.  External Exposure.  The exte-nal  whole-body dose  to  an  individual  is
estimated using personnel dosimeters.    A personnel  dosimeter should  be worn  by
each individual who is occuoetionclly  exposed to sources  of  ionizing  radiation.
Dosimeters must be worn in rcdietion areas and should be  worn by anyone who
periodically enters £ controlled arec  and is likely  to receive more  than 5* of
the quarterly dose-equivelent limit listed in Table  3.2 (Chapter 3).   An indi-
vidual under the age of 18 who enters  a controlled area and  is likely to
receive more than.5« of the quarterly  dose-equivalent limit  for  minors  should
also use a personnel dosimeter.  The dosimeters designated by the DA  and other
methods of controlling external exposures are described in Chapter 6, "External
Exposure."
4.3.5  Air Monitoring
     The purpose of air monitoring is  to determine the cleanliness of the air
in the work area.  The need for stringent controls on airborne radioactivity
should be stressed in SOPs.  High concentrations of  airborne  radioactive con-
tamination can lead to contamination of surfaces in  a facility or the environ-
ment, and can result in internal exposure to personnel.
     Inhalation is the principal means by which radioactive materials can
enter the body.  The amount of material deposited in the body depends largely
upon the concentration in the air inhaled, the particle size  of  the contami-
nant, and the length of time the individual is exoosed.  Control levels for
various isotopes are given in 10 CFR 20, Appendix B, Table I.  To determine
whether control levels are being met,  routine air samples are collected and
evaluated.
     Criteria for the development of an air monitoring program are given in
Chapter 5.  Several  useful references  are included in the bibliography.
Equipment used to monitor air is discussed in Chapter 2, "Radiation
Instrumentation."
4.3.6  Tritium Monitoring
     Tritium is a radioisotope of hydrogen that decays to helium by the emis-
sion of a beta part :le with a maximum energy of 18 keV and an average energy
of 5.7 keV.   The weak beta particle has a maximum range of 6 ym in water or
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DARCOM-P 385-1
 0.5 cm in air.  When released to the environment,  tritium can  enter biological
 materials by several routes.  It can be taken into the  body in water,  in foods,
 or as tritium or tritium oxide in inhaled air.   In both gaseous and liquid
 forms, tritium can readily penetrate directly through human skin surfaces.
 Tritium's ability to be readily incorporated into  biological  systems makes it
 of concern from the standpoint of internal exposure.
      The low energy of the beta particle emitted by tritium creates a special
 monitoring problem.  Portable detection instruments cannot be  used because the
 distance between the tritium source and the detector is usuall.y greater than
 the particle's range, end even in detectors with a window, the window may be
 too thick to be penetrated by the beta particle.  Windowless gas-flow propor-
 tional counters and liquid scintillation counters  are  therefore used to assay,
 or test, for tritium.  In the special case of tritium  gas, ionization chambers
 may be used.  These instruments are described in Chapter 2, and their applica-
 tion for monitoring tritium levels in water, in urine,  on surfaces, and in air
 is reviewed briefly below.  Additional references  specific to tritium measure-
 ments are provided in the bibliography at the end of the manual.
      A.  Water.  The maximum permissible concentration  (MPC) of tritium in
 drinking water is 3 x 10   yCi/ml (10 CFR 20, Appendix B).  This MPC corre-
 sponds to 110 disintegrations per second in each ml of water  (dps/ml).  Liquid
 scintillation counting is the method of choice for measuring  tritium in
 water.
      B.  Urine.  A radioassay for tritium in urine should be  performed every
 2 weeks for all personnel who routinely work with tritium, and  immediately
 following any unusual occurrence involving the spread of tritium contamination.
 If tritium is found in urine, additional urine samples should  be obtained daily
 to determine the biological half-life of the tritium deposited  in  the body.
 Biological half-lives between 7 and 12 days are commonly observed.
      Several hours are needed before tritiated water becomes  equally distri-
 buted throughout the body.  Consequently, urine samples  should  not be taken
 immediately after a :otenf;cl tritium inhalation.  Generally,  2 to 4 hours
 should elapse between the time of the exposure and the time of sample collec-
 tion.  When a urine sample is collected, personnel should  remove all protective
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                                                             DAKCON-P
clothing and wash their nands  to  avoid  contaminating the sample.  The
sample should oe placed in an  eir-tignt container and refrigerated.  Liquid
scintillation counting is used for  the  radioessay of tritium in urine.
     C.   Surfaces.   Because the energy  of  the beta particle emitted by  tritium
is too low to allow the particle  to enter  portable detectors, ?. smear test
should be used to monitor for  surface contamination.  The procedure is  similar
to that described in Section 4.3.2  except  that the smear should be lightly
coated with glycerin or moistened with  water to increase its efficiency in
collecting contamination.  Smears should be placed into vials immediately after
each sample is taken.   The sample can be counted using liquid scintillation.
     D.   Air.  In air, tritium occurs primarily as water vapor  or  hydrogen  gas.
Flow-through ionization chambers  and proportional counters can  be  used  to
monitor air for tritium.  Ionization chambers cannot distinguish tritium from
some other types of radioactive particles  and are sensitive to  interference
from cigarette smoke,  aerosols, and external gamma fields.  Ges-flow  propor-
tional counters can partially  discriminate against other radionuclides  and  are
less sensitive to aerosols.  The  sensitivity of ion chambers  is similar to  that
of qas-flow proportional counters (in the  pCi/cm  range).  To detect  tritium
                               3
levels much below about 1 pCi/cm   in air,  it is necessary to  remove tritiated
water vapor from the air using silica gels and bubblers.  Information on this
procedure is given in  Report No.  47 of  the National Council on  Radiation Pro-
tection and Measurements (NCRP 1976).   Liquid scintillation counting  can be
used to essay the water vapor  samples.
                Section 4.4  NONMEDICAL X-RAY INSTALLATIONS^

     X-ray equipment poses e potential  hazard,  both for those who operate it
and for those who may be in the vicinity,  because of the extremely high dose
rates generated by the devices at the flio of a switch.  Extensive engineered
safeguards and administrative controls  are used to minimize normal operating
(e) For -his section of the manual,  an installation is defined as the space
    occupied by a radiation-generating source with its associated equipment.
                                     4.23

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DARCOM-P 385-1
 exposures and prevent accidental exposures.  Radiation protection is accom-
 plished through the combined efforts of the manufacturers of the devices, the
 designers and builders of the installations where the devices are used, the
 operators of the equipment, and radiation protection personnel.
      Requirements for the design and operation of x-ray installations are dis-
 cussed in two reports of the American National Standards Institute, ANSI N543-
 1974 and ANSI N537-1976.  Installations, including necessary shielding, should
 be designed by a, qualified expert and should meet applicable regulations of
 federal, state, and local agencies.
      This section describes the classification of nonmec'ical x-ray installa-
 tions, the engineerec and administrative safeguards used in them to minimize
 exposures, and procedures for surveying them.  A discussion of surveys for
 medical x-ray installations is beyond the scope of- this manual; information on
 this- topic can be found in NCRP Report No. 33 (1968).
 4.4.1  Classification of Nonmedical X-Ray Installations
      Installations are divided into four classes, which are described briefly
 below and in greater detail in ANSI 543-1974.  A separate classification for
 x-ray diffraction and fluorescence analysis equipment is described in ANSI
 N43.2-1977.
      A.  Protective Installation.  An x-ray unit within a permanent, shielded
 enclosure is considered a protective installation if the exposure  rate at  any
 accessible surface of the enclosure is less than 0.5 mR/hr during  operation of
 the device.  Personnel may not remain inside  the enclosure during  irradiation.
      B.  Enclosed Installation.  An enclosed  installation is similar to  a  pro-
 tective installation in that the x-ray ur.it is within a permanent, shielded
 enclosure.  However, a higher exposure level  is allowed for this class of
 installation.  The exposure rate at any accessible,  occupied area  30 cm  from
 the outside surface of the enclosure must not exceed  10 mR/hr and  the  expos-ure
 rate at any accessible but normally unoccupied area  may not exceed 100 mR/hr.
 During operation of the device, personnel may not remain  inside  the  enclosure.
      C.  Unattended Installation.  An x-ray unit in  a shielded enclosure that
 is small enough to prohibit personnel occupancy  is  considered an unattended
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                                                             D/vRCOM-P 3E5-1
installation if the exposure rete 30 cm from the  outside  surface  of  the  device
does not exceed 2 mR/hr during operation of  the unit.   The  shielded  enclosure
may not be used for any purpose other than  to enclose  tne  x-rey unit.
     D,  Open Installation.  An x-rey unit  that is  not in  a  shielded enclosure
and that is located in en aree that may potentially be occupied by  personnel
during operation of the device is considered an open installation.
4.4.2  Engineered Safeguards
     Engineered safeguards are safety systems such  as  warning  devices,  shields,
and interlocks that are built into either the >-ray installation  or  the  x-ray
device itself.  They should be designed by  a qualified expert  in  accordance
with the requirements of the installation class.  The  fail-safe principle is
used whenever possible in the design and construction  of  safety systems.   A
fail-safe system is a system in which any malfunction, including  malfunction
of the safety system, causes the device to  stop functioning or to fail  in a
manner that does not expose personnel to radiation.
     Examples of the engineered safeguards  required for each installation class
are described below.  Greater detail can be  found in ANSI  N543-1974.  Engi-
neered safeguards for x-ray diffraction and  fluorescence  analysis equipment
are described in ANSI N43.2-1977.
     A.  Protective Installation.
 1.  Each machine must be totally enclosed  within physical  barriers  that have
     sufficient shielding to reduce exposure rates  during operation  to less
     than 0.5 mR/hr at all points accessible to personnel.
 2.  All  entrances to the installation must  have  a  fail-safe interlock system
     that prevents inadvertent entry during  machine operation.
 3.  The  enclosure must be equipped for emergency exit when the  doors are
     locked from the outside.   A least one  clearly  marked scram  button
     (emergency power-cutoff switch) must be located conspicuously in the
     exposure room.   Enough switches must be installed to allow  a person to
     reach  a  switch within 5 sec after e warning  alarm is activated.  The
     purpose  of the scram button must be clearly  marked.
                                     4.25

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DARCOK-P 385-1
  4.  Fail-safe visiole and aucible warning signals within  the  enclosure must
      be actuated et least 20 sec before irradiation begins.  The  visible
      sianal must stay on during the entire operation of  the  equipment.   Speci-
      fications for audible signals are provided in ANSI  N2.3-1967.
  5.  A steady red light activated by the control circuit must  be  located out-
      side the entrance to each enclosure.  A warning sign  showing the radia-
      tion symbol and the words "Caution:  Entering Radiation Exposure Room"
      must also bfe posted.
      B.  Enclosed Installation.  The engineered safeguards for protective
 installations also apply to enclosed installations with  the  exception of item
 1  pertaining to exposure rates.  For enclosed installations, each machine must
 be totally enclosed within physical barriers that have sufficient shielding to
 reduce operational exposure rates at all accessible and  occupied points to
 less than 10 mR/hr and at all accessible but normally unoccupied points to
 less than 100 mR/hr.  The following additional  safeguards  are also necessary:
  1.  All accessible areas in which the exposure rate exceeds 5 mR/hr must be
      posted with a sign showing the radiation symbol and the words "Caution:
      Radiation Area."
  2.  All entrances to the installation must have  a  sign posted showing  the
      radiation symbol and the words "Caution:   Entering Radiation Area."
      C.  Unattended Installation.
  1.  The exposure rete at any accessible area 30  cm from the outside surface  of
      the shielded device may not exceed  2 mR/hr during  operation.  Service
      doors to areas with exposure  levels higher than 2  mR/hr must be locked.
  2.  The device must be posted with a  sign  showing  the  radiation symbol  and the
      words "Caution:  X-Rays."
  3.  A steady red light that is activated by  the  control  circuit must  be
      installed near the head and beam  port(s)  of  each device.
  4.  All beam port; thet are not  in use  must  be secured in  a  closed  position in
      a manner that prevents their  casual opening.
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                                                                  y-P 3E!-1
 5.   The shielding must  be  secured  in  a  manner  that  prevents  its  cesoal  remove!
     or the exposure of  personnel.
     D.  Open Installation.
 1.   A steady red light  that is  activated  by  the  control  circuit  must  be  mounted
     on or near the source  of radiation.
 2.   Steady or fleshing  red .lights  activated  when the  device  is operating must
     be located at the radiation area  boundary  in sufficient  numbers  to  ensure
     that at least one is  visible from each avenue of  approach.
 3.   The perimeter of any  area where the radiation level  exceeds  5  mR/hr must
     be posted with a sign  displaying  the  radiation  symbol  and  the  words
     "Caution:  Radiation  Area."
 4.   The radiation source  and all exposed  objects must be within  a  conspicu-
     ously posted perimeter that limits  access  to arees where the exposure
     rate is greater than  100 mR/hr.  A sign  displaying the radiation symbol
     and the words "Danger:   High-Radiation Area" must be posted  at tre
     perimeter of this area.  During periods  of unattended irradiation,  this
     area must be locked to prevent access.
4.4.3  Administrative Controls
     Administrative controls are procedures used to  minimize the  radiction
exposure of operating personnel.  These procedures require the  cooperation of
radiation protection and operations personnel.   Enclosed, unattended, and open
installations require more  extensive administrative  controls than protective
installations because of their higher potential exposure rate.
     A.  Training.  All  individuals who use x-ray equipment must be trained to
operate it safely.  Information on tne content  of a  training program can be
found in NCRP Report No. 61 (1978).
     B.  Standing Operating"Procedures.   An SOP should be prepared for each
x-ray device.  The SOP should be posted where it is  easy to see, on or next  to
the  console for the device, and should contain  the following information:
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DARCOK-P 3S5-1
  1.  the class of the installcti&n
  2.  survey and monitoring requirements
  3.  a list of all required administrative and engineered safeguards
  4.  operating procedures
  5.  emergency procedures
  6.  a list of trained operators
  7.  the name of the individual responsible for the device.
      C.  Operation and Maintenance Logs.  The individual  responsible for an
 x-ray device should keep two log books:  an operations  log and a maintenance
 log.  The operations log should contain a complete description of all work
 performed with the device.  The maintenance log should  contain a description
 of all maintenance work.  All log entries should be signed and dated.
      0.  Radiation Area Requirements.  X-ray units'must be operated only within
 a radiation area.  When a qualified operator is not present, the area must be
 locked or else the device must be locked prevent its unauthorized operation.
 Before using the device, the operator must make sure that only required per-
 sonnel are present within the area and that any exposure of personnel within
 the area will be minimal.
 4.4.4  Surveys of Nonmedical X-Ray Installations
      Surveys of nonmedical x-ray installations should include both physical
 inspection of the facility and measurement of radiation levels.  Each installa-
 tion should be inspected to verify the current and expected occupancy of all
 areas surrounding the installation.  Devices that affect radiation protection
 (e.g., audible and visible warning signals, shielding,  interlocks, and devices
 that restrict the positioning of radiation sources) should be inspected to
 determine whether they are operating properly.  Administrative controls for
 each class of installation should be reviewed.
                                *
      A.  Frequency.  All new installations must be surveyed before routine
 operation is begun.  Existing facilities should be surveyed every 6 months or
 whenever changes in the installation could affect  radiation protection
 procedures.
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                                                             DARCOM-P  3£5-i
     B.   Procedure.   The RD0 should maintain  a  'ii:t  of  ell engineered  end
aoministrative safeguards necessary for  the  sefe  ooerction of each  nc-r.medical
x-ray installation,   bei ore becinninc  a  survey  of  an installation,  the  RPO or
a member of the radiation protection staff  should  review  this list  and  the
general  procedures outlined in Section 4.2.   The  following items  should bc-
included on the check list for the inspection:
 1.  Check for a posted,  up-to-date SOP.  All operators'  names  must be  listed
     on  the SOP.
 2.  Check for modifications to the device  that may  affect any  safety  system
     (e.g., shielding,  interlocks).
 3.  With the device operating at full power, check  for measurable  beams of
     radiation at ell appropriate locations.  Measurements should be taken at
     all points accessible to personnel  and  in  other normally occupied  spaces,
     such as offices not related to machine  operation.  A strong  effort must
     be  made to maintain exposure rates  ALARA even if they fall within stated
     guidelines.  Thus,  it is important  to  determine and  document any  exposure
     rate that could be  reduced by administrative or engineered safeguards.
 4.  Test all engineered safeguards listed  on the SOP,  including  interlocks,
     warning lights, alarms, and scram buttons.
 5.  Verify that the device is operated  within  a  radiation area that is
     adequately posted.
 6.  Determine that  all  operators are  trained.
     C.   Radiation Survey Report.  A report  of  a  radiation survey of an instal-
lation should include:
 1.  who conducted the survey and the  date  of the survey
 2.  the dev'ice and  installation being surveyed,  identified  by  suitable means
     (e.g., serial number, room number,  and  building number  or  name)
 3.  the survey instrument used end the  date  of its  lest  calibration
 4.  tne potential and current at v.nich  an  x-ray  tube v.cS operated during  the
     survey, and any measured x-ray beams
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DARCOK-P 385-1
  S.  the classification of the installation
  6.  the location of tne x-ray source and the orientation  of  the x-ray beam in
      relation to each exposure measurement (a diagram may  be  useful)
  7.  a description of all engineered and administrative  safeguards along with
      a verification that they were tested or inspected
  8.  all deficiencies found during the survey and the corrective action to be
      taken.
                   Section 4.5  ENVIRONMENTAL SURVEY PROGRAMS

      An environmental survey is a systematic investigation and measurement of
 radiation levels and radioactive contamination levels in the environment sur-
 rounding a facility.  The objectives of an environmental survey program
 include:
  1.  assessment of the natural radiation and radioactivity levels in the
      environment before operations begin
  2.  assessment of the actual or potential exposure of man from the additional
      radioactive materials or radiation contributed to the environment by the
      facility, or estimation of the probable upper limits of such exposure
  3.  determination of the fate of contaminants released to the environment
  4.  detection of sudden changes and evaluation of long-term trends, which can
      indicate failure or lack of adequate control in the operation of the
      facility
  5.  demonstration of compliance with applicable regulations and legal
      requirements concerning releases to the environment.
      The extent of an environmental survey program depends on  several factors,
 including the .nature of the facility, the type and quantity of radionuclides
 handled, and the potential for the release of radioactivity to the
 environment.
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                                                              DARCO.M-P  365-1
     Environmental surveys should be conducted  prior to  the  initiation  of
radiological operations at e facility and at least  once  a  year thereafter.
More frequent surveys may be needed depending on  the scope and nature of tne
facility's activities.  Trie results of an environmental  survey should be used
to determine any need to modify controls or operations.
     The development of a survey program should include  the  following general
steps:
 1.  Evaluate the facility as a source of direct  radiation and radionuclides,
     especially the composition, concentrations,  release rctes, points  of
     release, and physical and chemical forms of  the nuclides.
 2.  Identify the pathways leading to exposure to man, using analytical
    • models, the experience gained at other sites,.and preoperational data on
     local meteorology, hydrology, and population distribution and diet.
 3.  Select the pathways (e.g., water, food, air) that may be most critical in
     terms of their contributions to exposure, and determine the critical
     population groups.
 4.  Determine the measurements required to provide data for dose assessment
     for normal and abnormal conditions.
 5.  Allow for flexibility in the program design.  As operational experience is
     accumulated, other types of measurements or measurement frequencies may be
     desirable.
     Details on establishing and carrying out environmental survey programs can
be found in the bibliography.
                                  REFERENCES

American National Standards Institute (ANSI).  1967.  Immediate Evacuation
  Signal for Use in Industrie! Installations Where Radiation Exposure Nay
  Occur.  ANSI N2.3, New York.
American National Standards Institute (ANSI).  1974.  General Safety Standard
  for Installations Using Non-Medicel X-Rey and Sealed Gamma-Ray Sources',
  Energies UP to 10 MeV.  ANSI N543, New York.  Also published in  1875  as
  National Bureau of Standards Handbook No. 114,  Washington, D.C.

                                     4.31

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DARCOM-P 385-1

 American National Standards Institute (ANSI).  1976.   Radiological  Safety
   Standards for the Design of Radiooraphic and Fluoroscopic Industrie! X-Ray
   Ecuipment'ANSI N537, New York.

 American National Standards Institute (ANSI).  1977.   Radiation Safety for
   X-Ray Diffraction and Fluorescence Analysis Equipment.   ANSI N43.2,
   New  York.  Also published as National Bureau of Standards Handbook No. Ill,
   Washington, D.C.

 American National Standards Institute (ANSI).  1978.   Control  of Radioactive
   Surface Contamination on Materials, Equipment, and Facilities to be Released
   for  Uncontrolled Use"ANSI 13.12  (Draft), New York.

 National Council on Radiation Protection and Measurements (NCRP).  1968.
   Medical X-Ray and Gamma-Ray Protection for Energies up to 10 MeV - Equipment
   Design and Use.  NCRP 33,  Washington D.C.

 National Council on Radiation Protection and Measurements (NCRP).  1976.
   Tritium Measurement Techniques.  NCRP 47, Washington, D.C.

 National Council on Radiation Protection and Measurements (NCRP).  1978.  Radia-
   tion Safety Training Criteria for  Industrial Radiography.  NCRP 61,
   Washington, D.C.

 U.S. Code of Federal Regulations.  1982.  Title 10, Part 20, "Standards for
   Protection Against Radiation."  U.S. Government Printing Office,
   Washington, D.C.

 U.S. Code of Federal Regulations.  1982.  Title 10, Part 30, "Rules  of General
   Applicability to Licensing of Byproduct Material."  U.S. Government Printing
   Office, Washington, D.C.

 U.S. Department of the Army, Headquarters, Army Materiel Command.  Safety -
   Radiation Protection.  DARCOM-R 385-25, Washington, D.C.

 U.S. Department of the Army, Headquarters.  Safety -  Ionizing  Radiation
   Protection (Licensing, Control, Transportation/Disposal, and  Radia-~
   tion Safety).  AR 385-11, Washington, D.C.
                                       4.32

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                                                                                    DARCO.M-P 365-1
                                              APPENDIX  A
         MAXIMUM  PERMISSIBLE CONTAMINATION LEVELS  FOR  INANIMATE  OBJECTS1
                                                               npir Mpnt
  Contaminated  Item

1) Personal  clothing,
   including shoes
2) Protective  clothinc;

   t.  General
   b. Respirators
   c.  Laundry
3) Work areas, and
   equipment t*1'

   i. Uncontrolled
   b. Controlled
      (1)  Areas
      (2)  Hoods
      (3)  Glove boxes
      (4)  Workbench
          surfaces
      (5)  Other
          equipment

4) Tools,  equipment,
   containers
5) Vehicles

   a.  Used In  con-
      trolled  areas
   b.  Used  1n  uncon-
      trolled  areas
                            Corrective
Replace,  decontaminate,
or store  until  radio-
active contaminetior
has decavec  if  above:
Replace,  decontaminate,
or store  until  radio-
active contamination
has decayed if  above:

Replace,  decontaminate,
or store  until  radio-
active contamination
has decayed if  above:

Release only to
licensed  launoerer  if
contaminated
Control  and post, then
decontaminate  if above:

Decontaminate  (or if
decontamination 1s
impossible, fix and
then check fixation
periodically)  if above:
Prior to nonradio-
active use,  decon-
taminate if  above:
Decontaminate  (or  if
decontamination  is
impossible,  fix  and
then check fixation
periodically)  if above:

Decontaminate  if above:
                                                       200
                                                      1000
                                                       200
                                                       200
                                                       200
1000

 500
                                                                       None
                 200
                None
                30
                50
300

 30
                                                                                     C.05
              C.02
              0.6
              0.05
               0.25
0.25
                                                                      tT bets
                             Fixed'  '.     Removable ,     (mrao/hr    Removable  I
                          (dpm/100 err.  )   (dpfn':00 cr")    at  2.5  cm)   (opmMOO cr )
                                                                                                   None
             1000
                                                                                                   None
              100
1000
1000
5000
1000
1000
200
200
1000
200
200
0.2
2.0
2.5
0.5
2.0
400
2000
5000
400
2000
               100
500

100
(a) Reference:  AMC 385-25 and -AR 385-11.   (Note:  These limits may be chanoed  to  reflect  those found in
    ASS:  is.::.)
(b) Measured  with a calibrated radiation measurement  instrument.
(c) Determined usinc smears analyzed with  a  calibrated counting system.
(d) For natural and depleted uranium and for "f^ levels for alpna contamination  should be  increased
    by *  factor of £, in accordance with NRC guidelines.  ::' "°Ra is a contaminant,  levels  for alpha
    contamination should be reduced by a factor of 2.
                                                  A.33

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                                                              DARCO.M-? 3E5-1
                         CHAPTER 5.   INTERNAL EXPOSURE
5.1  CONTROL OF INTERNAL EXPOSURE  	    5.4
     5.1.1  Contamination Control Through Design Ftatures   .     .     .    5.4
     5.1.2  Contamination Surveys During the Course of Work .     .     .    5.5
     5.1.3  Decontamination of Contaminated Objects and Individuals   .    5.5
     5.1.4  Air-Sampling and Air-Monitoring Programs   ....    5.6
     5.1.5  The Use of Protective Apparel	5.7
            A.  Protective Clothing     	    5.7
            B.  Respirators	5.10
     5.1.6  Administrative Guidelines   	    5.14
5.2  MONITORING INTERNAL EXPOSURE  	    5.17
     5.2.1  Bioassay Programs      ........    5.17
          .  A.  Preparatory Evaluation  	    5.17
            B.  Exposure Control   	    5.17
            C.  Diagnostic Evaluation   .          .    .    .    .     .    5.18
            D.  Removal of Work Restrictions	5.18
            E.  Termination of Employment    .     .    .    ,    .     .    5.18
     5.2.2  Action To Be Taken Upon Detection of an Intake  .    .     .    5.18
5.3  INTERNAL DOSIMETRY CALCULATIONS    	    5.19
     5.3.1  Calculation of Acceptable Intake ......    5.19
            A.  Determining the Critical Organ     	    5.20
            B.  Calculating the Maximum Permissible Body Burden  .     .    5.21
            C.  Calculating the Maximum Permissible Concentrations
                in Air and Water	5.21
                                      5.1

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DAJRCOM-P 385-1

      5.3.2  Estimation of  Internal Dase	5.21

             A.  Calculating the Initial Dose-Equivalent Rate
                 to an Organ    .........    5.22

             B.  Calculating the One-Year Dose Commitment
                 Based on a Single In-Vivo Measurement  ....    5.22

             C.  Calculating the Fifty-Year Dose Commitment   .     .     .    5.23

 REFERENCES	5.25

 APPENDIX A  -  ICRP 30 RECOMMENDATIONS FOR LIMITING
               RADIONUCLIDE  INTAKES        	    5.27
                                     TABLES


 5.1  Parameters for  Internal Dosimetry   	     5.24

 5.2  Weighting Factors Recommended  in ICRP 30	5.30
                                        5.2

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                                                             DARCOM-P  385-1
                         CHAPTER 5.   INTERNAL EXPOSURE
     Internal radiation exposuie is the exposure  of  the  body  to radioactive
materials deposited in the body.  Radioactive materials  can enter the body
through the inhalation of radioactive dusts,  mists,  and  fumes,  the incestion
of contaminated food or water, injection via  punctur.e  wounds,  or occasionally
absorption through the skin or via a wound.
     Several methods can be used to control  exposure of  the body to external
radiation (see Chapter 6).  However, once radioactive  material  has entered the
body, there is usually no practicable method  of reducing the  internal radia-
tion exposure or the resultant dose.  Moreover, if the radioactive material
has a sufficiently long half-life, it may continue.to  irradiate the individual
for the rest of his or her life.  Because of  these difficulties, the intake of
radioactive materials into the body must be  limited  and  programs for monitor-
ing the internal exposure of radiation workers should  be followed.
     When an intake of radioactive material  is detected, estimating the result-
ing internal radiation dose is difficult for  several reasons.   First, in most
cases the quantity of radioactive material  taken  into  the body is not known.
Some procedures for assessing this quantity  partially  solve  this problem.
Second, radionuclides tend to accumulate, or  concentrate, in  specific organs
of the body, which then receive a larger radiation dose  than  do other organs.
For example, plutonium, strontium, and radium concentrate in  the bone; uranium
concentrates in the kidneys or lungs (depending upon its solubility); and
iodine concentrates in the thyroid.  Third,  a fraction of the energy emitted
by o radionuclide in an organ is absorbed within  that  organ,  while the
remainder of the energy escapes to other tissues  of the  body  or leaves the
body.  The fraction of energy emitted that results in  a  dose  to any single
organ depends on several factors, including  the type of  radiation emitted,  the
size and shape of the organ and body of the  individual,  and  the distribution
of the radioactive material within the organ or body.
     In this chapter, procedures for controlling  and monitoring internal expo-
sure and for estimatina internal dose are discussed.
                                      5.3

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DARCOM-P 385-1
                    Section 5.1  CONTROL  OF INTERNAL  EXPOSURE

      Considerable effort should be expended to  prevent  any intake of radio-
 active material through the accidental  ingestion  of  removable surface con-
 tamination or the inhalation of airborne contamination.   Removable surface
 contamination is radioactive material  that is easily moved from a surface by
 wiping or dissolution using common solvents.  Removable contamination presents
 an external hazard and, more important,  an internal  hazard if it is ingested.
 (Fixed surface contamination, which is  bound to an object, presents only an
 external hazard.)  Airborne contamination is radioactive material that has
 become airborne as a result of normal  work procedures,  suspension or resuspen-
 sion of surface contamination, breach  of containment, sputtering of heated
 fluids, or vaporization of volatile compounds.   Once airborne, the material
 may be inhaled by personnel, resulting  in an internal radiation dose.  Airborne
 contamination can present an additional  external  and internal hazard if it
 settles out of the air onto surfaces as  removable contamination.
      Because of the internal radiation  hazard posed by removable and airborne
 contamination, every means of preventing the spread of contamination should  be
 used.  The following approaches are discussed in this section:
  1.  the use of design features to limit the movement of airborne contami-
      nation and the spreading or resuspension of removable surface
      contamination
  2.  routine surveys for surface contamination
  3.  decontamination of contaminated objects and individuals
  4.  air-sampling and air-monitoring programs
  5.  the use of protective apparel
  6.  administrative guidelines.
 5.1.1  Contamination Control Through Design Features
      Design features are a key element in contamination control.  0'  particu-
 lar importance is the design of a facility's ventilation  system.  Other  design
 features, such as the elimination of surfaces from which  material  can  be
                                       5.4

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                                                             DARCOM-P  385-1
resuspended (e.c., scaffolding, open rafters,  and ceble  runs),  are clso impor-
tant in preventing contamination.   Contamination-producing  substances  snould
be used orly in hoods or glove boxes.   Such substances would  include heated
solutions; volet'le substances, such as iodine and mercury; and hiqh-specific-
activity solutions of alpha-emitting nuclides, such as     Cm  end   Sr.   See
Chapter 8 for a complete discussion of facility design.
5.1.2  Contamination Surveys During the Course of Work
     Surveys for surface contamination should  be conducted  routinely,  with the
frequency dependent upon the radiotoxicity of  the material  handled, the quen-
tity used, and the relative ease of spreading  the contamination.  In areas
containing radioactive materials that include  more than  one level  of radio-
toxicity (see Chapter 1, Table 1.10), all  removable contamination  should be
assumed to be due to the most highly radiotoxic agent until  proven otherwise.
Personnel surveys should be conducted periodically during the course of work in
a radiation area and must be conducted as  each person leaves  the area.   All
surveys should be made using the procedures discussed in Chapter 4.
     Detection equipment appropriate for the type of contamination involved
should be available.  For most nuclides that emit beta-gamma  radiation, a
Geiger-Mueller (6M) survey meter is suitable.   If the area contains low-energy-
beta emitters (e.g.,   C,   C), special survey instrumentation such as a thin-
window GM should be used.  Alpha-emitting  nuclides are best counted with
windowless proportional counters or with ZnS crysta-1 scintillation detectors.
For additional  information on instrumentation, see Chapter 2.
5.1.3  Decontamination of Contaminated Objects and Individuals
     All contamination should be cleaned up at the earliest possible time.
Contaminated objects should be decontaminated  to levels  below the maximum
permissible levels shown in Appendix A of  Chapter 4.  When an individual is
contaminated, the person responsible for decontamination should be given as
much information es possible, including the radionuclide(s) involved and the
chemical form(s) of each radionucl ia'e.  Often, ell that is known  is that the
contaminant is  a beta-gamma emitter or an  alpha emitter.  In many  instances,
the exposure may be to mixed radionuclides that emit predominantly beta-gammc
or alpha radiations.
                                      5.5

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-ARCOK-P 385-1
       Instrumentation used to assess the extent of contamination must be able
  to detect  the  radiations in question.  Use of the wrong type of instrument can
  lead  to underestimation of hazards or failure to detect any contaminants, and
  to release of  the object or individual without proper decontamination.  Decon-
  tamination procedures for both personnel and objects are discussed in detail
  in Chapter 7.
  5.1.4  Air-Sampling end Air-Monitoring Programs
       Air-sampling and air-monitoring programs have two major purposes:  1) to
  detect the presence of radioactive dusts, mists, and fumes in the air; and
  2) to quantify the amount of radioactive material in the air.  Sampling devices
  are designed simply to collect dusts, mists, or fumes; the radioactivity of
  the sampled material is quantified at a later time.  These devices are useful
  in identifying the amount and type of airborne radi'ation to which an  individual
  has been exposed.  Monitoring devices, on the other hand, detect radioactive
  material and usually sound an alarm when a specified limit is exceeded.  Moni-
  tors  are generally not as accurate as samplers; however, they do provide an
  immediate  indication of airborne radiation in the work area.
       Continuous monitoring or sampling for airborne particulate radioactivity
  should be  conducted whenever personnel have a significant potential for air-
  borne exposure because of radiological conditions in the work area.   Continuous
  air monitors should have both a visual and an audible  alarm.  Areas where  the
  potential  for  personnel exposure exceeds the limits of 10 CFR 20, Appendix  B,
  Table I, shall  be provided with an air monitor that is sensitive enough to
  alarm at <30 maximum permissible concentration-hours  (MPC-hr).   (An MPC-hr  is
  a unit that expresses the total' MFCs  an individual  has been  exposed to.   It is
  the product of the number of MPCs the individual was  exposed to and the  number
  of hours the individual was exposed.  For an individual  exposed  to  2  MPCs  for
  2 hours, for example, the product would be 4 MPC-hr.)
       When  a continuous air monitor alarms, the following actions  should  be
  taken:
  1.   Personnel  who are not wearing respiratory equipment shall  immediately
       leave the area.  However, these  individuals shall  remain  in  the  general
                                        5.6

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                                                             DARCO.M-P 385-1
     vicinity and shell  be  surveyed  for  contamination by a member of  the
     radiation protection  staff.
 2.   The Radiation Protection  Officer  (RPO)  shall be notified  immediately.
 2.   Personnel v/p.o are wearing respiratory equipment may regain  in  the area  to
     stop operations that  might be  the source  of eirborne  radioactivity.
     Other personnel may enter the  area  only if they are wearing appropriate
     respiratory equipment and only  for  the  purposes of evaluating  the source
     of airborne radioactivity or stabilizing  it.  When the  source  of the
     immediate p"obletr has  been identified and controlled, ell  personnel  shall
     leave the area.
Air samples shell be taken in  all potentially  contaminated work  locations  that
are not continuously monitored.  These  samples shall be analyzed to ensure that
personnel are not exposed  to levels  of  eirborne  radioactivity  higher than  the
leve.ls given in 10 CFR 20,  Appendix  B,  Table I.  Sampling  devices  should  be
located where they will  ensure detection of  abnormal concentrations of airborne
radioactivity.  Examples of good sampling locations  include  on hood faces  and
above laboratory benches.'
5.1.5  The.Use of Protective Apparel
     The purpose of protective apparel  is to place  a  barrier between radioac-
tive material and the individual.  This  barrier  has  negligible shielding  char-
acteristics; that is, it does  not effectively  attenuate,  or reduce the intensity
of, the radiation reaching the wearer.   Its  main purpose  is  to prevent con-
tamination of the skin of personnel  and inhalation  of airborne radioactive
materials.  The two classes of protective apparel  discussed in this  section
are protective clothing, which minimizes the contamination of an individual's
skin, and respirators, which minimize the inhalation of eirborne radioactive
material.
     A.  Protective Clothing.   Protective clothing  includes gloves,  laboratory
coats, coveralls, and shoe covers.   All  protective  clothing for use  in radia-
tion areas should be clearly marked and essily identified so that  it can  be
kept separate from other clothing.
                                      5.7

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DARCOM-P 385-1
      Because protective clothing often becomes  contaminated,  it  must  be
 removed carefully so that contamination is  not  transferred  to  the  wearer's
 skin or street clothing.  In all cases, if  protective  clothing is  ripped  or
 torn while an individual is working with radioactive material,  the individual
 should leave the area immediately.   The following discussion  includes a  brief
 description of the proper methods for removing  the clothing.
      (1)  Gloves.  Gloves should always be  worn in radiation  areas, particu-
 larly for handling sealed and unsealed sources  or potentially  contaminated
 objects.  The best gloves are both strong enough not  to tear  and tight enough
 not to continually slip off or catch on experimental  apparatus.   Disposable
 surgical gloves are frequently used.  "One  size fits  all"  gloves tend to be
 large and to slip off the hands, and may promote the  spread of contamination
 because of the unconscious movements used to keep them on.   In some instances,
 for example during work with radioactive elemental iodine  or alpha-emitting
 radionucl ides, two pairs of gloves shou'io be worn.
      Glove removal can cause contamination  if not performed properly.  During
 the removal process, avoid quick movements  that may cause  dust to become
 airborne.  Touch the outside of gloves only with gloved hands, and touch
 uncontaminated skin only with ungloved hands.  Grasp the upper, inside wrist
 cuff of one glove with the opposite gloved hand and pull down on  the glove so
 that, as it is being removed, it is also being turned inside out.  When the
 first glove is off, it should be held, inside out, in the gloved  hand.  To
 remove the second glove, slide the fingers  of the ungloved hand down the inside
 of the gloved wrist until the fingers can grasp the inside cuff of the glove.
 Grasp the inside cuff with the bare fingers and pull  down on the  cuff while
 withdrawing the hand from the glove; this should cause the glove  to  be turned
 inside out.  Pull the second glove over the previously removed  glove.  The
 result should be two inside-out gloves, one inside the other, which  are dis-
 posed of as radioactive waste.  The wearer's hands should be  surveyed after
 the gloves are removed.
      (2)  Laboratory Coats.  Laboratory coats are required for  work  with radio-
 active materials.  The coats should be correctly sized for the  individuals
 wearing them and should be worn buttoned up.  They should be  worn only  in
                                       5.8

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                                                             DARCOhi-P  3E5-1
radiation areas and controlled areas, where contamination might" exist,  and
should never be worn in uncontrolled areas, where  food  or beverages  will  be
consumed.
      (3)  Coveralls.   In areas with a high likelihood of contamination  or
where loose laboratory coats would be inconvenient and  might  cause excessive
resuspension of radioactive materials, coveralls should be  used.   Coveralls
have  the relative advantage of protecting all  the  street clothing of an
individual.  They can be made of ordinary cloth, specie! fabrics, or chemically
treated papers.  Velcro  fasteners make it easier  to  remove coveralls.
      Coveralls are removed as follows.  First,  remove gloves  if they are being
worn.  Then insert the index finger and the middle finger of  each hand  inside
the front of the collar and loosen the Velcro  fasteners by  pulling the  hands
apart.  Slid-  the fingers down the front opening until  the  coveralls are open
below the waist.  Place the fingers inside the coveralls at about .the height of
the collarbone and pull the coveralls off the  shoulders and down until  the arms
are free.  Roll the coveralls, inside out, down the body to the ankles, then
step  out.
      (4)  Shoe Covers.  Shoe covers are required wherever  floors may become
contaminated.   They can be made of any durable material such as plastic,
fiber-embedded paper or cloth, or rubber.  Shoe covers  should be tight enough
so that they do not tend to fall off the worker's  shoes, but not so tight that
they are difficult to remove.  A step-off area or  pad for  removing shoe covers
should be located at the exit from the contaminated area.   To remove shoe
covers, approach but do not stand on the step-off  area.  Lift one foot so that
it crosses in front of the opposite leg, grasp the outside  of the cover at the
heel with a gloved hand, and pull it off the street shoe,  being careful to
maintain balance.   Do not remove the street shoe with the  shoe cover.  Place
the contaminated shoe cover in a receptacle, then  step onto the step-off  pad
with the street shoe.  Do not step on the pad  with the remaining contaminated
shoe cover.  Remove the remaining shoe cover as described  above.
®A trademark of Velcro U.S.A. Incorporated.
                                      5.9

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DARCOM-P 385-1
      If the protective gloves have already been removed,  shoe  covers can be
 removed by placing the index finger and middle finger of  the left hand between
 the right street shoe and its shoe cover at the inside of the  heel, pushing
 down until the heel of the street shoe is out of the shoe cover,  and then
 sliding the rest of the street shoe out of the shoe cover and  placing the right
 street shoe on the step-off pad.  Reverse the procedure for the left foot.
      (5)  Care of Contaminated Clothing.  Contaminated clothing should be placed
 in receptacles specifically designed for contaminated apparel, and  should  be
 sent only to laundries that are equipped to handle contaminated clothing.  If
 protective clothing is worn many times before laundering, it should be stored
 so that any contamination on it could not be transferred  to other items of
 apparel.  Protective clothing contaminated with more than 50 mrad/hr of beta-
 gamma  radiation or more than 40,000 dpm of alpha radiation shall  be considered
 contaminated waste and shall be removed from service.
      B.  Respirators.  Respirators are devices designed to keep the wearer
 from inhaling airborne radioactive material.  Some devices also protect against
 oxygen-deficient atmospheres.  They are not a substitute for either good ALARA
 (as low as is reasonably achievable) or good engineering practices.  Respira-
 tors are considered an acceptable method of protecting the health of personnel
 only under the following circumstances:
   1.  when the Ionizing Radiation Control Committee  (IRCC) has determined that
      no feasible engineering or work practice controls can be used to control
      the airborne radioactive material
   2.  during intermittent, nonroutine operations (1  hour/day for  1  day/week)
   3.  during interim periods when engineering controls  are being  designed and/or
      installed
   4.  during emergencies.
 Respiratory protection programs, the selection of  respirators, and  the  types  of
 respirators available are discussed below.
      (1)  Respiratory Protection Program.  An effective  respiratory  protection
 program requires the cooperation of the commander,  the RPO, supervisors, and
                                      5.10

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                                                             DARCOM-P  385-1
medical personnel.  An adequate program includes, at e minimum,  the  require-
ments detailed below.   Radiation Protectior  Officers who  ere  responsible  for
respiratory protection programs should obtain  a  copy of the Nuclear  Regulatory
Commission's NUREG-0041 (NRC 1976)  for further detail  concerning these
requi rements.
 1.  Air sampling and other surveys must be  sufficient to identify the  radia-
     tion hazard, to evaluate individual exposures, and to permit proper
     selection o.f respirators.
 2.  Written standing operating procedures  (SOPs) must be followed to ensure
     proper selection, supervision, and training of personnel  using
     respirators.
 3.  Written SOPs must be followed  to ensure adequate  individual fitting  of
     respirators, as well as procedures for  testing respirators  for operability
     immediately prior to each  use.  Individuals who  issue respirators  shall  be.
     provided with training in  these procedures.
 4.  Respirators should be assigned to individuals  for their  exclusive  use,
     where practicable.
 5.  Written SOPs must be followed  for respirator maintenance (including  clean-
     ing and disinfection), decontamination, inspection,  repair, and storage.
     Respirators issued for the exclusive use of one  indivudal should be
     cleaned after each day's use.   Respirators  used  by  more  than one individ-
     ual shall be thoroughly cleaned and disinfected  after each use.
 6.  Respirators shall be stored in a convenient and  sanitary location.  They
     must be stored where the potential for  contamination by  airborne or sur-
     face radioactive material  is minimal.
 7.  Before initial use, each respirator shall be properly fitted, leakage
     tests performed,  and the facepiece-to-face  seal  tested in  a realistic
     test situation.
 8.  Before each use,  both positive and negative pressure tests shell be con-
     ducted (see Standard Z88.2 of  the American  National  Standards Institute
     (ANSI 1980)).  Respirators shall not be worn when a beard  or sideburns, a
     skull cap that projects under  the respirator,  temple pieces on  corrective
                                     5.11

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OARCOM-P ,"'85-1
      glasses, the absence of one or both dentures, or other conditions prevent
      a good  facepiece-to-face seel.
   9.  Respirators shell be inspected during cleaning.  Experienced personnel
      shall replace worn or deteriorated parts with parts designed for the
      respirator.  No attempt shall" be made to replace components or to mal'e
      adjustments or repairs beyond the manufacturer's recommendations.
      Reducing-admission valves or  regulators shall be returned to the manu-
      facturer of to   trained technician for adjustment or repair.  Tne manu-
      facturer's parts replacement  schedule should be followed.
  10.  Respirators for emergency use, such as self-contained breathing devices,
      shall be thoroughly inspected at least once a month and after each use,
      and  a written record kept of  inspection dates and findings.
  11.  Supervisors and personnel shall be instructed and trained in the selec-
      tion, use, care, and maintenance of respirators.  Training shall provide,
      for  each user, an opportunity to handle the respirator, to have  it fitted
      properly, to test its facepiece-to-face seal, to wear it in normal air
      for  a familiarization period, and to wear  it in a realistic test
      atmosphere.
  12.  Personnel should not be assigned to tasks  that  require the use of respira-
      tors unles-s the installation's medical authorities have determined that
      they are physically and psychologically able to perform their work while
      wearing the prescribed respirator.  The medical status of the respirator
      user should be re iewed periodically, with the  frequency of  review depend-
      ing  upon the results of appropriate medical examinations, the type of
      respirator used, and the age  of the individual.
  13.  Bioassays and other surveys should be conducted as appropriate to evaluate
      individual exposures and to assess the protection actually  provided.
      (2)  Selection of a Respirator.  The selection  of a  respirator depends  on
  a number  of  health and safety factors, such £S  the  nature  of  the  radiation
  hazard, the  limitations and the intended use of the  respirator,  how much  the
  respirator limits movement and work rate, the  time  needed  to  escape  in case  of
  emergency, and training requirements.  Because  the  effectiveness  of  a respira-
  tory protection program can be determined largely by the  degree  to which

                                      5.12

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                                                              DARCOM-P
personnel accept the program, the human factor must  elso  be  considered.   Per-
sonnel acceptance of respirators is influenced by  comfort, ability  to breathe
without undue interference, confidence in facepiece  fit,  and  convincing
evidence that a respirator is necessary and that action is being  taken where
possible to eliminate the need for respirators.
     The degree of protection afforded by a given  respirator  is defined  in
terms of its protection factor (PF), which is  the  ratio of the concentr:tion
of the contaminant in the ambient atmosphere to that inside  the equipment
(usually inside the facepiece) under conditions of use.   Protection factors
are based on laboratory leakage studies and field  experience  with the device.
     Respirators should be selected to provide a PF  greater  than  the multiple
by which peak concentrations of radioactive materials are expected  to exceed
the values specified in Table I, Column I, of 10 CPR 20,  Appendix B. For
example, if the airborne concentration of a radionuclide  in  a work  area  is
expected to be five times as high as the permissible concentration  listed in
the table, then the respirator selected for use in that area should have a  PF
of 6 or more.  The equipment selected should be used so that the  average
concentration of radioactive material in the air inhaled  by  the  wearer,  during
any period of uninterrupted use in the area, does  not exceed the  values
specified in the table.  For the purpose of this manual,  the concentration  of
                                                          %
radioactive material inhaled when respirators are  worn may be estimated
initially by dividing the concentration in the air of the work area by  the  PF.
Additional measurements, however, must be taken to evaluate  worker exposure.
     The protection factors for respirators may not be appropriate where
chemical or other respiratory hazards exist in addition  to radiation hazards.
The selection and use of respirators for such circumstances  should take  into
account recommendations and requirements of the National  Institute for Occupa-
tional Safety and Health (NIOSH) and the Occupational Safety and Health  Admin-
istration (OSHA).
     The installation's medical authority, or personnel  under the  guidance of
the meoical  authority,  shall determine the type of respirator best suited to
each tesk.   The RPO should assist the responsible  individual by  providing
                                     c; 13
                                     W • A W

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DARCOM-P 385-1
 environmental evaluations and any other appropriate  information.   Only equip-
 ment that is certified by the National Institute  for Occupational  Safety and
 Health/Mine Safety and Health Administration (NIOSH/MSHA)  should  be used.
      (3)  Description of Respirators.  There are  basically two forms of
 respirators:  air purifying and air supplying.  An  air-purifying  respirator
 removes contaminants from the air of the work area  by either filtering out
 particulate contaminants or removing contaminated gases  and vapors by chemical
 mear-s.  An air-supplying (or atmosphere-supplying)  resoiretor furnishes respi-
 rabie air or oxygen to the wearer from an uncontaminated supply.
      Respirators are designed to be used with an  enclosure such as a facepiece,
 hood, helmet, or suit.  The enclosure excludes  contaminated air and ensures
 that clean, respirable air is supplied to the nostrils and mouth  of the wearer.
      A facepiece is a tight-fitting enclosure over all or a portion of the
 face.  Only full-facepiece devices should be used to protect against airborne
 radioactive material.  (Facepieces that enclose only a portion of the face are
 not acceptable for use in radiation areas; they are to be used only for indus-
 trial safety applications for protection from nonradioactive particulates,
 gases, and vapors.)  A full-facepiece mask is generally constructed from flex-
 ible rubber or plastic and has one or two transparent lenses for viewing.  The
 device completely encloses the wearer's eyes, nose, mouth, and chin.  A head
 harness is attached to the facepiece at five or six points to provide support.
      A hood is a loose-fitting, flexible enclosure over the head, neck, and
 shoulders that is gathered around the neck or shoulders to provide a snug fit.
 A helmet has a more rigid construction than a hood and protects parts of the
 head against impacts.  Air is supplied to the hood or helmet from a compressed-
 air :upply.. Suits are one-piece garments to which a coniinuous supply of
 respirable air is provided.
 5.1.6  Administrative Guidelines
      Some administrative, guidelines that will help personnel reduce any intake
 of radioactive materials are listed below.  The list may .not be all-inclusive
 and should not be substituted for common sense in the laboratory.
                                      5.14

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                                                        DARCOM-P 3E5-1
Smoking, eating,  and drinking shell  not  be allowed  in radiation  areas  or
controlled areas.   The danger of transmitting radionuclides  internally is
too great.
Food containers such as returnable  bottles and coffee cups shall  not be
taken into radiation areas or controlled areas.   If they are  inadver-
tently taken in,  they should be destroyed.
Refrigerators shall  not be used to  store both food  and  radioactive
materials.  Ice cubes from refrigerators used for storing  radioactive
materials shall not be used for human consumption.
Frequently while  working with radioactive materials,  or upon the comple-
tion of work, each individual shall  survey hands, shoes, and other  areas
of the body or clothing that may be contaminated.  Contamination should
be removed when found and shall be  removed before the individual leaves
the laboratory.  If significant levels of personnel contamination are
found, or if the  contamination cannot be readily removed,  the individual
shall contact the RPO.
Frequent radiation surveys shall be performed around radiation and/or
controlled areas  to determine whether there  is  any  deviation from normal
background levels of radiation (see Chapter  4).
All containers used for radioactive materials shall be labeled in accor-
dance with Army regulations (AR 385-11). Radioactive warning labels,
tape, signs, etc., shall not be used for purposes other than those for
which they are intended.
Radioactive materials shall be stored so that unauthorized individuals
are not likely to accidentally handle or otherwise  come in contact with
them.
Each person shall  wash hands and arms thoroughly after handling any radio-
active source (sealed or unsealed), and in  particular before touching any
object that goes  in the mouth, nose, or eyes.
Equipment or apparatus that has come in contact with radioactive materials
shall not be used for other purposes until  it  is demonstrated to be free
of contamination.
                                5.15

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JARCOM-P 385-1
 10.  Mechanical devices shall  be used  for  pipetting.   NEVER  PIPETTE  RADIOACTIVE
      SOLUTIONS BY MOUTH.   In addition,  to  preclude  accidental  ingestion  of
      radioactive materials through  cross-contamination,  mislabeling,  etc.,
      never pipette any substance by mouth  in  laboratories  where  radioactive
      materials are used.
 11.  Radioactive materials in liquid form  shall  be  stored  and  transported  in
      containers that, if dropped, will  not release  the materials,  for example,
      in plastic bottles or in glass bottles with styrofoam containers (see
      Chapter 9).
 12.  All transfers and dilutions should be performed  in  functioning  exhaust
      hoods or glove boxes, unless procedures  have been approved  for  working  in
      the ocen (see Chapter 8).
 13.  Work should be planned ahead;  whenever possible,  a  dry  run  to test  the
      procedure should be done first.
 14.  All items of equipment intended to provide features of  safety shall be
      evaluated periodically to ensure that they are providing  the  safety
      feature intended (see Chapter  8).   For example,  a fume  hood in  which
      radioactive materials are handled should provide  a  uniform  air  flow
      through the opening of the hood.   This air flow should  be checked
      periodically to ensure that the hood  is  operating properly.
 15.  Laboratories shall be kept neat and clean.  Equipment or  material not
      being used should be stored away from the work area.
 16.  Absorbent paper should be placed on work surfaces on which  radioactive
      materials are used.   If liquid radioactive materials are  used,  a container
      large enough to hold the entire volume of liquid  should be  positioned tc
    .  catch any spill.
 17.  Fingernails should be kent short and  clean.
 18.  If there is a break in the skin below the wrist,  gloves of  rubber,  plastic,
      or some other substance impervious to the material  being  worked with shall
      be worn to cover the break.
                                      5.16

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                                                              DARCCW-P 385-1
                   Section 5.2  MONITORING  INTERNAL  EXPOSURE

     Inhalation is the pathway by which radioactive  material  is  most likely  to
enter the body of an occupationally exposed individual.   After being inhaled,
the radioactive material  may be gradually or immediately  transferred to the
blood, depending upon zhe solubility of the material,  and then excreted from
or retained by the body,  depending upon other characteristics of the material.
     Two methods are used to estimate the amount  of  radioactive material  taken
into the body and the consequent radiation  dose:   radioanalysis and in-vivo
counting.  Radicanalysis  is the measurement of radioactivity  in urine, feces,
secretions, and other body samples, such as blood and  other tissues.  In-vivo
counting is the measurement of the radiation emitted from the body, using an
external detector.  Radioanalysis and in-vivo counting are bioessay procedures.
Because they are highly specialized techniques, assistance in carrying them
out should be sought from the Army Environmental  Hygiene  Agency.
5.2.1  Bioassay Programs
     Bioassay programs should be established whenever  there is a potential for
internal contamination.  Bioassays are appropriate for five purposes:  1) pre-
paratory evaluation, 2) exposure control, 3) diagnostic evaluation, 4) removal
of work restrictions, and 5) termination evaluation (ANSI N343-1978).
     A.  Preparatory Evaluation.  Bioassays should be  performed before an
individual begins work that could result in an internal exposure.  These evalu-
ations are performed to determine the nature and extent of any prior exposure
that could affect an individual's availability for job assignments.  Knowledge
of prior exposures is also helpful in distinguishing,  in later bioasseys, which
exposures are not attributable to the present working  environment.
     B.  Exposure Control^  Bioassays should be performed periodically to
ensure the adequacy of physical containment and contamination control meas-
ures.  Personnel should be evaluated often  enough so that unfavorable exposure
trends can be identified.  Bioassays may be required more frequentl-y whenever
new processes, procedures, controls, or equipment are  put into  use, to verify
that protective measures  are adequate.  An  increased frequency  is  also required
whenever surface or air contamination is detected.
                                     5.17

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DARCOM-P 365-1
      C.  Diagnostic Evaluation.   Bioassays  are  used  after  a  known  intake  of
 radioactive material to determine the location  and  amount  of the deposition;
 to provide data necessary for estimating internal dose  rates,  the  fraction  of
 the deposition retained in the body,  and dose commitments;  and to  determine
 the necessity of work restrictions or referrals for  therapy.
      D.  Reiroval of Work Restrictions.   If  an individual's  internal  dose  rate
 has approached or exceeded applicable limits  and the individual's  work in radi-
 ation areas has been restricted, bioassays  should be performed to  determine
 whether the dose rate has decreased enough  so that  the  work  restrictions  can
 be lifted.
      E.  Termination of Employment.  Bioassays  should be performed as a regular
 part of the formal termination sequence in  order to  determine the  level of
 internal exposure attributable to the individual's'job  function.
 5.2.2  Actions To Be Taken Upon Detection of  an Intake
      If a routine bioassay performed  to assess  control  indicates  an abnormal
 (i.e., unexpected) presence of a radionuclide in the body or excreta, further
 evaluations should be made to confirm that  an intake has actually  occurred.
 (False indication of an intake may result from contaminated skin  in the case
 of in-vivo counting, or from contaminated samples in the case of  radioanalysis
 of excreta.)  The individual should be surveyed for external contamination,
 procedures for external decontamination should  be used (see Chapter 7), and
 then another in-vivo measurement should be  made.  If the measured  activity
 decreases, the contamination is probably external.   Continue decontamination
 procedures until two consecutive measurements result in no significant change.
 If the measured activity remains constant and an intake cannot be  ruled out,
 then radioanalysis of excreta should  follow.
      The interpretation of in-vivo counting data is influenced by a number of
 variables.  Examples of equations that can  be used  to calculate internal  dose
 are provided in Section 5.3.  However, the  interpretation of bioassay data
 requires trained personnel.  The RPO should contact the Army Environmental
 Hygiene Agency for assistance.  The radioanalysis of excre.ta and other body
 samples is also performed by the Army Environmental  Hygiene Agency.   If
 activity is found in excreta .^mples, the agency can provide assistance  in
                                      5.18

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                                                             DARCOM-P 385-1
interpreting the  data  in  light  of. Pub!ications  10 and  10A  (1968,  1971) of  the
Internetu>",al  Commission  on  Radiological  Protection  (ICRP).
     The cause of a  confirmed  intake  should  be  investigated, especially  if the
contamination  occurs  in several  persons  or recurs from time to  time  in one
person.   The dose reduction  methods discussed  in Report No. 65  of the National
Council  on Radiation  Protection  and Measurements (NCFP 1980) should  be con-
sidered  for use under the supervision of medical personnel.
                 Section  5.3   INTERNAL  DOSIMETRY  CALCULATIONS

     Dosimetry is the  measurement of the  radiation  absorbed  by  an  object.
Calculations of internal  dosimetry,  or  the  radiation  absorbed by the body's
organs and tissues,  serve two purposes:   1)  to determine  the amount of
radioactive material  that can be inhaled  in  eir or  ingested  in  water by an
individual without a  radiation dose  limit being exceeded;  and 2) to estimate
the radiation dose an  individual will  receive from  radioactive  material that
has already entered  the body.  In the first  case, the calculations are used
for preventive purposes,  to limit the dose  that might be  received  by setting
limits for the uptake  of  radioactive material; in the second case, the
calculations are used  for diagnostic purposes, to determine  the dose that will
actually be received.   The two uses  of  internal dosimetry calculations will  be
discussed separately.
5.3.1  Calculation of  Acceptable Intake
     Most federal regulations concerning  safe concentrations of radionuclides
in air or water are  based on  the recommendations of the ICRP in its Publica-
tion 2 (1979), Report  of  Committee II on  Permissible Dose for  Internal Radia-
tion.  However, ICRP has  recently issued  revised recommendations  in ICRP
Publication 30, and  these recommendations are being considered  for  incorpora-
tion into the Environmental Protection  Agency's "Federal  Radiation  Protection
Guidance for Occupational Exposures" yreciercl Register, January 23, 1981).
     The major difference between the two ICRP publications  lies  in the sophis-
tication of the dose  calculations used.  In ICRP 30, mathematical descriptions
                                     5.19

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DARCOM-P 385-1

 of organ shapes are used, whereas in ICRP 2, organs  of  a  rather  nebulous  shape
 are assumed.  The limits in ICRP 30 also account for tne  radiation dose to an
 organ from radioactive material situated in an unrelated  organ,  and ICRP  30
 uses a more complex model of radionuclide distribution  kinetics  (that is, the
 rate of radiation's absorption into the body, distribution  within the body,
 and eventual excretion from the body) than does ICRP 2.
      Because current regulations are based on the earlier ICRP publication,
 the material in the text of this section relates to  ICRP  2.   The terms used in
 ICRP 30 and the equations developed there for calculating the radiation dose
 to various body organs are discussed in Appendix A.
      The ICRP 2 methodology for calculating acceptable  intakes of radionuclides
 in air or water involves three steps:
  1.  determining the critical organ; that is, determining which organ or tissue
      of the body would be most damaged by a given radionuclide entering the
      body
  2.  calculating the maximum permissible body burden; that is, calculating the
      maximum amount of the radionuclide that can enter the body without the
      maximum acceptable dose limit for the critical  organ being exceeded
  3.  calculating maximum permissible concentrations; that is, calculating  how
      much of the radionuclide can be in air that is breathed or water that is
      drunk without the maximum permissible body burden being exceeded.
 These steps are explained below.
      A.  Determining the Critical Organ.  The critical  organ or critical  tissue
 is the organ or tissue that, if damaged by radioactive material taken  into the
 body, would cause the greatest physiological damage to the body.   In concept,
 the critical organ or tissue for a given radionuclide is determined  by con-
 sidering:  1) which organ accumulates the greatest concentration  of  the  radio-
 nuclide; 2) the importance of each organ to the well-being of the  entire  body;
 3) which organs are most affected by the route of entry of the  radionuclide
 into the body (e.g., the lungs are most affected by the inhalation of  a  radio-
 nuclide); and 4) the radiosensitivity of each organ, that is, which  organ  is
 damaged by the lowest dose.  In practice, the first criterion (the organ  that
                                      5.20

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                                                             DARCOM-P  365-1
hes the greatest concentration of a given radionuclide)  is  used in ICRP 2  to
determine the critical organ because of the difficulty  of evaluating the other
criteria.  If the radionuclide is not concentrated  in any single organ, then
the whole body is considered to be the critical  organ.
     B.  Calculating the Maximum Permissible Body Burden.   The  maximum  permis-
sible body burden (MPBB) is the amount of a radionuclide, accumulated throughout
the body of an individual over 50 years of occupational  exposure, that  will
result in a maximum permissible dose-equivalent  rate to the critical organ for
that radionuclide.  (See Chapter 3, Table 3.2, for  maximum  permissible  dose-
equivalent rates.)
     C.  Calculating the Maximum Permissible Concentrations in  Air and  Water.
The MPBB must be considered in order to estimate the acceptable concentrations
of a radionuclide in air or water.  In ICRP 2, a maximum permissible concen-
tration for air, (MPC) , and a maximum permissible  concentration for water,
____               g        ____^^_^__^_^_____^_^_^___^___
(MPC) . are given.  The (MPC)  and (MPC)  are calculated based  on a constant
     W                       c          W
intake of a radionuclide into the body and an exponential elimination of the
radionuclide from the body by radioactive decay  and biological  excretion.   The
calculations account for the breathing rate of the  individual in the case of
(MPC), and for the amount of water the individual might consume during  the
     a
day in the case of (MPC) .  The fraction of the  material actually retained
                        W
in the body is also considered.  The ICRP 2 recommendations for (MPC)  and
(MPC)  limits have been incorporated into the permissible concentrations of
     W
radionuclides in air and water that are listed in 10 CFR 20, Appendix B.  The
MPC is given in uCi/ml.
5.3.2  Estimation of Internal Dose
     Following the ingestion or inhalation of radioactive material, three dose
computations can be made:  1) the initial dose-equivalent rate, which  is impor-
tant because it serves as the basis for calculating the total dose  received;
2) the dose equivalent the critical organ or the total  body will  receive over
1 year; end 3) the total  dose equivalent the critical  organ or  the  total body
will  rece-'ve as a result of the ingestion.  The  total  dose  equivalent  car. be
calculated either for an infinite time following the ingestion  or for  50 years
following the ingestion.   A calculation based on the 50-year period results in
                                     5.21

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DARCOM-P 385-1
 what is called the 50-year dose commitment..   The methods of dose calculation
 described in this section are for an individual  of standard size and average
 metabolism (which effects the rate of excretion  of the radioactive material).
 If the calculations are to be modified to fit a  particular individual,  the
 Army Environmental Hygiene Agency should be  contacted for assistance.
      Equations provided in ANSI Standard N343-1S78 can be used for calculating
 the initial dose-equivalent rate and the 1-year  and 50-year dose equivalents
 resulting from an intake of radioactive material.  In all cases, it is  neces-
 sary to know the amount of radioactive material  in the body or in the organ
 fc1- which the dose is being calculated.  The calculations would be based on a
 single in-vivo measurement (i.e., a measurement  of the radiation emitted from
 the body, made using an external detector soon after the intake).
      A.  Calculating the Initial Dose-Equivalent Rate to an Organ.

                     51.2 x q(t) X f, x e
                 H -- 5 - 1 -                              (5.1)

 where           H = the dose-equivalent rate to the organ  (rem/day)
              q(t) = the activity in the whole body at the time of measurement
                fo = the fraction of the total-body radioactivity in the organ
                     of reference, from ICRP 2^
                 e = the effective absorbed energy per disintegration (MeV/dis)
                 m = the mass of the organ of reference (g)
              51.2 = constant ([renrg'dis]/[uCi -MeV-day]) .
      B.  Calculating the One-Year Dose Commitment Base' on a Single In-Vivo
 Measurement.  Equation (5.1) allows the calculation of the dose-equivalent
 rate to an organ containing radioactive material.  One may be more interested
 in the total dose an individual will receive for a year and/or a lifetime
 following a deposition.  Equation (5.2) allows for the calculation of the
 1-year dose equivalent to an organ containing radioactive material.
 (a)  If the amount of radioactive material actually  in the organ of  interest
      is known, then that activity, in units of microcuries, may be used  in  the
      equation rather than the product f  q(t).
                                      5.22

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                                                             DARCO.N-P  385-1

                    51.2 x q(t) x f, x c x eAt  [1  -  e"365X]
               V	STl	          (5"2)

where          H  = the 1-year dose equivalent based  on  a  single in-vivo
                    measurement (rem)
             q(t) = the activity in the whole body  at the  time  of measurement
                    (yCi)
               f? = tne fraction of the total-body  radioactivity in the organ
                    of reference
                c = the effective absorbed energy per disintegration (MeV/dis)
                e = the base of the natural  logarithms (e  = 2.71828)
                X = the effective removal  constant  (X =  0.693/t ff) (days"1)
                t = the time between the intake and' the  in-vivo measurement
                    (days)
                m = the mass of the organ  (g)
             51.2 = constant ([rem-g-disj/tviCi'MeV-day]).
     C.  Calculating the Fifty-Year Dose Commitment.   The  50-year dose equiva-
lent can be calculated by modifying the exponent (-365X) in the above equation
to (-18250X), which corresponds to a 50-year time interval.
     Values of f.,, X,  and E for a few selected radionuclides are given in
Table 5.1.  The parameters f2 and X listed in this  table are based on a "stan-
dard man," defined in  the Radiological Health Handbook (1970) as having a body
weight of 70 kg.  The  use of these values  in an equation will provide an esti-
mate of the radiation  dose to an individual  who is  the same size as the
standard man.  If possible, bioassay procedures should be used to obtain esti-
mates of ^2 and * that more closely match  the individual.
     Another source of reference for calculating the 50-year dose commitment
is NUREG-0172 (NRC 1977).  This report lists 50-year committed radiation doses
to selected organs following the chronic intake of several radionuclides over
a 1-year period.  The  radiation doses are  calculated in  terms of mrem per
                     1 ?
50 years per pCi (1C"1' Ci) of radioactive material.   The dose Calculations
are for populations rather than occupetionally exposed individuals and include
1) radiation doses from liquid effluents,  2) radiation doses from gaseous
                                     5.23

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DARCOM-P 385-i
                  TABLE 5.1.  Parameters  for Internal Dosimetry^
                                Organ Mass (grans)  for  Standard Mar



Nuclide
3H
54Mn

59Fe

58Co

60Co

95Zr-Nb

95Nb

106Ru-Rh


131,
133,
134Cs

137Cs.137mB

140Ba-La


144Ce-Pr



To^al Boay
Lung
Thyroid
Oroan
Total Body
Lung
1 Liver
Lung
Spleen
Lung
Total Body
Lung
Total Body
Lung
Total Body
Lung
Total Body
Lung
Kidney
Total Body
Thyroid
Thyroid
Lung
Total Body
a Lung
Total Body
Lung
Bone
Total Body
Lung
Bone
Liver
Total Body
70,000
1,000
20
f2
1.0
(d)
1.0
(d)
0.02
(d)
1.0
(d)
1.0
(d)
1.0
(d)
1.0
(d)
0.07
1.0
0.2
0.2
(d)
1.0
(d)
1.0
(d)
0.7
1.0
(d)
0.38
0.19
1.0
Liver
Spleen
Bone
i(b)
3.2 (-4)
8.1 (-3)
3.0 (-2)
2.1 (-2)
1.7 (-2)
1.5 (-2)
8.3 (-2)
6.1 (-3) .
7.3 (-2)
1.7 (-2)
1.2 (-2)
2.6 (-2)
2.1 (-2)
7.7 (-3)
2.8 (-1)
9.6 (-2)
9.6 (-2)
8.0 (-1)
9.5 (-3)
1.1 (-2)
5.8 (-3)
9.9 (-3)
6.0 (-2)
6.5 (-2)
6.5 (-2)
8.2 (-3)
2.9 (-3)
4.7 (-3)
3.6 (-3)



c(-)
0.01
0.23
0.23
0.42
0.34
0.29
0.61
0.72
1.5
0.52
1.1
0.26
0.51
1.4
1.3
1.4
0.23
0.54
0.57
1.1
0.41
0.59
1.4
4.2
2.3
1.3
6.3
1.3
1.3
1,700
150
7,000
[l-exD(-365*n
o.::
0.95
1.0
1.0
1.0
1.0
1.0
0.89
1.0
1.0
0.99
1.0
1.0

1.0
1.0
1.0
1.0
0.97
0.98
0.88
0.97
1.0
1.0
1.0
0.95
0.65
0.82
0.73
        (a) American National  Standards Institute 1978.
        (b) Units of day"'.
        (c) Units of (MeV/disintegration) x (rem/rjd).
        (d) Estimates of lung  dose should be based or, a  measured  lung burden.  However,
            a total-body in-vivo measurement can be used to  estimate an upper limit of
            the lung dose commitment by setting f- = 1.0 for the  lung.
                                            5.24

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                                                             DARCOM-P 365-1

effluents, and 3)  radiation  doses  from contaminated  surfaces  or  volumes  (i.e.,

external  radiation).

     The intake of the same  amount of  radioactivity  can  result in  different

radiation doses 'or neople of different ages;  consequently,  four sets  of dose

factors are presented in NUREG-0172.   The age  groups considered  are  infant,

child, teen, and adult.   The 50-year  dose commitment is  calculated by  reading

the dose factor from the approriate table and  multiplying  this value by  the

number of picocunes taken into the body.  The tables of NUREG-0172  are  not

reproduced in tnis manual.
                                  REFERENCES


American National Standards Institute (ANSI).  1978.   Internal  Dosimetry for
  Mixed Fission and Activation Products.  ANSI N343-1978.  Wasnington, D.C.

American Notional Standards Institute (ANSI).  1980.   Practices for Respiratory
  Protection.  ANSI Z88.2, New York.

International Commission on Radiological Protection (ICRP).  1959.  Report of
  Committee II on Permissible Dose for Internal Radiation.  ICRP 2, Pergamon
  Press, Oxford.

International Commission on Radiological Protection (ICRP).  1968.  Evaluation
  of Radiation Doses to Body Tissues  From Internal Contamination Due to
  Occupational Exposure.ICRP 10, Pergamon Press, Oxford.

International Commission on Radiological Protection (ICRP).  1971.  The Assess-
  ment of Internal Contaminetion Resulting from Recurrent or Prolonged Uptakes.
  ICRP 10A, Pergamon Press, Oxford.

International Commission on Radiological Protection (ICRP).  1978.  Limits for
  Intakes of Radionuclides by Workers.  ICRP 30, Part 1 and Supplement to
  Part 1, Pergemon Press, Oxford.

International Commission on Radiological Protection (ICRP).  1980.  Limits for
  Intakes of Radionuclides by Workers.  ICRP 30, Part 2,  Pergamon  Press,
  Oxford.

Nation?! Cc-jncil on Radiation Protection and Measurements  (NCRP).   1980.
  Management of Persons Accidentally Contaminated with Radionuclides.  NCRP  65,
  Washington, D.C.
                                     5.25

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DARCOM-P 385-1

 Radiological Health Handbook.  1970.   U.S.  Department  of Health,  Education,  and
   Welfare, Washington, D.C.

 U.S. Code of Federal Regulations.  1982.  Title 10,  Part 20,  "Standards for
   Protection Against Radiation."  U.S.  Government Printing  Office,  Washington,
   D.C.

 U.S. Department of the Army, Headquarters.   Safety - Ionizing Radiation
   Protection (Licensing, Control, Transportation. Disposal,  and Radiation
   Safety).AR 385-11,"Washington, D.C.

 U.S. Environmental Protection Agency (EPA).   "Federal  Radiation Protection
   Guidance for Occupational Exposures."  Federal  Register,  January 23, 1981.

 U.S. Nuclear Regulatory Commission (NRC).  1976.   Manual of Respiratory
   Protection Against Airborne Radioactive Materials.  NUREG-0041, Washington,
   D.C.

 U.S. Nuclear Regulatory Commission (NRC).  1977.   Age-Specific Radiation Dose
   Commitment Factors for a One Year Chronic Intake.   NUREG-0172,  Wasnington,
   D.C.
                                      5.26

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                                                             DARCOM-P  385-1
                                  APPENDIX A

             :RP 30 RECOMMENDATION'S FOR LIMITING  RADIONUCLIDE  INTAKES
     The most recent recommendations of the  ICRP  for  safe  limits  of  radioactive
material in air and water are found in Publication  30,  Limits  for Intakes  of
Radionuclides by Workers.  To date, ICRP 30  consists  of two  parts published  in
1978 and 1980, each with a supplement.  A third  part  and supplement  are expected
to be published.  Because ICRP 30 is so recent,  its recommendations  have not
been incorporated into current government regulations;  however,  they may be
incorporated into future regulations.

A.I  EXPLANATION OF TERMS USED IN ICRP 30
     The sequence of steps used in ICRP 30 to determine acceptable concentra-
tions of radionuclides in air or water is identical to  that  used in  ICRP 2 and
discussed in Section 5.3.  The terminology used  in  ICRP 30 is  different from
that used in ICRP 2, however, and is explained below.
     A.  Committed Dose Equivalent.  In ICRP 30,  the  Commission is attempting
to limit two types of radiation effects in the body:   1) stochastic  effects  are
those that are increasingly likely to occur  as the  radiation dose increases
(for examnle, genetic effects and malignant  diseases  such as cancer); 2) non-
stochestic effects are those that are .increasingly  severe as the radiation dose
increases and thet are unlikely to occur at  all  below a certain threshold dose
(fo- example, loss of heir, skin damage, and cataracts).
     The incidence of stochastic effects is  limited if  the risk of such effects
resulting from the radiation dose to any single  organ or combination of organs
in 1 year does not exceed the risk associated with  a  whole-body dose equivalent
of 5 rem in any 1 year.  The risk of stochastic  .effects is quantified by a
weighting factor for each organ; the weighting factor is an attempt to scale
both the relative importance of the .v can to the we'1-being of the body, and
the organ's relative radiosensitivity.  The  weighting factors can be used to
obtain a dose equivalent, HL, to a tissue that yields the same risk as 5 rem
                                      5.27

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DARCOM-P 385-1
 to the whole body.  The committed dose equivalent (H, 5n) in a tissue is the
 total  radiation dose equivalent received by an organ or tissue during the 50
 years  following an  intake.  The maximum intake of a radionuclide is limited in
 ICRP 30 by  the requirement that the sum of the ratios of H^ 5Q/HL in all
 irradiated  tissues  not exceed 1.0.  It is not possible to directly compare the
 doses  to  the critical organs given in ICRP 2 with the annual doses to the
 critical  organs given in  ICRP 30  (see Table 5.2 below).  This is because
 ICRP 30 restricts the sum of the  doses received by all the tissues of the body,
 whereas ICRP 2 restricts  the dose tc the critical organ only.

              TABLE  5.2.   Weighting Factors Recommended in ICRP 30

                                                          H     (rem)
               Organ or Tissue         Weiahtina Factor    L
           Gonads    .                        0.25               20
           Breasts                           0.15               33
           Red  bone marrow,  lung             0.12               42
           Thyroid, bone  surfaces          '  0.03              167
           Five other  tissues  receiv-        0.30               83
           ing  the greatest  dose in
           the  remainder  of  the body
           (a)  Dose  equivalent  to  a  tissue  giving  the  same  risk  as
               5  rem to  the  whole  body.

       In  order to prevent nonstochastic  effects,  ICRP- 30 limits the radiation
  dose  equivalent to any organ  over  the 50  years following  an intake (the
  committed  dose  equivalent)  to 50 rem.
       B'.  Annual Limit  of Intake.   In  ICRP 30, the  MPBB of ICRP 2  has been
  replaced by the annual  limit  of  intake  (ALI).  The ALI is the  amount of a
  radionuclide  that  can  be ingested  or  inhaled  such  that the sum of the ratios
  HT  CQ/H.  in all the tissues irradiated  is equal  to 1.  In addition, the
  committed  dose  equivalent  to  any organ  cannot exceed 50 rem in 1  year.
                                                                                    4
                                       5.28

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                                                             DARCOM-P  385-1
     The ALI is calculated based on a constant inhalation or ingestion over
the year.  Also considered is the rate at which radioactive material is elimi-
nated  from the body by both radioactive decay anc excretion.   The intake rate
can be exceeded at times as long as the total  yearly intake does not exceed
the specified ALI.
     C.  Derived Air Concentration.  The MPCs  given in ICRP 2  have been
replaced in ICRP 30 by a derived air concentration (DAC), which is the accept-
able concentration of a radionuclide in air.   The ICRP 30 recommendations are
                       3
listed in units of Bq/m , which can be converted to pCi/ml  by  multiplying by
the conversion factor 2.7 x 10"   (pCi-m )/(ml-Bq).  No derived water concen-
tration is defined in ICRP 30, nor is any value given that would be equivalent
to the MPCs.  The only mention made of a maximum concentration allowable in
air and water is that the total intake should  be less than the ALI.

A.2  DEVELOPMENT OF EQUATIONS USED IN ICRP 30
     A major change in ICRP 30 as compared to  ICRP 2 is that the radiation
dose to an organ is determined taking into account the radioactive material in
other organs as well  as in the organ of concern.  This change  is especially
important for intakes of radionuclides that emit gamma rays, x reys, or neu-
trons by spontaneous  fission.
     The committed dose equivalent to an organ (Hj ro) is a product of the
committed absorbed dose (D7 50), the quality factor of the radiation (Q),
and other modifying factors (N).  For the time being, ICRP has stated that N
is equal  to 1.   In the following paragraphs,  the equations used in ICRP 30 for
calculating the interne! dose ere developed.
     A.  Radiation Energy. £.  The dose equivalent to an organ is related to,
or proportional  to (symbolized = ), the energy  of the radiation.  In the case
of alpha  particles and gamma reys, E is the energy of the radiation listed on
periodic  tables  end in reference books.  In the case of beta particles, an
average energy  of the radiation must be calculated because beta particles are
emitted from the nucleus with a spectrum of energies.  As a general rule, the
                                     5.29

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DARCOM-P 385-1
  average  beta  energy  is about one-third of the listed or maximum enercy.  A
  more  exact  equation  is:
                  E. 0.33x^1-  ^  ,^1+  _»^,EMX          (5.3)

  where            E  = average  beta energy  (MeV)
                  ,2  = atomic number  of  the emitting nucleus
               E     = maximum  beta energy  (MeV)
               0.33  = constant.
  For positrons,  the equation  is:
                  E=0.33xl+          x   E                           (5.4)
  where           E  =  average  positron  energy  (MeV)
               E     =  maximum  positron  energy  (MeV)
               lUGA
               0.33  =  constant.
  The relation  of  the  committed  dose  equivalent  to the  radiation  energy  is:

             HT,50  «  E                                                  <5'5>

  where      Hj  rg  =  committed  dose  equivalent  to a  target  organ
                  E  =  energy of  the  radiation  (MeV).
       B.   Type of Radiation Emitted.   Each  type of  radiation  has a  character-
  istic rate  of energy deposition, or linear energy  transfer (LET),  as described
  in  Chapter  1. The quality factor,  Q,  is a function of  the radiation's LET and
  is  included in  the calculation of  the  dose equivalent.
       The  relation  can  now be written  a-s follows:

             HT>50  -  E  x  Q                                              (5.6)

  where      Kj  ^  =  committed  dose  equivalent  to a  target  organ
                  E  =  energy of  the  radiation  (MeV)
                  Q  =  quality  factor  of  the  radiation.
                                       5.30

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                                                              DARCQM-P 385-1
     C.  Mass of the Organ, IT*,.  The radiation dose received by an organ is
inversely proportional to the mass of the organ.   Because the absorbed oose is
defined in terms of energy absorbed per unit mass, if the amount of energy
deposited remains constant, then the absorbed dose necessarily decreases as
the mass of the organ increases.
     The relation for the committed dose equivalent is therefore:

            K     -iJLS                                             (5.7)
             T,bu    rr.y

where       KT 50 = committed dose equivalent to  a target organ
                E = energy of the radiation (MeV)
                Q = quality factor cf the radiation
               m- = mass of the target organ (g).
     D.  Absorbed Fraction of the Emitted Energy, AF(T-^S).  A fraction of the
energy emitted by radioactive material is absorbed in the organ containing the
material, and the remainder escapes.  The energy  that escapes from the organ
may penentrate through the body and produce a radiation dose in another organ,
or it may escape from the body.  The fraction of  the emitted energy absorbed
in a given organ is symbolized by AF(T«-S); T represents the target organ (the
organ receiving the dose), and S represents the source organ (the organ con-
taining the radioactive material).  The target organ and the source organ may
be the same organ, or they may be different organs of the body.  As a result,
it is now possible to calculate the radiation dose to an organ resulting from
radioactive materiel in a different organ.
     For the calculation of the absorbed fraction, radiations can be  placed
into Lwo categories:  nonpenetrating radiation and penetrating radiation.
     Nonpenetretino radiation is radiation that loses all of its energy after
traveling a short distance in tissue.  Examples of nonpenetrating radiation
are alpha particles, beie particles, and protons.  If the organ containing the
radioactive material is large compared to this distance, all the energy emit-
ted is deposited in the organ containing the radioactive material.  Thet is:
                                     5.31

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DARCOK-P 385-1
                            0,  if T is  not  S
                AF(T-S)  =
'5.8'
                            1,  if  T is  S
      Penetratinc radiation is radiation that  penetrates  through  the  body,
 depositing energy both in the organ that contains  the  radioactive  materiel  and
 in other organs.  Examples of penetrating radiation  incluoe  x  rays,  gamma  rays,
 and neutrons. The calculation of AF(T^-S) for  penetrating radiation is  quite
 complex and virtually impossible without the  aid  of  a  computer.   The computer
 is first programmed with a mathematical description :of a man of  average size,
 termed the reference man or standard man.  This mathematical  description is
 called a phantom and describes the shape, density, and relative  locations  of
 the various bones and organs of the body.  The absorbed fraction is  then cal-
 culated using a "Monte Carlo" computer calculation!   A description of  the
 basic principles behind these calculations follows.   The "Monte  Carlo" calcu-
 lations, although equivalent to this description,  are different  in detail  to
 save computer time.
      The radioactive nuclei are assumed to be distributed uniformly throughout
 the source organ.  A point within the source  organ is picked.   The computer
 model emits a photon of energy E in some direction picked at random from all
 possible directions.  The photon is followed  along its path; after it has  tra-
 versed a very short distance, the probability of  its interacting is calculated.
 The computer then "flips a coin" with this probability.  If a "head" results
 from the coin flip, the photon is considered  to interact at that point.  If
 the interaction is Compton scattering (see Chapter 1), the angle is picked at
 random with a relative probability determined by  the energy of the photon end
 by the interacting medium.  The energy of a recoil electron for scattering at
 that angle is calculated and deposited at the interaction site.   Similar pro-
 cedures are followed for the photoelectric effect and pair production.  The
 scattered photon is then followed in the same way.  If a "tail" occurs on the
 first coin flip, the photon is allowed to travel  another small distance and
 the probability of interaction is again calculated.   This procedure is
 repeated until  all the energy has been absorbed or the radiation leaves the
                                      5.32

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                                                             DARCOM-P  385-1
body.  The entire procedure -is repeated many times for each organ,  until  one
has a map of the radiation deposited in all  organs by gamma rays  (or other
penetrating radiation) leaving the specified point in the source  organ.  The
result of these calculations is the AF(T^S)  for penetrating radiation.   These
values are tabulated in ICRP 30 and its supplements.
     The relation for the committed dose equivalent is now written  as:
            HT ,n . E » 0 « [AF(T-51]                                 (5.9)
             T,50          ITU-

where       H- rr, = committed dose equivalent to an organ
                E = energy of the radiation (MeV)
          AF(T-t-S) = absorbed fraction of the emitted energy
               m,. = mass of the target organ (g)
                Q = quality factor of the radiation.
     E.  Radiation Yield, Y.  A radionuclide can undergo decay by different
pathways.  In the case of a beta-emitting nuclide, all pathways are similar in
that they entail  the emission of a beta particle followed by a gamma ray, but
they differ from each other in the distribution of energy between the beta
particle and the gamma ray.  The radiation yield, Y, is the fraction of dis-
integrations that yield a certain radiation type and energy.
     The committed dose-equivalent relation can now be written as:
            h*     . V x E x Q x [AF(T-S)1                   '          ,RIQX
            hT,50 c - 1~ -                             ^'1U)

where       H,. eg = committed dose equivalent to an organ
                Y = radiation yield (no units)
                E = energy of the radiation (MeV)
          AF(T*S) = absorbed fraction of the emitted energy
               m- = mass of the target organ (g)
                Q = quality factor of the radiation.
     The expression on the right side of Equation  (5.10) is collectively
referred to as the specific effective energy [SEE(T-S)].  "his indicates the
energy, in units  of MeV, deposited per gram of the target organ for each
                                     5.33

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DARCOM-P 3S5-1
 disintegration.  Because the radioactive material  may emit more tnan one type
 of radiation,  it is necessary to sum the contributions from all radiations
 emitted; that  is:

     SEE(7^S)total =  £  [SEE(T-S)]i                                   (5.11)

 where
     SEE(T-S)     , = specific effective energy of the nuclide (MeV/[g-Bq]),
              totai
                     which  is unique for any given combination of nuclide,
                     source organ, and target organ

    £  [SEE(T~S)]. = [
    i = l
  Thus, we  can write the relation for the committed dose equivalent as

             HT  50 «  SEE(T^S)                                            (5.12)

  where       HT  r^ =  committed dose equivalent to an organ
               I , OU
           SEE(T-S) =  specific effective energy of the radioactive nuclide
                      per  disintegration (MeV/g-dis).
       F.   Total  Number of Disintegrations  in the Source Organ, Uc.   The  total
  number  of disintegrations  in an organ over the 50 years following a single
  uptake  of radioactive material  is a complicated function of the  physical  decay
  of the  radionuclide  and  the metabolic characteristics of the  chemical  compound
  that  contains  the radionuclide.  For example, radioactive material  may be
  biologically eliminated  from one organ, perhaps the lung, only to be absorbed
  by a  second orgar,, such  as  the  liver.  The equations describing  the time-
  dependent distribution of the radioactive material car be found  in  ICRP 30  and
  are not discussed here.
                                       5.34

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                                                              DAKCOM-P 385-1
     If the 50-year cumulated activity in the source organ,  U<., is given in
disintegrations, then the relation for the committed dose equivalent may be
written as:
           HT 5Q  « U$ x [SEE(T*S)]                                   (5.13)

where       H. 50 = committed dose equivalent to an organ
               U<. = number of transformations in the source  organ S over
                    50 years following the intake of a radionuclide.
     G.  Conversion Factors.  Finally, the calculation of appropriate conver-
sion factors allows the replacement of the proportionality symbol by an equal
sign.  The conversion factors convert the energy deposition  to rem for the
traditional system, or to sievert if the SI system is to be  used.  In units of
rem, the appropriate equation is:

            HT 50 = (1.6 x 10"8) x U$ x [SEE(T-S)]                    (5.14)

where       H, CQ = committed dose equivalent to a target organ (rem)
               Ur = number of transformations in the source  organ S over
                    50 years following the intake of a radionuclide
         SEE(T*S) = specific effective energy of the radionuclide
                    (MeV/g).
In units of sievert, the equation is:

            HT 5Q = (1.6 x 10"10) x Us x [SEE(T*S)]                   (5.15)

where       HT CQ = committed dose equivalent to a target organ (Sv)
               U^ = number of transformations in the source  organ S over
                    50 years following the intake of a radionuclide
         SEE(T*S) = specific effective energy of the radionuclide
                    (MeV/g).
                                    5.35

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                                                              DARCOM-P 385-1
                         CHAPTER 6.  EXTERNAL EXPOSURE

6.1  CONTROL AND REDUCTION OF EXTERNAL RADIATION DOSE  .     .     .     .     6.3
     6.1.1  Exposure Time     .........     6.4
            A.  Basic Principle    ........     6.4
            B.  Control of Time    ........     6.4
            C.  Reduction of Time  ........     6.5
     6.1.2  Distance from the Source    .     .     .     .     .     .     .     6.6
            A.  Basic Principle    ........     6.6
            B.  Control of Distance     .......     6.7
     6.1.3  Shielding	     6.9
     6.1.4  Other Methods of Controlling External Exposure  .     .    .    6.11
            A.  Inventory Limitations   ....     .     .    .6.11
            B.  Good Practices     .    .     .     .     .     .     .    .    6.11
6.2  MONITORING OF EXTERNAL RADIATION DOSE   	    6.12
     6.2.1  Dosimetry Service .    .    .	6.12
     6.2.2  Review of Radiation Doses   	    6.13
6.3  ESTIMATION OF EXTERNAL RADIATION DOSE   	    6.14
     6.3.1  External Dose from Alpha Particles    .....    6.14
     6.3.2  External Dose from Beta Particles     .     .     .     .    .    6.14
     6.3.3  External Dose from Gamma Radiation    .....    6.15
            A.  Exposure Rate from Any Gamma Point Source   .    .    .    6.17
            B.  Other Methods of Calculating Gamma Exposure .     .    .    6.19
REFERENCES	6.19
APPENDIX A - ESTIMATION OF EXTERNAL GAMMA DOSE    	    £.21
                                      6.1

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DAKCOM-P 385-1
                                    FIGURES

6.1  The Inverse-Square Relationship    	     6.7
6.2  Line Source	6.24
6.3  Plane Disk Source	6.25


                                    TABLES

6.1  Half-Value and Tenth-Value Layers  	     6.10
6.2  Specific lonizetion for Electrons  .......     6.15
6.3  Conversion Factors for Computing Dose Equivalent from Exposure   .     6.16
6.4  Gamma Radiation Levels for One Curie of Some Radionuclides  .     .     6.18
6.5  Gamma-Ray Energy Absorption in Tissue   	     6.26
                                      6.2

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                                                             DARCOM-P  385-1
                         CHAPTER 6.   EXTERNAL  EXPOSURE
     External radiation exposure is the exposure of the body to radiation origi-
nating outside of the body.   For example,  an external  radiation exposure may
be received from radioactive material  in a package, from fixed contamination
on a bench top, or from an x-ray machine.   The hazard  presented by external
radiation and the methods used to control  external  exposure are dependent upon
the penetrating ability of the radiation and the dose  rate encountered.   Pene-
trating radiations such as photons and neutrons, which can pass into the body
and irradiate the internal organs, are considered more hazardous than tht rela-
tively nonpenetreting charged particles, such as alpha and beta particles.
     If a radioactive source material  is shielded so that the radiation  is
emitted as a beam, then only those parts of the body that traverse the beam
will be irradiated.   This causes a partial-body irradiation.  Common sources
of severe partial-body irradiation are radiation-producino machines such as
x-ray machines and'accelerators, which are capable of producing intense  beams
of radiation.  If the beam is large enough, or if the radioactive source mate-
rial is not shielded, then the entire  body may receive a dose of radiation;
this is called a whole-body dose.
     Exposure to external radiation can be controlled or reduced by a number of
methods, primarily the judicious use of time, distance, and shielding.  In this
chapter, these and other methods are discussed, the monitoring of external doses
is described briefly, and procedures for estimating external dose are given.
         Section 6.1  CONTROL AND REDUCTION OF EXTERNAL RADIATION DOSE

     The primary methods of reducing external radiation dose are the use of
time, distance, and shielding.   Other methods are also available.  Each task
involving radioactive materiel  should be carefully evaluated to determine which
control procedures are appropriate.   Ths ALARA (as lov, as is reasonably achitv-
able) philsophy should always be considered in the development of control
procedures.
                                      6.3

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DARCOM-P 385-1
6.1.1  Exposure Time
     The longer the t'me spent working in a radiation field,  the higher the
dose received.  An individual's working time can be reduced if work is planned
and if dry runs, complete in every detail except for the use  of radioactivity,
are performed before any work with radioactive materials or radiation-
producing machines is begun.
     A.  Basic Principle.  The total dose received at a given distance from a
particular source ,i's a linear function of the exposure time;  that is, doubling
the exposure  time doubles the total dose, and halving the time halves the total
dose.  This relationship can be expressed by Equation (6.1):

                D = D x t                                             (6.1)

where           D = radiation dose
                D = radiation dose rate, or dose per unit time
                t = time of exposure to radiation.
This equation assumes that the dose rate is constant during the exposure time.
     Minimizing an individual's exposure time is one of the simplest weys of
reducing the  individual's total dose.  For example, if the dose rate from an
unshielded source is 2 rad/hr and the time of exposure is 30 minutes, then the
radiation dose received is:
                D = 2 rad/hr x 0.5 hr = 1 red
However, if the time of exposure to the source can be reduced to 15 minutes,
then the radiation dose received is:
                D = 2 rad/hr x 0.25 hr = 0.5 rad
     B.  Control of Time.  Time spent in a radiation area can be controlled
by the use of timekeepers.  This practice requires that the dose rate  in a
given work area be known.  The maximum allowable residence time in the area
can then be calculated using Equation (6.2):

                t - D                                                  iz •"
                t	                                                  o. .
                                      6.4

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                                                             DARCOM-P  385-1
where           t = maximum allowable residence time in  the  radiation aree
                0 = maximum dose to be received by the  individual
                D = dose rate of the source.
     In instances of very high dose rates or  where rigid control  of exposure
is needed, a timekeeper should be available for each individual.   The time-
keeper stands away from the radiation source  but within  sight  of  the inci-
vidual.  When the specified time has elapsed, the timekeeper notifies the
individual, who then leaves the aree.  Personnel should  be instructed to leave
the area immediately and without question upon notification  by the timekeeper.
     C.  Reduction of Time.  Time spent working in a radiation area can be
reduced by a number of methods; examples include training, the use of power
equipment, easy access to equipment, and modification of the task to be
performed.
     The amount of time an individual spends  in a radiation  area  can depend on
how quickly and efficiently he or she can perform a task.  Training can improve
work efficiency and thus reduce exposure in day-to-day use of  radioactive
material.
     Training programs should include actual  performance of  a  procedure, com-
plete in every detail (including the use of protective clothing,  survey instru-
ments, etc.) with the sole exception that radioactivity is absent.  In some
instances, this may mean that full-scale mockups constructed.   Personnel can
tnen practice the procedures, becoming more proficient and confident.  At the
same time, the procedures should be observed  and analyzed by the  Radiation
Protection Officer (PxPO) in an attempt to reduce the working time.  Training
is discussed in greater detail in Chapter 12.
     The use of power equipment can reduce the time spent on a job.  Examples
of time-saving equipment include motorized carts for transporting materials in
warehouses; impact wrenches;  and power screwdrivers, saws, and drills.  Most
power tools can be used on the job without modification, although tools for
specialized applications may require modifications.  Equipment used in a radia-
tion area  should always be monitored for contamination before  being removed
from the area.
                                      6.5

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DARCOM-P 355-1
     Efficient access  to components, systems, or equipment can significantly
 reduce  the  time  required for their operation, maintenance, repair, or replace-
 ment.   The  ease  of  access  to equipment and components should be assessed wnen
 equipment or  work areas are being designed and should be evaluated frequently
 in existing situations.  For example, the fabrication of work platforms or the
 removal of  obstructions may improve access to equipment and reduce the time
 spent  in a  radiation area.
     Task modifications that result in decreased exposure time also reduce the
 radiation dose received.   A conscientious review of all repetitious tasks is
 the  best metnod  of  maintaining  radiation exposure ALARA.  After each task is
 completed,  all participants should discuss the task and methods to improve
 performance.  Task  modifications may also be identified in training sessions.
 All  standing  operating procedures (SOPs) should be continually upgraded and
 improved.
 6.1.2   Distance  from the Source
     Often, the  time spent near a radiation source cannot be reduced.  Person-
 nel  should  then  either work farther away from the radiation source or place
 shielding between themselves and the source.
     A.  Basic Principle.  If time and shielding remain constant, then the
 radiation dose decreases as the square of the distance from the source of
 radiation.  Consequently,  the relationship between distance and dose rate is
 commonly called  the inverse-square law.  This relationship is illustrated in
 Figure  6.1.
     The equation for  the  inverse-square law is:

               .    -      (s/
               D? = D, x   —^                                       (6.3)
                 2    l        2
where           D,  =  the dose  rate at distance 1
                Dp  =  the dose  rate at distance 2
                Sj  =  distance  1
                s~  =  distance  2.
                                      6.6

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                                                             DARCOM-P  385-1
            30
            25
          •£ 20
            15
          o
          o
            10
                 FIGURE 6.1,
2345
 DISTANCE (ARBITRARY UN ITS)
The Inverse-Square Relationship
The inverse-square law assumes that the radiation source is very small (a
point source).  If the distance between a nonpoint source and the irradiated
object is at least five times the largest dimension of the source, then the
inverse-square law can still  be used.   The inverse-square law also holds only
in a vacuum.  Attenuation of gamma rays and neutrons by air is usually negli-
gible and does not influence the dose  rate to an appreciable extent.  However,
alpha end beta particles are greatly attenuated by air, and as a result,
inverse-square calculations overestimate the eciual radiation dose for three
types of radiation.
     B.  Control of Distance.  Distance, as e method of reducing radiation
exposure, can include remote operation, moving work away from radiation
sources, and moving extraneous radiation sources away from the work are?..   Each
task should be carefully evaluated to  determine whether these procedures or
others can be used to increase tne distance between personnel and radiation
sources.
                                      6.7

-------
DARCOM-P 385-1
      (1)  Remote Operation.  Remote operation general!}' requires the use of a
manipulating  device, or remote-handling tool, to place distance between the
operator  and  the radioactive source.  For example, small radioactive sources
are  commonly  eouipped with a detachable handle or tool; most sealed sources
come with a handling device; and forceps can be used to manipulate swipes for
leak testing  sources.  Remote operations can also be performed using specially
designed  manipulators operated from behind barriers.  Manipulators range in
complexity from  simple devices used in conjunction with temporary shielding to
complex devices  built into specially  constructed hot cells.
      Some manipulations are difficult to perform using  remote-handling tools
and  can be performed faster and with  a lower resultant  dose using the fingers.
However,  direct  handling of radioactive sources should  be minimized and should
be  performed,  when  absolutely necessary, as quickly as  possible to minimize
the  high  dose rates that can result from direct handling.  A 2-cm-diameter,
 l-Ci  source  of   J  Cs,  for example,  gives  a  dose-equivalent rate of about
 1.5  x 10   rem/hr to  the  hand  when  held  in the hand.  At this rate, the maxi-
 mum  allowable dose equivalent to the  hand for one  calendar quarter (18.75  rem)
 would be  received  in  about  45 seconds.  When sources must be handled  directly,
 a  finger  dosimeter should be  worn.
      (2)   Moving Away from  Sources.   A  simple,  often-overlooked technique  for
 reducing  exposure  through the use  of  distance is for individuals  to move away
 from the  radiation source whenever possible.  For  example, if  personnel need
 to discuss a procedure,  they  should move  away from the  source.  If a  defective
 part of a machine  needs  to  be serviced,  it  should  be removed and  serviced
 elsewhere.   Tradeoffs  might be required  if -the  object to be worked on is
 bolted onto  or  close  to  the radiation source and removal time  exceeds ser-
 vicing time. The  ease of removing components should be considered during  the
 design of equipment  and  of  the building  in  which it is  to be housed.   Ideally,
 components that can  be removed from the  radiation  area  quickly and safely
 should be used.
      Another example  of  moving away from  the source is  found  in the  use  of
 gauging devices, such  as those used to  determine the surface density  of  road
 beds and  the moisture  content of roofs.   During the operation  of  these devices,
                                       6.8

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                                                              DARCOM-P 385-1
the radiation source is moved from a well-shielded configuration to a less
well shielded configuration.  The operator shou'id step back from the device
while the timer is operating and the measurement is being taken.
      (3)  Removing Other Sources.  Moving other sources, away from the work
area is the third method of using distance to reduce exposure.  For example,
piping can be backflushed to dislodge and remove radioactive debris.  Other
extraneous sources that should not be overlooked are contaminated stock bottles
and accumulations of radioactive waste.
6.1.3  Shielding
     Shielding is the use o. barriers or absorbers placed between a source and
en individual to stop some of the radiation reaching the individual.  Alpha
particles can be totally absorbed by a few centimeters of air or a few sheets
of paper.  Beta particles can be stopped by a few meters of air or a few milli-
meters of lead or plexiglass.  Gamma radiation can penetrate even dense mate-
rials such as lead; however, the intensity of gamma radiation can be reduced
to negligible levels by the use of shielding.
     The attenuation of gamma radiation by an absorbing materiel can be
described by the equation:

                I = IQ x e'vs                                         (6.4)

where           I = radiation intensity after traversing a thickness, s, of
                    material
               I  = originel radietion intensity, i.e., the radiation
                    intensity that would be observed had the attenuating
                    material not been present
                e = base of the natural logarithms (e = 2.71828)
                u = linear attenuation coefficient (cnf )
                s = thickness of the ettenueting material (cm).
     The linear attenuation coefficient, u, is related to both the attenuating
material  end the energy of the photon.  Ir, many instances, the me s s attenua-
tion coefficient is available in references, rather than the 1inear attenuation
coefficient.
                                      6.9

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DARCDM-P 385-1
The mess attenuation coefficient is the linear attenuation coefficient divided
by the density of the medium.  That is:
               vm - U/P                                               (6.5)
                                                    2
where          v  = mass attenuation coefficient (cm /g)
                m                                     -1
                v = linear attenuation coefficient (cm  )
                P = density of the attenuating material  (g/cm ).
Thus
                u = um x p                                            (6.6)

Mass attenuation coefficients as a function of photon energies are listed for
many materials in the Radiological Health Handbook (1970).  The densities of
common materials can also be found in the Radiological Health Handbook.
     The half-value layer concept  is useful in determining the necessary
shielding for gamma radiation.  A  half-value layer (HVL)  is the thickness of
material required to reduce the radiation intensity by a  factor of 2.  This
concept  is similar to the half-life of radioactive decay.   A related  term, the
tenth-value layer (TVL), is the thickness of an attenuating medium necessary
to reduce the radiation intensity  by a factor of 10.  Both HVLs and TVLs for
selected gamma sources and absorbing materials are given in Table 6.1.

                 TABLE 6.1.  Half-Value and Tenth-Value Layers

Radionuc i ide
60,
uO
i37Cs
192ir
19EAu
77 f-
""Pa
Gamma Energy
Hal
c
•J t
21
7£
2.

f-Life
24 vr
yr
d
7 d

1621 yr

1.
0.
0.
0.

0.
(MeV)
17, r.33
66
13 to 1.06
41

047 to 2.4
Hal
f-Value Laver
Concrete
6.
4.
4.
i

6.
6
8
3
1

c
Steei
2.1
1.6
1.3
_ _

2.2
(cm)
Lead
1.20
0.65
0.60
0.33

1.66
Tenth-Vel
Concrete
20.8
15.7
Id. 7
13.5

23.4
ue Leve:
Steel
6.9
5.3
4.3
..

7.4
r (cm)
Lead
4.0
2.1
2.0
1.1

5.5
                                      6.10

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                                                              DARCOM-P 3E5-1
6.1.4  Other Methods of Controlling External  Exposure
     Inventory limitations, access restrictions,  and a variety of other
approaches can be used to help control  exposure to external  radiation.
     A.  Inventory Linr'tetions.  The hazard presented by radioactive materiel
is a direct function of the quantity of material  present.   Inventories of
radioactive material in laboratories can be reduced by the frequent collection
of radioactive waste.  An inventory of  a radioactive chemical  reagent can also
be reduced by separating aliquots of the materiel  into individual vials and
storing the material that will not be used immediately away from the work area.
The material  can be separated by the user after receiving  it,  or it can be
ordered in multiple containers from most suppliers, for a  nominal fee.  Two
advantages result from this separation:  1) the radiation  hazard resulting from
spills or other accidents is reduced, and 2)  inventory recordkeeping is simpli-
fied.  The use of a centralized storage room for radioactive material net in
use or used only occasionally is often  convenient, relatively inexpensive, and
secure.  Such a facility is also helpful in keeping exposures  ALARA, since less
radioactive material is stored in laboratories  or other areas  occupied by
personnel.
     B.  Good Practices.   Other methods of reducing radiation exposures, which
are discussed in more detail in other chapters  of this manual, include the
following:
 1.  Restrict access to areas that present a  radiation hazard, through the use
     of locked doors, intrusion alarms, or guards.  The means  of restriction
     selected depends upon the radiation dose rates that are anticipated, the
     presence of interlocks, security restrictions, and budget.
 2.  Minimize the number  of authorized  radiation  workers present by limiting
     the  number of persons in an area at a given  time.
 3.  Post  signs  in radiation areas.   The work area should  be surveyed every few
     months to ensure that the signs adequately describe the hazard associated
     with  the area.   The  posting should indicate  the £ctJ=l  nazarc1 invc-~'ved; d:
     not  "overpost."   Habitual  overstatement  of radiation  hazards may cause
     personnel  to ignore  the warning signs.
                                     6.11

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DARCOM-? 385-1
 4.  Keep copies of SOPs readily available to all radiation workers.
 5.  Maintain operating Iocs for all radiation-producing machines and radioac-
     tive sources.  These logs should contain information such as date, time
     in, time out, and the names of the individuals working with the machines
     or sources.   In some cases, it may De desirable to include tne readings
     of a pencil dosimeter as each  individual enters end leaves the area.
 6.  Use a  "buddy, system" so that an individual  never works alone in a radia-
     non area, particularly in one that  is locked.
 7.  Establish  areas that require an estimation  of the dose rate before a per-
     son can  enter.
               Section 6.2  MONITORING OF  EXTERNAL RADIATION DOSE

     The  primary  DA dosimeter  is the film badge  (see Chapter 2).   Pocket dosim-
 eters  and  thermoluminescence dosimeters  (TLDs) can be used to supplement the
 film badge.   Supplementary dosimeters should  be  used when an individual is
 likely to  receive more  than 5  mrem  in 1  hour  and must be used when an  indi-
 vidual  enters  a high-radiatior. area where the dose rate may be greater than
 100 mrem/hr.
     The  dosimetry service fcr Army personnel and the responsibilities of the
 RPO in reviewing  radiation doses to personnel are discussed in the following
 sections.
 6.2.1   Dosi metry  Service
     Dosimeters for all personnel  (army,  civilian, and  contractor) working
 with DA,  ARNG, and USAR are provided by  DARCOM.  The dosimetry service is coor-
 dinated through the Lexington-BlueGress  Army  Depot  (Attn:  AKXLX-ME-1), and  an
 informational  packet  that describes the  procedures for  obtaining dosimetry
 services  is  available upon request.  Because  these procedures  are updated  peri-
 odically,  they will not be detailed here.  Actual  requisitions for dosimetry
 service should be sent  to the  appropriate Army  depot designated  in the informa-
 tional  packet  obtained  from Lexington-BlueGrass.

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                                                             DARCOM-P  385-1
     When dosimetry service is requested for an individual,  the RPO should be
prepared to provide the following information about that person:
 1.  name of individual
 2.  date of birth
 3.  social security number
 4.  work classification
 5.  type of dosimeter required (i.e.,  whole-body or extremity)--If extremity,
     include the body part it is to be  worn on (i.e.,  wrist, finger).   If
     whole-body, include the radiation  of interest (i.e., beta, gamma,  x ray,
     or neutron).   If a neutron badge is required, a beta-gamma badge  should
     also be requested because neutron  radiation is. almost always accompanied
     by gamma radiation.
6.2.2  Review of Radiation Doses
     The RPO is responsible for reviewing the radiation dose received  by per-
sonnel  (10 CFR 20, AR 40-14).  These evaluations provide the basis for showing
compliance with existing regulations and can be used to spot trends in doses
                                           t
received by personnel.
     Dosimetry services that process dosimeters report personnel  doses in terms
of reir.; no further calculations need to be performed by the RPO.   The  dose and
the date the information is received are transferred onto each individual's
record.  The RPO should review the individual records et least once each calen-
dar quarter to check for administrative overexposures and to spot any  unusual
trends  in both individual  and collective dose equivalents.  If any trencs are
noted,  especially  increases in dose equivalents, an investigation should be
conducted to determine the cause and correct any situations contributing to the
increases.  Criteria for judging whether an individual overexposure has occurred
and for reporting  any overexposures are discussed in Chapter 11,  Section 11.3.
Briefly, any monthly whole-body dose equivalent exceeding 500 mrem is  cate-
gorized as an overexposure.
                                     6.13

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DARCOM-P 385-1
              Section 6.3.  ESTIMATION OF EXTERNAL RADIATION DOSE

     Factors  thet effect  the external radiation dose a person may receive from
 a  radiation source  -.ncluae time, distance, shielding, and the activity cf the
 source.  The  first  three  factors have already been discussed.  The activity of
 the  source material, often referred to as the source strength, has a direct
 linear  relationship to the dose rate.  That is, if the source activity is
 douoled, then the dose rate is doubled.  Source activity is expressed cs the
 activity of the  parent radionuclide and  is given  in units of curies.  Terms
 such as  intense  source, large source, or sme'.l source are relative terms and
 should  be avoided.
     Many methods can be  used to estimate radiation doses from radioactive
 sources  outside  the body.  The more sophisticated methods are computer-based
 calculations  that must be performed by experienced individuals.   However, for
 evaluating a  facility's safety  requirements,  rapid estimates of  radiation
 doses  that are  relatively accurate are often  sufficient.
 6.3.1   External  Dose from Alpha Particles
     An  alpha particle must have an energy of at  least 7.5  MeV to penetrate  the
 0.07-mm-thick protective  layer  of the skin.   The  vast majority of alpha-emitting
 radionuclides have  alpha  energies less than 7.5 MeV.  For this reason, alpha
 particles do  not present  an appreciable  external  radiation  hazard,  and dose
 calculations  are generally not  required.
 6.3.2   External  Dose from Beta  Particles
     The dose rate  10 cm  from a source of beta  particles  is given by Equation
 (6.7),  which  is  valid over a wide range  of beta energies.

                 D = 2700  x A                                   '        (6.7)

 where            D = the dose rate (rad/hr)
                 A = the activity of  the  source  (Ci).
 In order to calculate the dose  rate  at distances  other  than 10  cm, the inverse-
 souare  relationship can be used.  Equation  (6..')  neglects  the ability of air,
                                      6.14

-------
                                                             DARCOM-P  385-1
and even of the source material  itself,  to  reduce  or  attenuate  the  dose rate.
Trie attenuation of beta particles  by air can  be  appreciable,  pnd  large  errors
in the calculated dose rate occur  at distances beyond  about  1 meter from the
source.
     The dose rate, in air, at the surface  of a  beta  source  is  given by Equa-
tion (6.8):
                    A
where
D =

D =
A =
S =
                        Pi
                                                  (6.6)
               Pi
dose rate (rad/hr)
source activity (mCi)
                              p
surface area of the source  (cm )
specific ionization of the  radiation,  or the average
number of ion pairs produced per  centimeter of the
radiation's path in air (taken from Table 6.2).
               TABLE 6.2.   Specific lonization  for Electrons
                                                            (a)
             Radiation
            Energy (MeV)
               0.05
               0.10
               0.20
               0.30
               0.50
               1.00
               1.50
                 Pi
            (Ion Pairs/cm)
                 250
                 175
                  96
                  76
                  60
                  £3
                  47
                                     Ranae in Air
                                        "(cm)
                                          3.02
                                         10.80
                                         32.50
                                         59.60
                                        122.00
                                        310.00
                                        526.00
            (a)  Brodsky and Beard 1960.

6.3.3  External Dose from Gamma Radiation
     Host equations for calculating the gamme-rey dose result in the exposure
(the measure of the ior.ization of air by  gamma  radiation,  measured in roentgen
(R)), rather than the absorbed dose (rad) or dose equivalent (rem).  The factors
                                     6.15

-------
DARCOM-P 385-1
for converting from exposure in units  of roenigen to dose equivalent in units
of rem are nearly eaual to 1 for photons with energies greater than about
600 keV.  Photons with energies less than about 600 keV are greatly scattered,
resulting in a dose-equivalent rate in rem that is higher than the exposure
rate in roentgen.  Therefore, for photons with energies above 662 keV, the
conversion factor 1.03 should be used, and for photons with energies below
662 keV, the conversion factors listed in Table 6.3 should be used.  The three
depths included  in the table are for dose equivalents to 1) the whole body
(1.0-cm depth, or deep dose equivalent); 2) the lens of the eye (0.3-cm depth);
and 3) the skin  (0.007-cm depth, or shallow dose equivalent).

 TABLE 6.3.  Conversion Factors for Computing Dose Equivalent from Expcsure^3'
     Photon Energy     	Conversion Factor at a Depth of
(keV)
15
20
30
40
50
60
70
80
90
100
110
120
130
140
150
662
1.0 cm ("deeD")
0.28
0.58
1.00
1.28
1.46
1.47
1.45
1.43
1.41
1.39
1.37
1.35
1.33
1.32
1.30
1.03
0.3 cm
0.67
0.79
1.07
1.29
1.46
1.47
1.45
1.43
1.41
1.39
1.37
1.35
1.33
1.32
1.30
1.03
0.007 cm ("shallow")
0.90
0.94
1.11
1.34
1.50
1.52
1.50
1.48
1.45
1.43
1.40
1.36
1.34
1.32
1.30
1.03
         American National Standards Institute (ANSI) Standard
         N13.11-1978.
                                     6.16

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                                                              DARCOM-P 385-1
     A.  Exposure Rate frorr Any Gamma Point Source.  A point source is e small
source of radiation.  The commonly used equations for calculating the exposure
rate to an individual from a point source assume that the distance between the
source and the individual is at least five times the diameter of the source or
the diameter of the individual, whichever is larger.  The simplest equation
used to calculate the exposure rate from a gamma-emitting radionuclide is based
on the specific gamma-ray constant (r) of the radionuclide, as given in
Table 6.4.
where           X = exposure rate (R/hr)
                A = source activity (mCi)
                                                      p
                T = specific gamma-ray constant ([R«cm ]/[hr-mCi])
                s = distance from the source (cm).
     If the specific gamma-ray constant for a gamma-emitting radionuclide is
not listed in Table 6.4, then the following two equations can be used.  For a
distance from a source measured in meters:
                    0.54  A £  Ei ni
                X= 	1±1	                                 (6.10)
where           X = exposure rate (R/hr)
                A = source activity (Ci)
               E. = energy of photon i (MeV)
               n. = number of photons of energy E. emitted per
         ,           disintegration
        ^ Ei ni = E:  ni + E2 r'2 + ...... Ek nk
                s = distance from the source (m)
             0.54 = constant ([R-m2]/[Me\'-hr-Ci )].
                                     6.17

-------
DARCOM-P 385-1
   TABLE 6.4.  Gamma Radiation Levels for One Curie of Some Raoicnuclides^°
                                                                     r(b)
Nucl ide
Actinium-227
Antimony-122
Antimony-124
Antimony-125
Arsenic-72
Arsenic-74
Arsenic-76
Barium-131
Berium-133
Barium- 140
Beryl 1 ium-7
Bromine-82
Cadmium- 11 5m
Calcium-47 ,v
Carbon-lll '
Cerium-141
Cerium-144
Cesium-134
Cesium-137 , ,.
Chlorine-38 '
Chromium-51
Cobal t-56
Cobal t-57
Cobalt-58
Cobal t-60
Copper-64
Eurooium-152
Europium-154
Europium-155
Gal 1 ium-67
Gellium-72
r(b)
^2.2(c)
2.4
9.8
-«2.7
10.1
4.4
2.L
<-. .0
'-2.4
12.4
•x.0.3
-14.6
-vO.2
5.7
5.9
0.35
^0.4
8.7
3.3
8.8
1.16
17.6
0.9
5.5
13.2
1.2
5.8
^6.2
•^0.3
<\,i _ \
11.6
(2) Radiolocical Health
Nucl ide
&old-198
Gold-199
Haf nium-175
Hafnium-181
Indium- 11 4m
Iodine-124
Iodine-125
Iodine-126
Iodine-130
Iodine- 131
Iodine-132
Iridium-192
Iridium-194
Iron-59
Krypton-85
Lanthanum-149
Lutecium-177
Magnesium-28
Manganese-52
Manganese-54
Manganese-56
Mercury-197
Mercury-203
Molybdenum-99
Neodymium-147
Nickel -65
Kiobium-95
Osmium-191
Palladium- 109
Platinum-197
Potassium-42
Handbook 1970.
                                           '_;	       Nuclide

                                             2.3     Potassium-43
                                            -vO.9     Radium-226
                                            ^2.1     Radium-226
                                            ^3.1     Rhenium-186
                                            ^0.2     Rubidium-86
                                             7.2     Rutherium-106
                                            ^0.7     Scandiurri-46
                                             2.5     Scendium-47
                                            12.2 '    Selenium-75
                                             2.2     Silver-llOm
                                            11.8     Silver-Ill
                                             4.8     Sodium-22
                                             1.5     Sodium-24
                                             6.4   .  Strontium-85
                                            ^0.04    Tantalum-182
                                            11.3     Tellurium-121
                                             0.09    Tellurium-132
                                            15.7     Thuliuir-UO
                                            18.6     Tin-113
                                             4.7     Tungsten-185
                                             8.3     Tungsten-187
                                            ^0.4     Uranium-234
                                             1.3     Vanadium-48
                                            ^1.8     Xenon-133
                                             0.8     Ytterbium-175
                                            ^3.1     Yttrium-88
                                             4.2     Yttrium-91
                                            ^0.6     Zinc-65
                                             0.03    Zirconium-95
                                            "-0.5
                                             1.4
     5.6
     8.25

    ^0.2
     0.5
     1.7
    10.9
     0.56
     :.o
    14.3
    ^0.2
(d)
12.0
18.4
 3.0
 6.8
 3.3
 2.2
 0.025
M.7
^0.5
 3.0
^0.1
15.6
 0.
 0.
14.
 0.01
 2.7
 4.1
       1
      ,4
      ,1
 (b)   r  = specific  gamme-ray  constant  =  R-cm /hr-mCi,  or  r/10  =  R-m"/hr«Ci.
 (c)   ^  = approximately.
 (d)   A  Manual  of Radioactivity  Procedures  1961,  Appendix A,  pp.  137-140.
                                      6.18

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                                                             DARCOH-P  385-1

When the distance from the source is  measured in  feet,  an  approximation of the

exposure rate is given by Equation (6.11).
          k
         £
x= - 1  -                                     (6.11)
                     6 A     E.  n.
where           X = exposure rate (P./hr)
                A = source activity (Ci)
               E. = energy of photon i  (MeV)

               n. = number of photons of  energy E^  emitted per
                    disintegration
                s = distance from the source (ft)
                6 = constant ([R-ft2]/[MeV-hr-Ci]) .'

     B.  Other Methods of Calculating Gamma Exposure.   In special cases, such
as for calculating of gamma dose from line sources  or from planar disc sources,
more complex equations than those listed  above are  needed.  These equations,
presented in Appendix A, are for estimating exposure based on the intensity of

the photon radiation.
                                  REFERENCES


American National Standards Institute (ANSI).  1978.  Criteria for Testing
  Personnel Dosimetry Performance.  ANSI K13.ll, New York.

Brodsky, A., and G. V. Beard.   1960.  A Compendia of Information for Use in
  Control line Radiation Emergencies.  TID-8206 (Rev.), U.S. Atomic Energy
  Commission, Washington, D.C.

National Council on Radiation Protection and Measurements (NCRP).  1961.
  A Manual of Radioactivity Procedures.  NCRP 28, Washington, D.C.  Also
  published in 1961 as National Bureau of Standards Handbook No. 80,
  Washington, D.C.

Radiological Health Handbook.   1970.  U.S. Department of Health, Ed-.-ation, end
  Welfare, Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 10, Part 20, "Standards for
  Protection Acainst Radiation."  U.S. Government Printinc Office, Washincton,
  D r
  L> • w «
                                     6.19

-------
DARCOM-P 385-1
U.S. Department of the Army and Defense Logistics Agency.  Medical Services -
  Control and Recording Procedures for Exposure to Ionizing Radiation and Radio-
  active Materials.  AR 40-14, DL.AR 1000.28. Washington, D.C.
                                     6.20
                                                                                        4

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                                                              DARCOM-P 385-1
                                  APPENDIX  A
                      ESTIMATION OF EXTERNAL  GAMMA  DOSE
     Equations were presented in Section 6.3.3  for  estimating  the exposure
rate from point sources of gamma radiation.   A  slightly  more  complicated
method of dose estimation involves first calculating  the flux, c.r intensity,
of the radiation, which is measured in photons  per  unit  area  in unit time
                      2
(usually in photons/cm -sec), and then using the  flux to calculate the absorp-
tion of the radiation's energy by body tissues.
     To calculate the flux from any source,  it  is necessary to consult a decay
scheme to determine the number of photons emitted per disintegration.
Cobalt-60, for example, emits two gamma rays per  disintegration, and both of
these must be taken into account in the calculation of the flux.
A.I  FLUX FROM A POIK" SOURCE
     For a point source, the photon flux can be calculated from:

                j  = (3.7 x 1010) x A x n                              (6
                         4 x IT x s
where           I  = photon flux for photons  of a given energy
                                2
                    (photons/[cm -sec])
                A  = source activity (Ci)
                n  = fraction of disintegrations that yield a gamma
                    ray of a given energy (photons/disintegration)
                s  = distance from the source (cm)
                r  = pi = 3.1416
       3.7 x 10   = constant (disintegrations/[sec-Ci]).

A. 2  FLUX FROM A LINE SOURCE
     A typical problem might entail calculating the dose  rate from a pipe thai
contains radioactive materiel.   In principle, the problem could be solved by
                                     6.21

-------
DARCOM-P 385-1
considering the pipe (or line) to be a series of point sources, calculating the
flux from each point, and then aading up the dose rates from all the points.
At best, tnis would be tedious. Therefore, the following equation has been
derived to calculate the photon flux from a line source.  The equation is valid
for any point, p, along the source.
                     [3.7 x 10
                             10
where
                I =
            I  =
                           )  x A   x n
                         4 x
                               x s
               A  =
                s
                n =
             photon flux for photons  of a  given  energy
             (photons/[cnT-sec])
             source activity per  unit length of  pipe  (Ci/cm)
             fraction of disintegrations that yield  a  gamma ray
             of a given energy (phctons/disintegration)
             3.1416
             distance from the pipe (cm)
             the angles shown in  Figure 6.2 (radians)
     j.   t-
3.7 x 1010 = constant (disintegrations/[sec-Ci]).
           6,, e, =
                   t-\v\Vv\\\\\\\\ \V\\\\\\\\\\\\\\\N
                                         . 62  /
\ "i ' y
^xv ' '

p
                            FIGURE 6.2.   Line Source
h.
:LUX  FROM  A  PLANE  DISK SOURCE
     The dose rate from a  plane disk  source  can  be  used  to  approximate  the  dose
 received from radioactive  materiel on the  ground.   The photon  flux et a point,
 d,  from a  plane disK source can be estimated from the equation:
                                     6.22

-------
                     3.7 x 1010) >: A  x n
                                          x log
                                                 R
2
                                                              DARCO.M-P 385-1
                   (6.14)
                                                        J
where           I = photon flux for photons of a given energy
                                o
                    (photons/tern -sec])
                                                        2
               A  = source activity per unit area (Ci/cm )


                n - fraction of disintegrations that yield  a photon

                    of a given energy (photons/disintegration)


                R = radius of the source (cm)

                s = distance from the source (cm),  as shown in Figure 6.3

       3.7 x 10   = constant (disintegrations/[sec'Ci]).
                        FIGURE 6.3.   Plane Disk, Source



A.3  ABSORPTION OF ENERGY BY TISSUES


     The absorption of energy by body tissues is given by the energy absorption

coefficient for the radiation in tissue.   The basic equation is:


                              -  k
                X = 5.75 x 1C'3 X;  I.  (ven). I.                      (6.15)




where           X = exposure rate (R/hr)
                                            2
               I. = photon flux (photons/[cm -sec])
                                                          2
           (v  )• = mass energy absorption coefficient (cm /g)

               E. = the photon energy (MeV)

   k  5.75 x 10"5= constant  ([R-c-secj/[Me\•hr-pho-on])
                                    6.23

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                                                              DARCOM-P 385-1

                          CHAPTER 7.  DECONTAMINATION




7.1  GENERAL DECONTAMINATION PROCEDURE  	    7.3

7.2  PREPARATION FOR DECONTAMINATION    	    7.4

     7.2.1  Area Definition and Access Control    .....    7.5

     7.2.2  Personnel Protection During Decontamination     .     .      .    7.5

     7.2.3  Evaluation of Decontamination Needs   .     .  •        .      .    7.5

7.3  PERSONNEL DECONTAMINATION     	    7.6

     7.3.1  Personnel Decontamination Methods     .....    7.9

     7.3.2  Specific Personnel Decontamination Procedures   .     .      .    7.12

     7.3.3  Personnel Decontamination Kit     	    7.13

7.4  EQUIPMENT AND MATERIAL DECONTAMINATION   	    7.14

     7.4.1  Decontamination Methods     ...'....    7.14

            A.  Cleanino, Abrasive, Chemical and Electrochemical
                Methods"	7.14

            B.  Acing and Seeling  ........    7.15

     7.4.2  Selection of Decontamination Methods  .....    7.15

     7.4.3  Specific Decontamination Techniques   .....    7.16

REFERENCES	7.16

APPENDIX A - PERSONNEL DECONTAMINATION PROCEDURES 	    7.17

APPENDIX B - EQUIPMENT AND MATERIAL DECONTAMINATION METHODS .     .      .    7.29

APPENDIX C - EQUIPMENT AND MATERIAL DECONTAMINATION PROCEDURES   .     .    7.39
                                      7.1

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DARCOM-P 365-1




                                    FIGURES






7.1  Personnel  Contamination Record      .......     7.8



7.2  Decontamination and Evaluation Log  .....               7.10










                                    TABLES






7.1  Personnel  Decontamination Methods   .....               7.11



7.2  Contamination  Removal Methods  ........     7.31



7.3  Sealing Methods	7.36



7.4  Decontamination Methods for Various Surfaces 	     7.37
                                      7.2

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                                                             DARCOM-P  385-1
                          CHAPTER 7.   DECONTAMINATION
     The presence of contamination, or unwanted radioactivity,  can result from
normal operations, maintenance activities,  and abnormal  events  such es  equip-
ment failure, accidents involving radioactive materials,  and  improper work
practices.  Detecting and determining the extent of contamination usually
require use of the survey techniques described in Chapter 4 of  this manual.
When the extent of contamination has been determined and  appropriate barriers
have been established to limit further spread, the process of cleanup,  or
decontamination,  can begin.
     Decontamination has three purposes:   1)  to prevent  any uptake of radioac-
tive material into the human body; 2) to limit external  radiation exposure;
and 2) to prevent further spread of contamination.  Decontamination may be
required for personnel, for equipment of all  types and sizes, and for large
surface areas such as land,  floors, roads,  or buildings.   Tne basic method of
decontamination is to remove radioactivity  by one or more wet or dry processes.
Two other approaches that decrease the level  of removable contamination are
allowing short-lived radionuclides to dissipate through  radioactive decay and
fixing contamination in place by covering or  sealing it.   These approaches are
not generally recognized as  decontamination processes; however, under some
circumstances they may be the best possible actions.  For that reason they are
included in this  chapter.
                Section 7.1  GENERAL DECONTAMINATION PROCEDURE

     The specific decontamination methods and procedures selected for use in
particular circumstances depend on the type,  extent, and location of the con-
tamination;  however,  the general  approach to  decontamination outlined below
applies to most situations.
 1.   Control  access  to contaminated areas.
 2.   Provide  personnel protection, including  appropriate clothing, for workers,
                                      7.3

-------
DARCOM-P 3E5-1
 3.  Evaluate whet is to'be decontaminated.
 4.  Obtain necessary equipment and materials.
 5.  Survey ell  items to be released to an unrestricted area.
 6.  Begin with  the mildest decontamination methods and progress to harsher,
     more abrasive, or caustic methods as required.
 7.  Work from the outside of the contaminated area to the inside.
 8.  Isolate  all  clean areas from contaminated areas.  Clean areas adjacent to
     those being decontaminated should be covered with taped-down paper or
     plastic  to  prevent  recontamination.
 9.  Minimize the generation of contaminated  liquids and airborne radioactivity
     during the  work, and collect and treat as contaminated waste all liquids
     generated and materials used during decontamination.
 10.  Survey between major steps in the decontamination process  (i.e., between
     successive  applications of each technique and between different
     techniques).
 11.  Continue decontamination until contamination  levels are reduced to
     appropriate levels  as given in Chapter 4, Appendix A, of this manual.
 12.  Document the completion of decontamination, including the  name  of the
     individual  performing the final survey,  the date, and the  survey results.
     (Documentation of intermediate survey results may also  be  desirable.)
 These  steps ere  discussed further in the following sections  on  preparation  for
 decontamination  end on methods for decontaminating personnel, equipment,  and
 materials.  Specific  procedures for applying  these methods are  given in  the
 appendixes at the end of this chapter.


                  Section 7.2.  PREPARATION FOR DECONTAMINATION

     Preparation fcr  decontemir.atior includes establishing boundaries within
 which  contamination is to be contained  and controlling access to the area;
                                       7.4

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                                                             DARCCtt-P  3E5-1
provicing radiation protection for personnel  involved  in  the  decontamination
operation; and evaluating the specific  items  to  be  decontaminated.
7.2.1  Area Definition and Acress Control
     Contaminated areas (e.g., floor or land  areas)  should  be posted and
barriers established to limit access to and further  spread  of contamination.
In more complex situations (e.g., pieces of contaminated  equipment  or several
rooms within a building), it may be necessary to segregate  and  isolate  areas  of
relatively high contamination from those of relatively low  contamination.
Segregation can be useful in determining what effort will be  required to com-
plete decontamination, and it helps in  the establishment  of priorities, or a
sequence for the work.
7.2.2  Personnel Protection During Decontamination
     Radiation protection requirements  for decontamination  operations are the
same as those for work in contaminated  or high-dose-rate  areas.   The key
concerns are to protect personnel from  becoming  contaminated  and to keep both
individual and collective radiation doses at  levels  that  are  es  low as  is
reasonably achievable (ALARA).
     Personnel can be protected against contamination  by  the  use of protective
clothing.  For decontamination operations involving tritium,  organic solvents,
or other wet substances, clothing impervious  to  the liquids involved should be
selected to prevent absorption of contamination  through the skin.  Respiratory
protection should be used in highly contaminated areas, particularly when
decontamin~f :n methods may generate or stir  up  loose  contamination.  Step-off
pads should be positioned at exits from the contaminated  area.
     The radiation dose to personnel during decontamination cen be monitored
and controlled using standard instruments and techniques  (e.g.,  thermo"lumines-
cence dosimeters, pocket dosimeters, dose rate monitoring,  and surveys of
individuals).
7.2.3  jiyBluation of Decontamination Needs
     Many materials such ts wood, damaged equipment, scrap  metcl, cables,
cords,  hoses,  and clothing require more time  and effort to  decontaminate than
                                      7.5

-------
DARCOM-P 365-1
 they  are worth.   In general, these items should be disposed of as contaminated
 waste (see  Chapter 10).  If proper control procedures are used,  some contami-
 nated items  can be assigned for use in permanently contaminated  areas (e.g., in
 nuclear  reactor facilities or radiochemistry laboratories).
      Decontamination is begun at the perimeter of a large contaminated area and
 progresses  toward the center.  When appropriate, decontamination is from top to
 bottom of vertical surfaces.  Perimeters should be surveyed and  reestablished
 as  the size  cf the contaminated area is reduced.  The environment or topography
 may impose  additional considerations for sequence; on sloping or windy terrain,
 decontamination should begin with the highest or upwind points,  respectively.
 The presence of drains, sumps, or sewers warrants special consideration.  Where
 they exist  specifically for the collection of radioactive liquids, they should
 be  used  during decontamination; however, if they could become pathways for the
 further  spread of contamination to the environment, every effort should be made
 to  ensure their isolation.
      Where  areas with varying degrees of contamination can be identified,
 adequately  segregated, and controlled, the priority for decontamination is less
 critical.   In general, work should begin where  the most significant reduction
 in  personnel  dose can be achieved through early decontamination.  Other factors
 that may contribute to setting decontamination  priorities  include the avail-
 ability  of  materials, equipment, and personnel, and how immediate the need is
 for uncontrolled  access to or use of the area or equipment to be decontaminated.
                    Section 7.3.  PERSONNEL DECONTAMINATION

      Before  external decontamination of an individual is begun, the following
 steps  should be  taken to help establish priorities for decontamination and
 for  follow-up efforts:
  1.   Observe any physical effects to the contaminated person, such as bleeding,
      irregular breathing rate, burns, or shock.
  2.   Assess  the  extent of any injuries:  medical treatment of injuries takes
      priority over decontamination.
                                       7.6

-------
                                                             DARCOM-P 385-1
 3.  Immediately flush with water any skin contamination  involving  caustic,
     corrosive, or organic-solvent solutions.
 4.  Deterir.-ine tne extent and magnitude of contamination  using  personnel  survey
     techniques.
 5.  Document survey results.
 6.  Remove contaminated clothing, place it in  a  plastic  bag, and hold  it for
     further disposition.
 7.  Obtain assistance from medical  personnel  if  decontamination of eye:, ears,
     nose, or mouth is necessary or if harsh chemicals  (other than  soap  and
     water) will be required.
 8.  Investigate to determine how the contamination  occurred.
     For accident situations involving both contamination and personnel  injury,
medical treatment must take priority over decontamination.   The only exceptions
to this are 1) when an extremely high level of  contamination presents a  greater
hazard to the victim than does the physical injury,  and  2)  when decontamination
can be performed prior to treatment of minor injuries,  and  the  medical  officer
concurs.  In all cases, decontamination must be performed in a  manner that pre-
vents indiscriminate soreading of contamination.
     When personnel contamination is suspected  or detected, a thorough  personal
survey should be performed.  Contaminated clothing should be removed and bagged
for subsequent disposition.  During the survey, particular attention should  be
paid to locating any hot spots of contamination.   The  results of this survey,
including the locations and measured levels of  contamination, should be docu-
mented.  Figure 7.1 is an example of a data sheet for  assessing personnel
contamination.  Refer to AR 385-40 to determine whether an accident/incident
report is required.
     In the event of a known or suspected internal deposition of radioactivity
(by inhalation, injection, consumption, etc.),  arrangements must be made for a
prompt bioassay (see Chapter 5) and for consultation with the medical stc^f.
The treatment and removal of internally dsrcsited radioactivity is  i highly
specialized field, and the assistance of qualified medical  personnel is
essential.
                                      7.7

-------
DARCOM-P  385-1
         Name:
         Date of Incident:
 PERSONNEL CONTAMINATION RECORD




	 Social  Security Numoer:




               Time of Occurrence:
         Location of Incident:
         Description of  How  Contamination occurred:
         How was contannation discovered?
                                        SURVEY RESULTS
         Survey Performed  by:
         Survey Instrument  Manufacturer and Model:



                   Serial Number:



             Indicate type,  extent, and magnit
                     o«^miriation  on  figure below
                      FIGURE  7.1.   Personnel  Contamination  Record
                                             7.8

-------
                                                             DARCOM-P  385-1
7.3.1  Personnel Decontamination Methods
     Personnel should be decontaminated as  quickly  as  possible  using  the  least
drastic means necessary.  Decontamination  efforts  should  begin  with mild
methods, which should be continued as  long  as  the\  are  effective,  and progress
to harsher methods only as required.   Medical  supervision is  required when
harsh materials or methods are used.   Extreme  care  should be  taken to prevent
the spread of contamination to any skin or  body  opening,  and  all  liquids
generated and materials used during decontamination should be collected and
treated as contaminated waste.  Personnel  performing the  decontamination  should
take all necessary precautions to protect  themselves.
     The progress of decontamination  should be closely  monitored  by surveying
between successive washings or techniques.   A  log  of methods  used  and survey
results should be maintained.   A typical  log sheet  for personnel  decontamina-
tion is shown in Figure 7.2.
     Basic methods for personnel decontamination are listed in  Table  7.1  in
increasing order of harshness, along  with  their  advantages, disadvantages,  and
decontaminating action, and some commonly  available agents for  each method.
This is not a complete listing; many  other  agents  have also been  used
effectively.
     Simple washing methods (mild soaps,  abrasive  soaps,  and detergents)  are
streightforvsrd in their use.   Generally, mild soap and water is  sufficient  for
localized skin decontamination.  A modification  of  simple washing is  to make  a
pcste by applying a powdered household laundry detergent  to wet skin  and
rubbing.  This method provides somewhat more effective decontamination,
although it is also more irritating to the  skin.  Cool  or lukewarm water should
be used for all  washing and rinsing.   Hot water  causes the skin pores to open,
driving contamination deeper into the skin.  Cold  water closes  the pores,
trapping contamination in the  skin.
     If extensive washing is required or harsher methods  must be  used, obtain
assistance from medical personnel before proceeding.  In  these circumstances,
particular attention must be given to preventing skin  carnage.  Chapping or
cracking of the skin from repeated washing  or  abrasion can lead to the intake
                                      7.9

-------
DARCOM-P  385-i
                                    c ECO'- • »V"! '.A" 11)!,  -1L JL-'fiD
                   •'. c r i' c *." c r L c- v e !
         ternfiatinr     --ir ~r£o   Uecontannet'f.'     .o^tar-matior.  ^tve
         Decor teriir.atir.r  Conpletec Dy:	




         c IOALE>Y  (Cnect-  is eprlictble.  Attach results »men




                D T'-vivo count                   D nasal  swipe-




                D urine  safi'p'ic




                L3 none  recuired



         FOLLOW-UP




         Further evaluation nee.




               Tyos?
         Siniler to  previous occurrencps?      Yes




             L f D1 c i n
               taken  to  prevent  recurrence:
         Conwer.ts:   (attach  if more  space recuired)
         Radiation  Protection Officer:



         P.eviewec bv:
Date:_



Date:
                   FIGURE 7.2.   Decontamination  and  Evaluation  LOG
                                            7.10

-------
                                                                                                                   DARCOK-P   385-1
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                                                                       7.11

-------
DARCOK-P 385-1
of radioactivity through minor cuts.   The use  of  a  hand  cream  or lotion
between washings can help prevent chapping.   If contamination  still  remains
after extensive washing, covering the contaminetion with plastic (e.g.,  a
plastic glove taped over the hand) and allowing the skin to  sweat can  provide
further decontamination.
     Chemical complexing agents, which should  be  used  only under medical
supervision,  remove contamination by chemical  interactions  such as  ion
exchange and bonding.  A solution of EDTA (ethylene-diamine-tetra-acetic acid)
can be prepared by dissolving 10 grams of EDTA salts  in  100  ml  of water.   This
solution, which can be prepared in advance and stored,  is applied to the skin
with cotton  swabs or sponges.  Following each  application of the solution, the
area should  be rinsed.
     Oxidizing agents decontaminate by chemically removing  the contaminant and
a thin layer of skin.  Household bleach is a weak oxidizing  agent that can be
applied full strength using cotton swabs or soonges.   A stronger oxidizing
agent is potassium permanganate (KMnOJ followed  by sodium bisulfite (NaHSO^).
Saturated solutions of each of these chemicals should  be made  up at the  time
of need by dissolving crystals of each in a small amount of  water.  (A satu-
rated solution is one in which no more crystals will  dissolve.)  The KMnO.
solution is  painted thickly onto the skin and allowed  to dry.    It is then
removed by gently scrubbing with the NaHSO^ solution.   The skin should be
rinsed after each use of oxidizing agents, and their  use should be discon-
tinued if the skin becomes tender.  Medical supervision is required for the
use of this method.
     Commercial decontamination agents—soaps, detergents, and  complexing
agents—are  available under various trade names.   They should be used with
medical assistance and the manufacturer's instructions should be followed.
7.3.2  Specific Personnel Decontamination Procedures.
     Specific procedures for personnel decontamination are provided in Appen-
dix A of this chapter.  Procedures for decontaminating the skin, heir and
scalp, body, face, eyes, ears, mouth, and nose are included.
                                     7.12

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                                                               DARCQM-P 385-1
7.3.3  Personnel Decontamination Kit
     A personnel decontamination kit should be assembled for field use, or
supplies should be available at designated decontamination stations.   Typical
materials that should be included are as follows:
                        Item
      Applicators, cotton-tipped
      Cotton bells
      Cleansing tissues
      Sterile gauze pads (5 cm x 5 cm)
      Hand brushes
      Masking tape
      Plastic cups (4 02.)
      Plastic cups (1 oz.)
      Plastic bags (for waste)
      Scissors
      Surgical gloves (talced)
      Flexible tube
      Filter paper (for smears)
      Envelopes (to hold smears)
      Hand cream
      Soaps:  Regular bar soap
              Abrasive soap
              Detergent (household laundry  type)
      Reagents:   Household  bleach
                 Potassium  permanganate  crystals
                 Sodium bisulfite  crystals
                 EDTA salts
      Basin  (for field use)
      19-liter jug (for field  collection
         of  liquids)
      Pencils  or pens
      Paoer
Approximate Quantity
    500
    200
      4 boxes
    400
      4
      1 roll
     25
     25
     20
      1 pair
      1 box
      1.2 meters
      1 box
      1 package
      1 jar
      2 bars
      1 bar
      1 box
      1 bottle
      1 smell jar
      1 small jar
      1 smell jar
      1
      1
      3
      1 p£
                                     7.13

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DARCOM-P 385-1
              Section 7.4.  EQUIPMENT AND MATERIAL DECONTAMINATION

     Equipment and materials may need to be decontaminated for a number of
 reasons,  including:
  1.  for  release for unrestricted use
  2.  for  the  salvage of  valuable material
  3.  to reduce the potential for exposure of personnel  to radiation
  £.  to reduce the volume of contaminated waste.
 Decontamination should be performed as soon as possible after contamination
 occurs.   This  is particularly true for liquid contaminants, which can penetrate
 farther into  materials es contact time increases.
     Materials that cannot be easily or cost-effectively decontaminated should
 be  evaluated  for possible limited use in restricted'areas, or disposed of.
 Porous  items  (such as wood and unsealed concrete), intricately designed eouip-
 ment, and  items of low replacement cost tend to fall in this category.
 7.4.1   Decontamination Methods
     Many  methods and techniques have been developed for decontaminating equip-
 ment (TM  3-220).  Most are physical or chemical cleaning processes.  Two other
 methods, which are not considered true decontamination, are radioactive decay
 (aging) and sealing contamination in place.
     A.   Cleaning, Abrasive, Chemical, and Electrochemical Methods.  True
 decontamination entails  removing radioactivity by cleaning, abrasive,
 chemical,  and  electrochemical methods.  Cleaning methods are nondestructive
 but may require that equipment be disassembled for maximum effectiveness.
 Cleaning  includes both manual (wiping, mopping, vacuuming) and mechanical
 (soaking,  spraying, vibrating) techniques.  Abrasive methods are destructive,
 involving  the  progressive removal of the contaminated material.  Chemical
 methods include both nondestructive techniques  (e.g., the use of detergents
 and complexing agents, which remove contamination by emulsifying and  ion
 exchange), end destructive techniques (e.g., the use of caustics and  acids,
 which dissolve and corrode contamination and sometimes the base material).
                                     7.14

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                                                              DARCOM-P  385-1
Electrochemical  methods are destructive,  electrolyticolly  removing contami-
nation and some  of the base material.   Table  7.2 in  Appendix  E  summarizes  the
applicability,  advantages,  and disadvantages  of specific methods  "in each of
these broad classes.
     E.   Aoing  and Sealing.  Aging involves  isolating  a  contaminated object
until radioactive decay has reduced the contamination  to an acceptable level.
This approach is suitable only when short-lived rac'ionucl ides are involved.
Aging for 10 half-lives reduces the contamination  level  to one-thousandth
(i/1000) of the  original  level.
     Sealing involves fixing radioactivity in place  by covering  it with an
impermeable material  such as earth, asphalt,  cement, paint, or plastic.
Sealing  is most  effective for alpha end low-level  beta-gamma  contamination.
Most sealants are adequate  for shielding alpha and-some beta  contamination.
However, thick,  high-density materials (e.g., concrete or  several inches of
earth) ere needed to  sufficiently attenuate  gamma  rays.   Sealing  is of most
value where the  primary concern is preventing the  spread  of relatively low
levels of contamination,  and where dose rate  is not  a  serious concern.
Table 7.3 in Appendix B provides a brief description of methods  used for
sealing  contamination in  place.
7.4.2  Selection of Decontamination Methods
     The selection and application of decontamination  methods is  dependent
upon the material or  equipment to be decontaminated.  For  extensive decontami-
nation,  outside  assistance  may be necessary.   Methods  may be  used individually
or in combination.  When  more than one method is  to  be used,  the least harsh
or abrasive method should be used first.   Table 7.4  in Appendix B lists some
types of surfaces, materials, and equipment,  and  identifies methods suitable
for decontaminating each.  In the case of contaminated commodities, consult
the appropriate  technical manual for decontamination procedures.
     Where extensive  decontamination work is  to be performed, several methods
or combinations  of methods  can be tested on  different areas of the  same sur-
face end t;'i results  can  be compared using the decontamination facto- (DF),
the commonly used measure of decontamination  effectiveness.  The DF is cal-
culated  as follows:
                                     7.15

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DARCOM-P 385-1
                        surface contamination before decontamination
               DF =
or             DF =
residual  surrace  contamination  after  (Decontamination

dpm before decontemineton
dpm after decontamination
The hiciher the DF, the more effective the method.  High DFs are generally
achieved with the initial application of any method, but subseouent applica-
tions may be less effective.  Rinsing usually improves the DF of any decon-
tamination procedure.
     All other factors being equal, the decontamination method with the high-
est DF should be used.  However, the resources available for decontamination
and the destructiveness of each method also affect the choice of decontamina-
tion methods.  The RPO should maintain records of the DF obtained during each
decontamination  in order to assist in the selection of procedures for future
decontaminations.
7.4.3  Specific  Decontamination Techniques
     Specific techniques for decontaminating equipment and materials are
described in Appendix C.  The techniques include the use of tape patches,
vacuum cleaning, wiping or mopping, water jets,  detergents, complexing agents,
organic solvents, acids, and caustic solutions.
                                  REFERENCES
U.S. Department of  the Army, Headquarters.  Chemical, Biological  and
  Radiological  (CBR) Decontamination.  TM 3-2300, Wasninoton,  D.C.
U.S. Department of  the Army, Headquarters.  Safety  - Accident  Reporting  and
  Records.  AR  385-40, Washington, D.C.
                                     7.16

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                                                               DARCOM-P 385-1
                                  APPENDIX A

                     PERSONNEL DECON'TAKINATION PROCEDURES

A.I  LOCALIZED SKIN DECONTAMINATION
Prerequisites
 1.  Survey tc identify extent end magnitude of contamination.
 2.  Obtain medical assistance if harsh decontamination methods will be
     necessary.
 3.  Collect materials needed for decontamination.
 4.  Document steps and survey results in the appropriate log.
Precautions
 1.  Medical treatment takes priority over decontamination.
 2.  Do not spread contamination to clean areas.
 3.  Do not reuse applicators (replace after each time skin is touched).
 4.  Handle all waste materials as contaminated waste.
 5.  Stop decontamination procedures if evidence of skin damage appears or if
     person complains of soreness or stinging; contact medical personnel for
     assistance.
 6.  Person performing decontamination should take  precautions not to become
     contaminated (i.e., wear gloves and other protective clothing as
     required).
Procedure for Spot Decontamination
 1.  Press masking tape over contaminated area.
 2.  Slowly remove end discard.
 3.  Repejt is  nece::-.cry.  evading sk'r. -Irritation.
 4.  Proceed with area decontamination if tape method is not effective.
                                    7.17-A

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DARCOM-P 385-1
Procedures for Area Decontamination (in increasing  order of  harshness)
 1.  Soap and water
     Use one or more of the following techniques until  no further reduction in
     contamination occurs:
     (a)  Wash with mild bar soap and cool or lukewarm water.
     (b)  Wash with abrasive soap and water; this method is  particularly
          applicable to toughened skin areas such as fingertips and the palms
          of the hands.
     (c)  Swab with mild liquid soap using cotton-tipped applicators, then
          rinse with water.
     (d)  Use a soft hand brush in combination with any of the above
          techniques.
     Consult with medical personnel before proceeding with harsher techniques.
 2.  Detergent and water
     (a)  Wash using a detergent and water.
     (b)  Make a paste by first lathering the skin area with mild soap and
          water, then applying detergent powder to lathered skin and working
          into a paste; rub skin area and rinse paste off.
 3.  Mild oxidizing agent
     Apply household bleach full strength using cotton sponges or applicators.
     Rinse after each application. Continue until no further contamination
     reduction occurs.
 4.  EDTA solution
     Prepare a 10% EDTA solution by dissolving  10 grams of EDTA salts
     (Na^EDTA) in 100 ml of water.  (This solution can be prepared in advance
     and stored.)  Apply the solution to the skin using cotton sponges.  Rinse
     after application.  Do not apply more than two times.
                                    7.io-A

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                                                              DARCOK-P  385-1
 5.  Strong oxidizing agent
     Prepare a saturated solution of potassium permanganate (KMnO^)  by
     dissolving KMnO, crystals in 1 ounce of water until  no more crystals
     will dissolve (solution will be a dark -ed or brown).   Prepare  a  saturated
     solution of sodium bisulfite (NaKSOJ by dissolving  NaHSO,  crystals in
     1 ounce of water until no more crystals will  dissolve.  Paint contaminated
     skin area with KMnO. solution using cotton applicators or sponges.
     Allow to dry, then repeat two more times.  Remove brown itain by  gently
     swabbing with NaHSO., solution using cotton swabs. Then rinse with
     water.  If necessary, repeat the application  one time.
 6.  Further decontamination
     If contamination remains after ell these procedures  have been tried, a
     medical expert should be consulted for assistance.
 7.  Post-decontamination
     Following successful decontamination, apply hand lotion to skin to  prevent
     chapping.
 8.  Sweating
     If soreness or tenderness develops during decontamination,  the procedure
     being used should be stopped for a time.  During this interval, the
     contaminated area can be covered with plastic and allowed to sweat, thus
     cleansing the area from the inside out.  The  area should then be  gently
     washed in lukewarm water.  (This method is particularly useful  for
     decontaminating the hands,  using surgeons' gloves for covering.)

A. 2  HAIR AND SCALP DECONTAMINATION
Prerequisites
 1.  Survey to identify extent and magnitude of contamination.
 2.  Collect materials needed for decontamination.
 3.  Document steps and survey results in the appropriate log.
                                    7.19-A

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DARCOM-P 385-1
 Precautions
  1.  Medical treatment takes priority over decontamination.
  2.  Do not spread contamination to clean areas.
  3.  Do not reuse applicators (replace after each time skin  is touched).
  4.  Handle all waste materials as contaminated waste.
  5.  Stop decontamination procedures if evidence of skin damage appears or if
     person complains of soreness or stinging; contact medical personnel for
     assistance.
  6.  Person performing decontamination should take precautions not to become
     contaminated (i.e., wear gloves and other protective clothing as
     required).
 Procedure
  1.  Contaminated person should remove outer clothing and put on overalls or a
     laboratory coat and surgeons' gloves.
  2.  Wrap a towel around the person's neck.
  3.  Bend the  person over a sink or basin and wash hair using mild soap or
     shampoo.  Massage hair and scalp carefully, preventing lather or water
     from entering the ears, eyes, nose, or mouth.
  4.  Rinse heir with water.  Change the towel if it becomes saturated.
  5.  Thoroughly dry the hair with towels (do not use  a blow dryer).
  6.  Resurvey  heir, also checking face and neck.
  7.  Repeat shampoo process as long as it is effective.
  8.  If shampooing ceases to be effective, contaminated heir  can be  cut with
     scissor or clippers and the scalp can be decontaminated  using the
     procedures for localized skin decontamination.
                                     7.20-A

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                                                              DARCOM-P 385-1
A. 3  GENERAL BODY DECONTAM NATION
Prerequisites
 1.  Survey to identify extent and magnitude  of  contamination.
 2.  Collect materials  needed for decontamination.
 3.  Document steps and survey results  in  the appropriate  log.
Precautions
 1.  Medical treatment  takes priority over decontamination.
 2.  Do not spread contamination to clean  areas.
 3.  Do not reuse applicators (replace  trter  each  time skin  is  touched).
 4.  Handle all  waste materials as contaminated  waste.
 5.  Stop decontamination procedures if evidence of skin damage appears or if
     person complains of soreness or stinging;  contact medical  personnel  for
     assistance.
 6.  Person performing  decontamination  should take precautions  not to become
     contaminated (i.e., wear gloves and other protective clothing as
     required).
Procedure
 1.  Remove clothing.
 2.  Shower with  lukewarm water.
 3.  Lather, using mild soap and soft brush or scrub pad.
 L.  Rinse, taking care not  to spread contamination to skin  or body openings.
 5.  Survey and  repeat  as necessary.
 6.  If only localized  contamination remains, follow procedures for localized
     skin decontamination.
                                    7.21-A

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DARCOM-P 385-1
A.4  FACIAL DECONTAMINATION
Prereoin'si tes
  1.  Survey to  identify extent and magnitude of contamination.
  2.  Collect materiels needed for decontamination.
  3.  Document steps  and survey results in the appropriate log.
Precautions
  !.  Medical treatment takes priority over decontamination.
  2.  Do not spread contamination to clean areas.
  3.  Do not reuse applicators (replace after each time skin is touched).
  4.  Handle ell waste materials as contaminated waste.
  5.  Stop decontamination  procedures if evidence of skin damage appears or if
     person complains of soreness or stinging; contact medical personnel for
     assistance.
  6.  Person performing decontamination should take precautions not  to become
     contaminated (i.e., wear gloves and other protective clothing  as
     required).
Procedure
  1.  Use only mild soap and water to decontaminate the face.
  2.  Exercise special caution to prevent the spread of contamination to eyes,
     ears, nose, or  mouth.
  3.  Avoid the  use of oxidizing agents because of the sensitivity of facial
     skin and to prevent harm to the eyes.
  4.  Take nasal smears to  assess the presence of  nasal contamination.
  5.  Contact medical personnel for assistance in  treating  persons with  hrgh
     levels of  facial contamination or a suspected internal deposition  of
     radioactivity.
                                     '.22-A

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                                                               DARCOM-P  365-1
A.5  EYE, EAR, AND MOUTH DECONTAMINATION
Prereoir: sites
 1.  Obtain assistance of medical  personnel.
 2.  Survey to identify extent and magnitude  of  contamination;
 3.  Collect materials needed for decontamination.
 4.  Document steps and survey results  in the appropriate  log.
Precautions
 1.  Medical treatment takes priority over decontamination.
 2.  Do not spread contamination to clean areas.
 3.  Do not reuse applicators (replace  after  each  time  skin  is  touched).
 4.  Handle all  waste materials as contaminated  waste.
 5.  Stop decontamination procedures if evidence of  skin damage appears  or if
     person complains of soreness  or stinging; contact  medical  personnel for
     assistance.
 6.  Person performing decontamination  should take precautions  not to become
     contaminated (i.e., wear gloves and other protective  clothing as
     required).
Procedure for Eye or Ear Decontaminetion
 1.  Flush with  water.  A fountain can  be prepared by attaching a flexible
     tube to a faucet or water bottle.
 2.  Survey.
 3.  Repeat as necessary.
 4.  If eye becomes irritated or activity cannot be  removed, obtain further
     medical assistance.
 5.  Fluids or agents other than water  should not  be used  unless approved by
     medical personnel.
                                    7.23-A

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DARCOM-P 385-1
Procedure for Mouth Decontamination
 1.  Special Cautions:
     (a)  Under no circumstances should a person with mouth  contamination be
          allowed to eat, drink, chew, or use tobacco until  decontaminated.
     (b)  In no cases shall  oxidizing agents (bleach, potassium permanganate,
          or sodium bisulfite) be used in the mouth because  they will  damage
          the mucous membranes.
 2.  For localized mouth contamination (spot on tongue or tooth), swab with  an
     applicator or cotton sponge.
 3.  For general mouth contamination, flush using tap water  and a flexible
     tube connected to a faucet or water bottle (the fountain method).
 4.  If contamination cannot be effectively removed by flushing, further
     medical assistance should be obtained.
 5.  Bioassay should be initiated for individuals with mouth contamination.

A.6  NASAL DECONTAMINATION
Prerequisites
 1.  Obtain the assistance of medical personnel.
 2.  Survey to identify extent and magnitude of contamination.
 3.  Collect materials needed for decontamination.
 4.  Document steps and survey results in the appropriate log.
Precautions
 1.  Medical treatment takes priority over decontamination.
 2.  Do not spread contamination to clean areas.
 3.  Do not reuse applicators (replace after each time skin is touched).
 4.  Handle all waste materials as contaminated waste.
 5.  Stop decontamination procedures  if evidence of  skin damage  appears  or  if
     person complains of soreness or  stinging; contact medical personnel for
     assistance.
                                    7.24-A

-------
                                                               DARCOM-P  385-1
 6.   Person performing decontamination  should take  precautions  not to become
     contaminated (i.e.,  wear gloves  end other protective  clothing as
     requi red).
Procedure
 1.   V.'nen nasal  contamination is  suspected,  have  the  person  blow nose into
     disposable  tissue.   Survey  used  tissue  and nose.
 2.   Take smears  externally on the  nose and  uoper lip  area using filter  papers
     moistened with water.
 3.   Take smears  inside  each nostril  using  cotton-tipped  applicators  moistened
     with water.
 4.   Gently swab  nasal  passages  using wet cotton  applicators and periodically
     have the person blow nose into tissue.
 5.   If contamination is  not removed, obtain further  medical assistance  in
     performing  nasal  irrigation.
 6.   Bioassay should be  initiated for individuals with nasal contamination.
                                    7.25-A

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                                                    DARCOM-P 385-1
                       APPENDIX B

     EQUIPMENT AND MATERIAL DECONTAMINATION METHODS

Table 7.2.  Contamination Removal Methods
Table 7.3.  Sealing Methods
Table 7.4.  Decontamination Methods for Various Surfaces
                         7.27-B

-------
DARCOM-P  385-1
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                                               7.35-B

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                                               DARCOM-P 385-1
                   APPENDIX C

EQUIPMENT AND MATERIAL DECONTAMINATION PROCEDURES

          C.I  Tape Patches
          C.2  Vacuum Cleaning
          C.3  Wiping or Mopping
          C.4  Water Jets
          C.5  Detergents
          C.6  Complexing Agents
          C.7  Organic Solvents
          C.8  Acids and Acid Mixtures
          C.9  Caustics
                    7.37-c

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DAKCOM-P 385-1
                                  APPENDIX
               EQUIPMENT AND MATERIAL  DECONTAMINATION  PROCEDURES
C.I  TAPE PATCHES
Materials
 1.  Masking, adhesive,  friction,  or duct tape
Procedure
 1.  Place tape over contaminated  area.
 2.  Remove tape and discard as radioactive waste.
 3.  Repeat as long as effective.

C.2  VACUUM CLEANING
Materials
 1.  Conventional wet or dry vacuum cleaners may be used if modified to include
     a high-efficiency particulate air (HEPA) filter on the exhaust.
Procedure
 }.  Use conventional vacuum-cleaning techniques.
 2.  Periodically monitor build-up of radioactivity or dose rate from bag or
     canister during operation.
 3.  Dispose of bag or collection  container as radioactive waste.
 4.  For extensive use,  monitor build-up of dose rate from collection
     container and HEPA filter.

C.3  WIPING OR MOPPING
Materials
 1.  Mop, cloth, or tovel.
                                   7.3S-C

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                                                               DARCO.M-P 385-1

Procedure

 1.  Wipe or wet-mop using a decor-.aminoting agent and hot water.


 2.  Rinse with clean water, damp-mopping.


 3.  Repeat as necessary.



C.4  WATER JETS


Materials


1.   High-pressure, low-volume jet and/or low-pressure jet or spray.


Procedure


 1.  Spray from top to bottom at an angle of 30° to 45°.


 2.  Use high-pressure jets to loosen decontamination.


 3.  Use low-pressure jets or sprays to wash and flush.

                                                                    2
 4.  Determine cleaning rate experimentally or else use 0.5 to 0.9 m /min.



C.5  DETERGENTS


Materials


 1.  Detergent.


Procedure


 1.  Apply full strength or per manufacturer's recommendations.


 2.  Wipe with towel or rag.


 3.  Powered brush may be used.


 4.  Key be applied by a mist applicator, using caution to prevent spread to

     other surfaces.



C.6  COMPLEXING AGENTS


Materials


 1.  Solution containing 3« (by weight) of complexing agent (e.g., EDTA).
                                   7.39-C

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DARCOM-P 385-1
Procedure
 1.   Spray surface with agent.
 2.   Keep moist for 30 minutes.
 3.   Flush with water.
     Note:  Kay be applied to vertical  and overhead surfaces  by adding chemical
     foam (sodium carbonate or aluminum sulfate).

C.7   ORGANIC SOLVENTS
Materials
 i.   Kerosene, paint thinner, or acetone.
Procedure
 1.   Use standard wiping techniques.
 2.   Immerse in solvent bath.
     Caution:  High flammability and  toxic fumes.   The use of acids and
     complexing agents is generally preferable.

C.8  ACIDS AMD ACID MIXTURES
Materials
 1.   Single Acids (1 to 2 nonnality)
     3%-6« sulfuric acid
     9%-l8% hydrochloric acid
     5% oxalic acid
 2.   Acid Mixture
     0.4 liter hydrochloric acid
     90 grams sodium acetate
     4 1iters water
 3.   Other acid mixtures may include  acetic acids, citric acids, acetates,
     citrates.
                                   7.40-C

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                                                               DARCOM-P 385-1
Procedure
 1.  Use dip bath for movable items.
 2.  Leave weethered surfaces in contact with  acid  solution  for 1  hour.
 3.  Allow pipe circulation systems to soak for 2  to 4  hours.
 4.  Flush with water.
 5.  Flush with neutralizing solution.
 6.  Flush with water.
     Caution:   Personnel  hazard, toxic and explosive fumes generated.   Provide
     good ventilation.

C.9  CAUSTICS
Materials
 1.  Lye (sodium hydroxide)
 2.  Calcium hydroxide
 3.  Potassium hydroxide
 4.  Typical solution for removing paint:
     38 liters water
     1.8 kg lye
     2.7 kg boiler compound
     0.34 kg cornstarch
Procedure
 .1   Apply caustic solution to painted surface.
 2.  Keep solution in contact with paint until paint is soft enough to be
     washed off with water.
 3.  Wash off  paint and caustic solution with  water.
 4.  Remove remaining paint with scraper.
     Caution:   Caustics pose personnel burn hazard.
                                  7.41-C

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                                                               L'ARCO.M-P 385-1

           CHAPTER 8.  SELECTION AND DESIGN OF RADIATION FACILITIES
8.1  GENERAL PRINCIPLES   	    8.5

6.2  INITIAL PLANKING PROCESS  	    8.6

     8.2.1  Designs for New Facilities	8.7

            A.  Commander's Responsibility   	    8.8

            B.  Ionizing  Radiation Control Committee   ....    8.8

     8.2.2  Review of Designs for Modifying Existing Facilities   .     .    8.9

8.3  SITE SELECTION	8.10

     8.3.1  Impact of Surrounding Operations Upon the
            Proposed Facility  .                   	    8.10

            A.  Background Radiation    .......    8.10

            B.  Effluents From Facility and Nearby Operations     .     .    8.11

            C.  Fire and  Explosion Hazards	8.11

            D.  Chemicals Spills   ........    8.11

            E.  Access Control	8.11

     8.3.2  Impact of Proposed Facility Upon Surrounding Area     .     .    8.12

            A.  Potential Environmental Releases  .    .     .     .     .    8.12

            B.  Accident  Analysis  	    8.12

            C.  Future Land Use	8.13

            D.  Additional Considerations    	    8.13

     8.3.3  Natural Phenomena  .........    8.14

            A.  Regional  Climate   ........    8.14

            E.  Hydrology	8.14

            C.  Geologic  and Seismic Considerations    ....    8.14

8.4  FACILITY DESIGN	8.15
                                      8.1

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DAJICO>;-? 385-1



     8.4.1  General Considerations in Facility Design



     8.4.2  Building Areas    	



            A.  Radiation Area     ....



            B.  Controlled Area    ....



            C.  Uncontrolled Area  ....



     8.4.3  Work Stations     .....



            A.  Class A Laboratories



            E.  Class B Laboratories



            C.  Class C Laboratories



     8.4.4  Building Materials     ....



            A.  Ease of Decontamination  .



            B.  Corrosion Resistance



            C.  Fire Resistance    ....



     8.4.5  Building Access   .....



8.5  CONTROL  OF EXTERNAL RADIATION



     8.5.1  Shielding Requirements ....



            A.  Integrity     .....



            B.  Materials     	



            C.  Entryways     .....



            D.  Quality Assurance  ....



     8.5.2  Access  Restrictions  for  Radiation  Areas



            A.  Interlocks and Warning Systems



            B.  Guards   	



8.6  CONTROL  OF INTERNAL RADIATION ....



     8.6.1  Containment Devices    ....



            A.  Sealed Sources     ....
8.15



8.16



8.16



8.17



8.18



8.18



8.18



8.19



8.20



8.21



8.21



8.22



8.22



8.22



8.23



8.23



8.23



8.23



8.25



8.25



8.25



8.25



8.26



8.27



8.27



8.27
                                       8.2

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                                                               DARCOK-P 385-1
           ..B.  Hoods	8.27
            C.  Glove Boxes	8.29
            D.  Hot Cells	8.29
     8.6.2  Ventilation Systems    	    8.29
            A.  Ventilation Zones  	    8.30
            B.  Air Flow Patterns	8.30
            C.  Pressure Differentials  	    8.30
            D.  Duct Routing	8.30
            E.  Filtration	8.31
     8.6.3  Sampling and Monitoring Equipment     	    8.32
            A.  Air Samplers and Monitors    .     :     .     .     .     .    8.32
            B.  Radiation Area Monitors	8.32
8.7  FACILITY SUPPORT    	    8.33
     8.7.1  Change Room Facilities	8.33
     8.7.2  Personnel and Property Decontamination Facilities     .     .    8.33
     8.7.3  Water Supply and Sanitary Sewers 	    8.33
REFERENCES	8.34


                                    FIGURES

8.1  Facility Layout	8.16
8.2  Class A Laboratory	8.19
8.3  Class B Laboratory  ..........    8.20
8.4  Class C Laboratory	8.21
                                      8.3

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                                                               DARCO.M-P 3E5-1
            CHAPTER 8.  SELECTION AND DESIGN' OF RADIATION FACILITIES
      Facilities in which radioactive materials are used have specific needs
 that  must be recognized and planned for from the initial design phase through
 the construction and operation of each facility.  The location of the facility
 must  be considered in relation to the work that will  be carried on there.  The
 building must be designed to keep radioactive materials in certain areas while
 still allowing efficient operation.  Finally, equipment must be built in or
 brought in to control external and internal radiation doses to personnel and
 to keep the amount of radioactive material leaving the facility within permis-
 sible 1imits.
      The purposes of this chapter are:  1) to help the Radiation Protection
 Officer (RPO) and the Ionizing Radiation Control Committee (IRCC) judge whether
 a facility is adequate for handling radioactive materials, and 2) to delineate
 wnat  should be considered when a facility is being designed and the rationale
 behind each item.  Because DARCOM and the installation's engineering staff have
 ultimate responsibility for facility design, this chapter is for information
 purposes only.
                        Section 8.1  GENERAL PRINCIPLES

     Safety should be achieved as much as possible through engineered sefe-
guc'-ds rather than aaministrative controls or the use of personnel protective
equipment.  The National Council on Radiation Protection and Measurements
(M,. ••') recommends in Report No. 59 (1978) that a qualified expert be consulted
during the planning and design of new and modified radiation facilities to
ensure the incorporation of proper radiation safety procedures.  Certification
by the American Board of Health Physics or the American Board of Radiology is
evidence of a consultant's qualifications.
     Items that m-j.-t be considered when a new facility is being planned or an
existing structure is being renovated include:
                                      8.5

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DARCOX-P 385-1
 I.  meteorological and hydrologies!  parameters  of  the  site
 2.  facility layout, which should be compatible with  the  establishment of
     contamination areas
 3.  shielding, especially witn respect to floor-loading limits
 4.  ventilation, which should be capable of controlling the  movement of air to
     prevent or minimize the spread of contamination within  the  facility
 5.  types of monitoring equipment needed.
The facility should be arranged to meet the following  objectives:
 1.  keep dose equivalents received by personnel as low as is reasonably
     achievable (ALARA)
 2.  confine radioactive materials accidently released within the facility and
     control releases from the facility to levels below the  concentration
     guides in 10 CFR 20, Appendix B, Table 11,  averaged over 2  hours
 3.  achieve a uniform level of safety through physical and  engineered
     safeguards
 4.  accommodate normal or anticipated changes in mission requirements without
     compromising radiation protection.


                     Section 8.2  INITIAL PLANNING PROCESS

     The terms "facility design," "radiological  design," and "radiological
engineering" are often used interchangeably, although they have different mean-
ings.  Design is the planning and development of a facility as opposed to its
actually construction and operation.  Facility design refers to a plan for a
building or installation as a whole, and thus includes  nonrediological as well
as radiological design features.  Radiological design refers to the  specific
set of design features included because of the planned  presence of radioactiv-
ity or radiation-generating machines.  Radiological requirements should  be made
known to the architect ana/or engineer responsible for  designing a facility as
early as possible, to minimize the cost of incorporating  safety features; it  is
                                      8.6

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                                                               DAHCOM-P 385-1
 less expensive to  reoraw preliminary plans than final blueprints, and less
 expensive to  revise final blueprints thar to rebuild or alter the finished
 facility.
     Radiolopicel  engineer~.r.g refers to the implementation of the radiological
 design  (i.e., the  actual construction).  Radiological engineering requires the
 use of  quality control procedures during construction.  For example, precau-
 tions should  be taken to minimize air pockets in concrete walls used for
 shielding, to sufficiently overlap lend sheets used for shielding, and to
 ensure  that foundations, footings, and pilings have sufficient loadbeering
 capacity so that concrete shield wells do not buckle or crack.  In essence,
 good radiological  engineering ensures that the design criteria are met
 (Kathren and  Selby 1980).
     Review of the radiological hygiene aspects of blueprints, drawings, and
 other documents relating to the design of facilities and devices for generat-
 ing radiation should be coordinated through channels with the DARCOM Field
 Safety  Activity and the U.S. Army Environmental Hygiene Agency (USAEHA).
 Therefore, contact with DARCOK and the agency should be made early in the
 planning process to avoid the necessity of expensive changes in the structural
 design.
 8.2.1   Designs for New Facilities
     When a facility is being designed, all proposed uses and needs of the
 facil ity--both current and projected—must be considered, especially if the
 projected needs will  exceed the current needs.  If possible, the facility
 should  be designed to meet tne maximum needs, because the cost of altering or
 rebuilding may be greater than the cost of overbuilding initially.  The scope
 of work to be performed in the building should be defined in terms of the
 purpose of the work,  the proposed inventories of•radioactive materials, the
 presence of radiation-generating devices, and the expected lifetime of the
 building.
     Many safety features mjst be considered early in the design of a facility.
With few exceptions,  shielding an: facility layout ere difficult to change, and
adequate safety often cannot be ensured in a redesigned or rebuilt facility
                                      8.7

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DARCOM-P 365-1
without high costs and the loss of usable  work  space.   Thus,  future  uses  of
the facility, which may include increased  workloaas,  must  be  considered  so
that shielding, containment, confinement,  and work spaces  can be  designed to
suit those uses.
     A.  Commander's Responsibility.   The  local  commander  is  responsible  for
providing proper facilities for the use of radioactive materials  (AR 385-11).
Therefore, the commander shall provide for the  review and  approval  of all
blueprints, drawings, and other documents  relating to the  design  of facilities
that will contain radioactive materials.   Assistance  in judging the adequacy
of new and renovated facilities may be obtained from  USAEHA and tne DARCOM
Field Safety Activity.
     B.  lonizatino Radiation Control  Committee.  The IRCC should have as part
of its responsibility helping to design safe facilities.  The committee should
include construction or general engineering personnel and  representatives from
maintenance, operations, health, and safety, including the RPO.  The committee
should be informed of all proposed uses for each building, both immediate and
future.  The local commander shall establish an approval process  that guaran-
tees that all safety-related concerns  (both radiological and nonradiological)
have been addressed and adequately resolved.
     (1)  Meinter.ance and Operations Representatives.  Representatives from
maintenance and operations should be consulted because they are usually aware
of the problems associated with various building designs.   They can advise on
whether a design will allow ease of maintenance and repair, which can minimize
work times in radiation areas.
     (2)  Health and Safety Representatives.  The RPO and the other health and
safety representatives snould be responsible for the following:
 1.  reviewing the general layout of the facility, giving particular attention
     to corridors, traffic patterns, radiation areas, change rooms, radiation-
     monitoring sites, and personnel decontamination facilities
 2.  working with the installation's environmental coordinator to prepare  or
     coordinate the preparation of the environmental  impact  statement  (if
     any)
                                      8.8

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                                                              DARCOM-P  385-1
 3.  identifying manuals and standards that deal  with radtojogicel  aspects of
     the fecility design
 4.  ensuring that the ventilation system will  provide the public and site
     personnel with maximum protection against  airborne contamination
 5.  ensuring that maximum practical  control  of liquid, solid,  and gaseous
     wastes is provided, to protect the environment
 6.  verifying that the proposed design and application of hoods, glove boxes,
     and shielded cells ensures ease  of decontamination and remote operation,
     to reduce occupational exposures
 7.  ensuring that the thickness of all shielding meets design  criteria, and
     coordinating shielding calculations and design to keep radiation doses
     ALARA
 8.  ensuring that needs for sampling and monitoring instrumentation have been
     identified and that the instrumentation being provided meets the latest
     occupational and environmental standards,  can be installed properly, and
     is capable of obtaining representative samples
 9.  ensuring that radiological safeguards and  safety systems are adequately
     protected from fires, floods, and other similar accidents, and are
     fail-safe
10.  assessing the adequacy of facilities for receiving, storing, and packaging
     any radioactive wastes that may  be produced during the operation of the
     building.
8.2.2  Review of Designs for Modifying Existing Facilities
     How extensively a facility is being modified influences the extent of the
design review needed.   Major modifications, such as extensive renovation of a
radioactively contaminated facility or preparation of a facility that has never
before housed radioactive materials,  may require application of ell steps
involved in the design of a new facility and may therefore require the same
attention  from the members of the IRCC.  The RPO, or the eopropriate health
and safety representative, has the following additional responsibilities
whenever an existing radiation facility is being upgraded:
                                      8.9

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DARCOM-P 385-1

 I.  If the building previously contained radioactive  material,  evaluate the
     modification plans to ensure that radiation dose  equivalents  received by
     construction workers during the renovation are kept ALARA.   (Consider
     removing radioactive sources end decontaminating  the facility.)
 2.  Evaluate the impact of the modification on existing safety  systems, such
     as air filters and ventilation systems.
 3.  Review the d'esign of any structures needed to contain radioactive
     materials (e.g., greenhouses and special waste containers).
 4.  Approve all modifications.
                          Section 8.3  SITE SELECTION

     The  initial step in selecting the site for a radiation facility is to
 establish the reouirements of the facility and the interrelations between the
 facility  and its environment.  Proposed sites and the area surrounding each
 should be reviewed for location and for distances from air, ground, and water
 traffic,  pipelines, and fixed manufacturing, processing, and storage
 facilities.
 8.3.1  Impact of Surrounding Operations Upon the Proposed Facility
     The  level  of background radiation at a proposed site can affect some
 operations and  should be considered during site selection.  Other external
 factors effecting site selection are the location of other facilities, the
 potential for fires, explosions, and chemical spills, and any need  for
 restricted access.
     A.   Background Radiation.  Background radiation is an important considera-
 tion for  facilities that will house laboratory counting instruments, which  are
 extremely sensitive to radiation.   Fluctuations  in  the  level of  background
 radiation can effect the instrument readings, and high  background  radiation
 rates, even  if  they are constant,  increase the lower limit of detection  for
 these  instruments.
                                      8.10

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                                                              DARCON-P  385-1
     Background radiation levels can be increased  by  either  natural  or marr-
maoe sources.  One natural  source of increased  background  radiation  is certain
types of rocks that have a  fixed raciation.   The evolution of  radon  gas  from
rocks and soil also raises  the concentration of radioactive  material  in  air
and thus results in more surface contamination  and higher  dose  rates.   Man-
made causes of increased background radiation levels  include nuclear power
reactors and mining and milling operations.   Uranium  mining  is  an  obvious
cause; however, phosphate mining and even coal  mining are  also  sources of
background radiation.
     Trie extent to which radiation background levels  fluctuate  because of
these sources is small  and  under ordinary circumstances  does not  present a
significant radiation  hazard to personnel.   However,  a radiation  hazard may
occur in submerged or  underground facilities, especially if  the  air  flow rates
a re 1 ow.
     B.   Effluents From Facility and Nearby  Operations.   A facility  should be
designed so that its air intakes are not likely to draw in its  own exhaust
materials.   As a general rule, air intakes  should  be  at the  upwind end of  the
facility and exhaust vents  should be at the  downwind  end,  with  the prevailing
wind direction used as  a guide.  In addition, air  intakes  should be  at least
155 meters away from the exhaust vent of any other facility  that is  venting
radioactive material or other toxic or hazardous materials.
     C.   Fire end Explosion Hazards.  Operations that might  present  fire and
explosion hazards include petroleum refineries  and storage facilities, docking
facilities (for example, for oil tankers),  and  chemical-manufacturing plants.
Also, military depots may be sites of storage for  explosive  compounds.  Radia-
tion facilities should  be located at a safe  distance  from such hazards.
     D.   Chemical  Spills.  The manufacture,  storage,  and transportation of
chemicals lead to the potential for chemical spills or releases.   The release
of toxic gas may require that a facility be  evacuated promptly.   However,   in
some facilities such as nuclear reactors, operators cannot be  evacuated
immediate!}.   In such cases, protection must be prcviced for the workers.
     E.   Access Control. Access to a facility  may be restricted for either
radiation safety or national security reasons.   Access control  for national
                                     8.11

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     M-P 3S5-1
security purposes is beyond the  scope  of  this  manual.  Whenever  a  high-
radiation area TS not mechanically secured  to  prevent unauthorized  entry,  a
guard must be posted (DARCOM-R 385-25).   Physical  safeguards  that  are  appro-
priate for the hazard or security classification  must be  used.
8.3.2  Impact of Proposed Facility Upon  Surrounding  Area
     The use of radioactive materials  at  a  facility  may  increase the  level  of
background radiation if any materials  used  outside of sealed  containers  are
released to the environment.   The releases  may be of two  types:   routine
low-level releases and accidental releases  that could be  of any  magnitude.   The
possibility of such releases  influences  the selection of  a facility site.  The
anticipated use of the land around a proposed  radiation  facility should  also be
considered in site selection.
     A.  Potential Environmental  Releases.   Routine  releases  usually  enter the
air from hood vents and enter sanitary sewage  systems via floor  and sink
drains.  Radioactive material  may also be transported to  the  environment on
the clothing of personnel and can be tracked about extensively if it  gets on
their shoes.  Facilities in which radioactive  materials  are  used should  be
located downwind from major metropolitan areas and in flat or gently rolling
terrain, so that any radioactive material accidentally  released  into the air is
dispersed rapidly and evenly,  with minimal  impact.  In  addition, engineered
safeguards should be provided to prevent or at least limit the release of
radioactive materials to the environment.  Such safeguards are discussed later
in the chapter.
     B.  Accident Analysis.  The potential  for accidents should  be analyzed
before any accident occurs.  The RPO and individuals familiar with ventilating
systems should review the proposed levels of radioactivity in each laboratory.
Accidents that could result in the release of radioactive materials should
then be analyzed.  This analysis can be detailed, involving determination of
the possible causes, probabilities, and impacts of an accident;  or it can be
as simple as assuming that the largest amount of material that might be
unsealed at any time is available for release (see Chapter il).
                                     8.12

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                                                              DARCOM-P  385-1
     Accident analysis is extremely important  for  major, facil ities.   It  can
help ensure that engineered safeguards  are  provided  to  prevent or minimize
radiation exposures of personnel  and the  public.   If an accident  analysis  is
performed early in the design process,  safeguards  may be  suggested that
otherwise would have been omitted.
     C.  Future Land Use.  Sources  of information  on projected population
growth and proposed land uses should be consulted.   County  engineers  can
provide information on public roads and traffic  volumes;  local government
councils may have information on  population growth,  proposed  new  industries,  or
future transportation routes; and zoning  boards  are  sources of information on
land use controls.  The increase  in the local  population  brmght  about by  the
construction and use.of the proposed facility  should also be  considered, as  it
may not have been included in the projections  of the state  and local  agencies
just mentioned.
     D.  Additional Considerations.  Before a  particular site is  selected, the
following topics should be considered:
 1.  personnel  traffic routes and their relation to  the flow  patterns for
     exhaust air where accidental or routine releases of radioactive  material
     could occur (radioactive material  should  not  be vented to high-traffic
     areas)
 2.  the relationship between the exhaust and  air  supply systems  of  various
     facilities (radioactive material  should not be  vented  where  it  is  likely
     to be drawn into other buildings or  back  into the  building  it came  from)
 3.  the impact of additional radioactive waste  on waste removal  systems (e.g.,
     consider stress on sewer systems that  may contain  radioactive material,
     on retention or diversion systems, and on systems  that handle liquid
     waste containing high levels of radioactivity)
 4.  the availability of emergency  systems  (fire,  ambulance,  and  radiological-
     emergency  response teams)
 5.  the ability to simultaneously  evacuate ell  neighboring facilities
                                     8.13

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DARCOM-P 365-1
 6.  the need for special transportation capabilities  (railroad  spurs,  rigs
     for moving heavy material)
 7.  the impact of future modifications.
8.3.3  Natural Phenomena
     Facilities should be designed to withstand the influences  of na*  ral
phenomena.  This requirement can be relaxed for facilities  located where the
only natural phenomena likely to occur are those that  can be  accurately
forecasted, such as hurricanes and floods.  In these cases,  adequate warning
time for securing materials anc evacuating personnel can be  provided.   Other
phenomena,  such as earthquakes, volcanic eruptions, and floods  caused by dam
failure, cannot be adequately forecasted and may occur with  little or no
warning.
     A.   Regional Climate.  Meterological conditions  that may  affect a facil-
ity include hurricanes, tornadoes, water spouts, thunderstorms,  lightning,
hail, and high levels of air pollution.
     Data on severe weather phenomena should be based on standard meteoro-
logical records from a nearby National Weather Service station  or from
military or other stations that are recognized as standard installations and
that have kept records for a long time.
     B.  Hydrology.  The hydrology of a site should be reviewed, especially
if a facility will house large quantities of special nuclear materials
(plutonium  or uranium enriched in isotope 233 or 235, or any material
artificially enriched by either isotope).  The hydrologic characteristics of
streams, lakes, shore regions, and existing or proposed water control
structures  (e.g., dams and irrigation ditches) should be considered as  they
relate to potential flooding of the structure.  The hydrology of both  surface
water and ground water should also be considered as it  relates  to the  possible
contamination of these waters by activities within  the  facility.
     C.  Geologic and Sersmic Considerations.  Ideally,  the site  should be  in a
geologically stable area—one low in reismicity, free of active  faults, under-
lain by competent foundation materials, and free from the adverse effects of
other geologic hazards.
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                                                              DARCOM-P 385-1
                         Section 8.4  FACILITY  DESIGN

     A properly designed facility can lead to reduced  radiation  doses  to
personnel through the establishment of designated  areas  for the  use  of
radioactive materials, and of designated  types  of  laboratories within  these
areas.  The materials used in the construction  of  a  facility and the ease  of
access to areas within the facility also  affect radiation  safety for
personnel.
8.4.1  General Considerations in Facility Design
     The layout of rooms, corridors, entrances, exits, ventilation systems,  and
other utilities in a building should be designed to  meet the following
objectives:
 1.  Keep the dose equivalent received by personnel  ALARA.
 2.  Confine radioactive materials accidentally released within  the  facility
     and control any releases from the facility so that  they remain  below  the
     concentration guides in 10 CFR 20, sections 20.106  and 20.303.
 3.  Accommodate routine programs or anticipated program changes without
     compromising radiation protection.
     The flow of people and materials in  a facility  is a function of building
design.   One design, shown in Figure 8.1, has a central  service  corridor for
equipment,  piping, and waste handling. Laboratories on  both sides  open  to both
the central corridor and the outer corridors, with offices located  between the
outer corridors and the outside of the building.  The advantages of  this
aesign are  that it allows for two exits from each  laboratory,  permits  easy
access to utilities for the laboratories, and allows radioactive materials to
be transferred without effecting the clean areas of  the  facility.  An
alternate design might have offices located in  one part  of the building  and
laboratories in another, so that only laboratory personnel  need  enter  the
laboratory  areas.
                                     8.15

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DARCOM-? 365-1
X N N
/
LAB
L OFFICES 1
\L_NL_NL_ V^_N 	 \
CORRIDOR
\/
LAB
\ /
LAB
J ,J

LAB
SERVICE CORRIDOR
LAB
\
LAB
/ \
\
LAB
/ \
\
LAB
                              CORRIDOR
             Frrrrrrrr
/
                       FIGURE  8.1.  Facility Layout
8.4.2  Building Areas
     Facilities are generally divided into  a series of sequential  areas that
are based  upon the presence of radiation or radioactive materials  and  are
designed to control personnel exposure to radiation.   The three types  of
areas—radiation,  controlled, and uncontrolled—are described below.
     A.  Radiation Area.  Radiation areas include three subclassifications:
radiation  areas, high-radiation areas, and  airborne-radioactivity  areas.  A
radiation  area is  defined in 10 CFR 20 as any area accessible to personnel in
which radiation levels could result in a major portion of the body receiving a
dose-equivalent rate in excess of 5 mrem in any 1 hour or 100 mrem in  any 5
consecutive days.   For practical purposes,  AR 40-14 defines this as any area
in which the dose-equivalent rate is greater than 2 mrem/hr but less than
100 mrem/hr.  A high-radiation area is any  area accessible to personnel in
which radiation levels could result in a major portion of the body receiving a
dose equivalent in excess of 100 mrem in any 1 hour.   All radiation areas must
be marked  and posted as described in 10 CFR 20.20.  An airborne-radioactivitv
     is any room,  enclosure,  or operating  area where the  concentration  of
airborne radioactivity exceeds the amounts specified in 10 CFR 20, Apper.dix B,
Table I, Column 1  or where the concentration, when averaged over the number of
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                                                              DARCOM-P  385-1
hours in any week an incividual works in the  area,  will  exceed  25* of the
amounts specified in 10 CFR 20, Appendix B, Table  I,  Column  1.
     To ensure that regulatory and administrative  limits  are not  exceeded,
consideration should be given during facility design  to  the  establishment of
radiation areas at:
 1.  any location where unsealed (unencapsulated)  radioactive materials  will
     be stored, handled, or processed
 2.  any e^ea containing a radiation-generating  oevice
 3.  any routinely occupied area where an individual  would  be expected to
     receive more than 500 mrem in 1 year
 4.  an.v area, regardless of the expected occupancy,  there  the  anticipated
     dose-equivalent rate exceeds 2 mrem/hr
 5.  any routinely occupied area where the concentration  of airborne radio-
     active materials may exceed 25% of the values  presented in 10 CFR 20,
     Appendix B, Table I, Column 1
 6.  any area, regardless of the occupancy, where  the concentration of air-
     borne radioactive materials may exceed the  values  presented  in 10 CRF 20,
     Appendix B, Table I, Column 1.
     Radiation areas should be remote from offices, lunchrooms, and conference
rooms, to preclude the exposure of support personnel  (e.g.,  secretaries and
clerks).  Persons entering a radiation area should  pass  through a controlled
area.  To keep nonradiation workers out of radiation areas  during the normal
course of their work, separate corridors should  be  provided.
     B.  Controlled Area.  A controlled area  is  any area to which access is
controlled and in which occupancy and working conditions are controlled for
the purpose of protecting personnel against exposure to radiation.  Such areas
include:
 1.  any area normally free of contamination  that  is adjacent to a radiation
     area and thet may become contaminated through accidental spreads o-
     releases from the radiation area
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DARCOM-P 3S5-1
 2.  any area that may occasionally contain radioactive material  because of
     the transportation of radionuclides between radiation areas  or the
     maintenance of contaminated process equipment that cannot be entirely
     placed inside a radiation area
 3.  any area where the anticipated dose-equivalent rate exceeds  0.2 mrem/hr
     but is less than 2 mrem/hr
 4.  any area where the concentration of airborne radioactive materials may
     exceed 50% of the values presented in 10 CFF; 20,  Appendix E, Table II,
     Column 1.
     C.  Uncontrolled A-ea.  An uncontrolled area is any area where direct
 radiation exposure is not necessary or anticipated in the performance of a
 job.  These areas include "cold" laboratories (those containing no radioac-
 tivity), offices, lunchrooms, conference rooms, and reception areas.  The
 traffic patterns in a building should keep radioactive materials  from being
 brought into  uncontrolled areas for any reason (such as by delivery per-
 sonnel).  Further, the building should be designed so that the dose-equivalent
 rate in uncontrolled areas does not exceed 0.2 mrem/hr.
 8.4.3  Work Stations
     Work stations are subdivisions of a radiation area.  One method of desig-
 nating work stations is to define three classes of laboratories,  A, B, and C,
 which depend  upon the radiotoxicity, dispersibility, and total quantity of
 unsealed radioactive materials to be used.  (See Chapter 1, Section 1.6.2, for
 definitions of the levels of dispersibility, and Chapter 1, Table  1.10, for
 groupings of  radionuclides by degree of radiotoxicity.)
     A.  Class A Laboratories.  Class A laboratories are specially designed
 end equipped  for the safe handling of 1) large quantities of  highly radiotoxic
 materials (groups VI through VIII in Table 1.10) in any dispersible form and
 2) large quantities of moderately radiotoxic materials  (Groups III through V)
 in highly dispersible form.
     Each Class A laboratory should be wholly within a  radiation area  and
 should be separated from uncontrolled areas by at least two confinement
 barriers (see Figure 8.2).  Within a Class A laboratory, a fume  hood should be
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                                                              DARCOM-P 385-1
                      UNCONTROLLED AREA
                      CONTROLLED AREA
                          RADIATION AREA

                              HOOD
                                 CONFINED WORK AREA,
                                   CLOVE BOXFS,
                                    HOODS etc.
                           SERVICE AREA
                        FIGURE 8.2.  Class A Labora'tory

used for work involving dispersible material, and sealed glove boxes, hot
cells, or similar devices should be used for work involving readily or highly
dispersible materials.   Class A laboratories should have access to a clothing
change room through which personnel pass before entering an uncontrolled
area.
     The air of a Class A laboratory should be exhausted through two stages of
high-efficiency paniculate air (HEPA) filters that are testable using a
dioctylphthalete (OOP)  mist.  (Because OOP is a suspected carcinogen, it
should be used with care.)  The air of hoods, glove boxes, or other sealed
enclosures where readily and highly dispersible materials ere used should be
exhausted through three stages of HEPA filters, at least two of which must be
OOP-testable.  Tne use  of gaseous materials (wet operations) may cause early
failure of HEPA filters.  Therefore, if these materials are used, HEPA filters
may need to be replaced frequently, air flow monitors should be used, and
additional filtration devices may be needed.
     B.  Class B Laboratories.  Class B laboratories are designed for the
handling of 1) large quantities of minimally radiotoxic materials (Groups I
and II in Table 1.10) or 2) moderate quantities of moderately or highly
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DARCOM-P 385-1
radiotoxic materials (Groups II I--through VIII).   The materials  may range from
dispersible to highly dispersible.
     Each Class B laboratory should be separated from uncontrolled areas by at
least two confinement barriers (see Figure 8.3).   A glove box or other enclo-
sure should be used for work with highly raciotoxic or highly dispersiole
materials.  Each laboratory should have at least one fume hood.
     The air of a Class B laboratory should be exhausted through at least two
OOP-testable HEPA'filters that are in series.   The exhaust system for hoods,
glove boxes, or other enclosures should contain two stages of OOP-testable HEPA
f i 1 ters.
UNCONTROLLED AREA

cor>


TROU-ED AREA


RADIATION AREA
— HOOD

CLOVE
BOX




                        FIGURE 8.3.  Class B Laboratory

     C.  Class C Laboratories.  Class C laboratories are designed for work
involving simple chemical processes and minimal  quantities of radioactive
material.  Materials of low and moderate radiotoxicity (Groups I through V in
Table 1.10) may be present in forms that are dispersible or of limited
cispersibility.
     Each Class C laboratory should be separated from uncontrolled areas by at
least one confinement barrier, which may be the laboratory wall (see Fig-
ure 8.4).  At least one hood should be provided in each laboratory.  The exhaust
system should contain at least a single-stage OOP-testable HEPA filter.
                                     8.20

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                                                               DARCON-P 385-1
UNC
;ON'TROLLED AREA







CONTROLLED AREA

RADIATION

El-'TRY

ARLA
HOOD








                        FIGURE 8.4.  Class C Laboratory

8.4.4  Building Materials
     The building materials used in a radiation facility should be easy to
decontaminate, extremely durable, corrosion resistant,  and fire resistant.
Unfortunately, very few materials combine all  of t^.ese  characteristics.
     A.  Ease of Decontamination.  The building materials chosen should be
nonporous and should have few, if any, cracks.   They should be readily remov-
able if contaminated, and chemically inert to  reduce the likelihood that con-
tamination would become chemically bonded to the materials.  (See Chapter 7
for details on the ease of decontaminating various materials.)
     (1)  Flooring.   Flooring materials should  be chosen based on price; avail-
ability; ease of installation, service, and maintenance; chemical inertness;
end any special  requirements imposed by the use of radioactive materials.
Porous materials such as concrete and wood are  not acceptable by themselves;
they must be covered by ether, removable materials to facilitate decontamina-
tion in the event of an accident.  Examples of  acceptable covering materials
include sheet flooring (such as vinyl  flooring) or poured vinyl or epoxy floor
covering.   The floor covering should be sealed  and waxed regularly.
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DARCOM-P 385-1
     (2)  Walls and Partitions.  Walls and partitions 'should be protected by
coatings that are hard, smooth, and er.3y to clear,.   If  extensive contamination
is possible, then strippable coatings should be used.   These form an effective
;eal over porous wall, ceiling, or floor materials  and  are easily stripped off
or removed when the surface must be decontaminated.
     (3)  Bench TODS and Laboratory Equipment.   Laboratory benches with syn-
thetic or plastic tops are now available.  Many of  these tops are quite imperme-
able and durable;•consul t manufacturers' literature for details.  Laboratory
equipment can be tested for susceptibility to contamination and ease of decon-
tamination, as described by Fitzgerald (1969).   In  general, furniture in
laboratories where low and intermediate levels  of radiation are used should be
of high-quality, impermeable materials.
     B.  Corrosion Resistance.  Bench tops, hoods,  walls, and floors should be
corrosion resistant because the pitted surfaces caused by corrosion are
difficult to decontaminate.
     C.  Fire Resistance.  Laboratory facilities should be fire resistant.
Where a fire could result in the dispersal of radioactive materials, exits and
a means of closing the facility to prevent the spread of radioactive materials
should be provided.  Fire extinguishers should be located throughout each
facility, and showers and fire extinguishers should be provided in  laboratories
where flammable chemicals are  used.
8.4.5  Building Access
     Consideration should be given to pathways for moving radioactive materials
in and out of buildings and laboratories.  Examples of  items that  should  be
considered are:
  1.  doorways - Because radioactive sources are usually  integrated  with  large,
     heavy shielding, motorized carts,  trucks, or fork  lifts may  be needed  to
     move them.  Doorways and  hallways  leading to exits  should  be  large  enough
     to allow passage of these machines.
  2.  ramps - Sealed, shielded  radioactive  sources can weigh  tons  and may exceed
     the lifting capacity of freight elevators.  Gently  sloping ramps  should  be
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                                                               DARCOM-P 365-1
     provided between floor levels so that such sources  can be transported by
     fork  •ifts from one building level  to another.
 3.  ceiling openings - Ceilings can be  designed so  that the roof is  easily
     dismantled, providing an opening large enough  for a crane to lift out a
     shielded source or any large heavy  object.
                  Section 8.5  CONTROL OF EXTERNAL RADIATION

     Dose rates to personnel from radioactive materials can be greatly reduced
by the placement of attenuating or shielding materials between personnel  and
the radiation source.  The shielding materials can be designed into the
building structure or they can be separate from the building.    Shielding may
be required to protect personnel  from radiation emitted from open, unsealed
radioactive materials and from radiation-generating devices.  External dose
rates are also controlled by restricting access to radiation areas through the
use of interlocks, warning systems, and guards.  (See Chapter 6 for details on
the control and reduction of external exposure.)
8.5.1  Shielding Requirements
     Shielding is required wherever the anticipated dose-equivalent rate will
exceed 2.0 mrem/hr.   The shielding should reduce the dose-equivalent rate to
0.2 mrem/hr or less.
     A.  Integrity.   Shielding must be designed so that the degree of
protection is constant from all angles of approach.  The simplest method of
achieving uniform protection is to surround a source with a uniform shield.  In
practice, however, a  shield is usually penetrated by cooling pipes, electrical
power and signal cables, rotating shafts, and removable plugs or covers, and
special considerations must be made for these penetrations in the shield.
Design features such  as shadow shields, baffles, and offsets can help ensure
adequate protection.
     B.  MaterieU.   The choice of shielding material depends upon factors such
as cost and the desired thickness and mess of the shield.  However, all of the
following should be  considered whenever shielding material is being selected:
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DARCOM-P 385-1
 1.  attenuation characteristics - Different shielding  materials  have
     different abilities to attenuate photons, neutrons,  and  beta particles.
 2.  structural integrity - The materiel selected must  be structurally stable.
 3.  nonflammability - The shielding material should be fire  resistant or
     noncombustible and should not release toxic gases  or smoke when heated.
 4.  confinement capability - The shielding material may  have to  contain gases,
     solids, and liquids in the shielded enclosure.
 Shielding materials commonly used for various types  of  radiation  are described
 below.
     (1)  Shielding for Ions and Electrons.  Virtually  any material  can be used
 as  shielding for ion and electron sources as long as the shield  is thicker than
 the  range of the particles.  Bremsstrahlung radiation may be  produced if
 shielding materials with a high Z number (atomic number), such as lead or iron,
 are  used.  The likelihood of bremsstrahlung radiation can be  reduced by the
 use  of  low-2 shielding materials, such as plastics.   If bremsstrahlung
 radiation is produced, it can be attenuated by lea'd, iron, or any material
 that shields against x and gamma rays (see Chapter  1).
     (2)  Shielding for X- and Gamma-Ray Sources.   Common shielding materials
 for photon sources are lead and iron.  Depleted uranium and tungsten are
 expensive materials for shielding but they can be used if a relatively thin
 shield  is required.  Concrete and water can be used if the thickness of the
 shield  is of no consequence.
     (3)  Shielding for Neutrons.  Shielding for thermal  (slow,' or  low-energy)
 neutrons is provided by thin layers of materials that have a high cross section
 for capture, for example, boron or cadmium.  A disadvantage of cadmium is that,
 after a neutron is captured, the material emits high-energy gamma rays for
 which shielding must also be provided.
     Fast neutrons are not easily shielded.   In addition, sources of fast
 neutrons are commonly also sources of gamma  rays; the shielding material must
 therefore be able to shield against both the photons and  the neu'.rons.  Shield-
 ing of  fast neutrons is generally a two-step process.  First, hydrogen-
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                                                              DARCOM-F 385-1
containing materials  such  as  plastic, water, of concrete are used to moderate,
or reduce, the neutron  energies  to  thermal levels by elastic scatter.  If :he
shielding material  contains  high-Z  elements, such as lead or iron, then the
neutrons may lose their energy through  inelastic collisions.  The thermal
neutrons may then be  captured, as described above, by boron, cadmium, or  (to a
lesser extent) the  hydrogen  in water.
     C.   Entryweys.   Wherever possible, entryways should consist of a
labyrinth, or passage with turnings, that  scatters radiation twice before it
hits a door.  This  scattering reduces the  amount of radulion reaching the
door, with two positive results:  first, the likelihood o:*  radiation streaming
around the door is  lowered;  and  second, the shielding requirements for the door
are reduced and the weight of the door  is  thus lowered.  Labyrinths can reduce
the shielding requirements for a door to negligible levels.
     D.   Quality Assurance.   Following  the construction of  any  shield, the
shield must be tested for  uniformity.   In  concrete, for example, voids may
occur or the aggregate  may settle,  making  the shielding characteristics uneven
and unacceptable.  Special scrutiny should be given to all  penetrations and  to
the crevices between  concrete blocks, if they are used.
8.5.2  Access Restrictions for Radiation Areas
     Access to radiation areas should be restricted whenever the dose-
equivalent rate exceeds the  levels  that define a  radiation  area (see Sec-
tion 8.4.2), and shall  be  restricted whenever the dose-equivalent  rate exceeds
the level thci defines  a high-radiation area.  Requirements for access restric-
tions are defined in  10 CFR  20.203. Access may  be  restricted by  interlocks
and warning systems or  by  guards.
     A.   Interlocks and Warning  Systems.   An  interlock  is  an electromechanical
device such as a switch that causes a radiation-generating  device  to  stop
producing radiation if  the access barrier  to  the  device  is  violated.   Examples
of interlocks include:
 1.  door interlocks  -  These interlocks turn  off  the  radiation-generating
     device if the  door to the high-radiation area  is  opened; they also  prevent
     operation of the device until  the  door  is  closed.
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DARCOM-P 385-1
 2.  device-mounted switches -  A switch  integrated with e timer  is mounted  on
     the radiation-generating device.  To  operate the device, the ooerator
     must enter the room, turn  on the  switch  (and timer), leave  the  room, and
     close the door before starting  the  device  from outsiae  tne  room.   The
     purpose of this type of switch  is to  ensure that the operator enters the
     room before every procedure is  begun  and  instructs ell  personnel  to  leave
     the room.  The timer allows sufficient  time for ell these  steps  to be
     performed without rushing.
                 »
 3.  emergency shutoff or SCRAK switches - These switches are located  through-
     out the room containing tne radiation-generating device.   Their  purpose
     is to allow personnel inadvertently left  in the high-radiation  area  to
     shut off the device or prevent  it from  starting up.  These  switches  must
     be reset before the device can  be operated.
     Warning systems may consist of  lights or  alarms or both, as follows:
 1.  lights - Rotating red warning lights  (the  kind used on  emergency
     vehicles) are located near eye  level  and  are bright enough  to  be seen
     anywhere in the exposure room even  if not  viewed directly.   The  lights
     should be on for 15 seconds before  an irradiation  starts and during  the
     entire irradiation.
 2.  alarms - Warning alarms sound for 15  seconds before an  irradiation can
     start.  When irradiation is started after the  15-second delay,  lights
     remain on and audible alarms stop.
     All interlocks and alarm systems  shall  be fail-safe  so  that a  radiation-
generating device cannot be operated if  the  warning  systems  or  interlocks are
inoperable.  Signs describing the systems  and  how  they  are  used should be
posted near each interlock or warning  system.
     B.  Guards.  Security guards can  prevent  unauthorized  personnel  from
entering radiation areas by cnecking the credentials  of each individual who
desires entry.  Security guards are  necessary  when  electrical  or mechanical
devices for restricting access have  been inactivated  (for  repair or testing),
tne radioactive material is at a temporary location,  or national security
requires the use of guards.
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                                                              DARCOM-P  385-1
                  Section 8.6 •CONTROL OF INTERNAL  RADIATI-QN

     Internal radiation is controlled by the use  of 1)  containment devices,
which prevent radioactive materials from entering work  areas where they might
be inhaled or ingested by personnel; 2) ventilation systems, which remove
radioactive materials from the air of work areas  to ensure clean breathing air;
and 3) air-sampling and air-monitoring systems, which have alarms to notify
personnel if concentrations of radioactive materials exceed permissible limits.
(See Chapter 5 for details on the control of internal exposure.)
8.6.1  Containment Devices
     The spread of radioactive materials can be  kept to a minimum by the use of
seeled sources and containment devices such as hoods, glove boxes, and hot
cells.
     A.  Sealed Sources.  A sealed (or encapsulated) source is defined as a
radioactive source sealed in a container that has a banded cover.  The con-
tainers are designed not to rupture and thus to  prevent dispersion of the
radioactive material under normal operating conditions  and following minor
accidents, such as a container inadvertently being  dropped.  The integrity of
sealed sources should be tested as described in  Chapter 4.
     B.  Hoods.   Open-face or fume hoods should  be  designed and located to
provide constant air flow into the hood.  The velocity of the air flowing into
the hood (the face velocity) must be sufficient  to  ensure that no contamination
enters the room.  For conventional hoods, a face  velocity of 46 ~ 8 linear
meters/min meets this criterion.   Supplied-air hoods and National Cancer Insti-
tute hoods have other criteria;  consult the manufacturer's literature for
details concerning a specific hood.
     Hoods should be illuminated with lights that can be serviced from outside
the hood.  Outlets for gas, air,  and water should be located along the back or
sides of the hood and should be  controlled through  knobs located outside the
hood.   Electrical outlets should be on the outside  of the hood.
     Each hood should be strong  enough to support all necessary shielding.
which should attenuate radiation in all directions.  The air from each hood
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DARCOM-P 385-1
should be exhausted through HEPA filters of a type appropriate  for tne labora-
tory classification (see Section 8.4.3).  In a"!  cases,  a  prefilter shoulc  De
placed ahead of the HEPA filters.  The exhaust ducts  should  be  designed to
allow in-place testing of tne filter systems.  In addition,  pressure taps
should be provided to allow measurement of the pressure  drop across the
filters.  The filters should be located to allow  rapid,  clean servicing with
little danger of the workplace being contaminated.
     The following general rules have been established for the  design of hoods
for work with radioactive and chemically toxic materials (Industrial Ventila-
tion 1980); they are applicable for glove boxes and hot  cells as well.
 1.  Operations in' which radioactive materials are handled should, as often as
     possible, be performed in enclosed areas to  prevent the contamination  of
     large air volumes.
 2.  High-velocity cross-drafts should be avoided because  they may increase
     contamination and dust loading.
 3.  The volume of air withdrawn from the hood must be larger than the volume
     of contaminated gases, fumes, or dusts created in the hood.
 4.  If possible, operations requiring large amounts of wet digestion, volatil-
     ized acid, or solvent treatment should be confined to one group of hoods,
     and dry materiel should be handleo in others.
 5.  Whenever possible,  radioactive aerosols should be removed by filtration.
     The filters should be as close to the hood as practical to prevent unneces-
     sary contamination of equipment and ductwork.
 6.  The value or accountability of the material  used in a hood may  require
     that the hood be designed so that even the smallest chips ar*d turnings
     can be reclaimed.
 7.  A supply of coolant inside the hood may be needed, depending on  the pyro-
     phoric nature of the contaminant (its ability to ignite spontaneously).
 8.  Hoods and duct systems should be designed to be easily accessible for
     decontamination, and should be constructed of materials that  are  easily
     decontaminated.  For this reason, stainless  steel  is frequently  used  for
     the metal parts of hoods.
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                                                              DARCOM-P  385-1
 9.  The hood fan should be located close to the  release  point from the
     building so that ductwork witnin the building  is  under  a negative
     pressure.
     C.  Glove Boxes.  Glove boxes can be designed  to  function as  a primary
containment to minimize the potential for release of radioactive materials.
Tne use of glove boxes minimizes contaminated-air volumes and simplifies a'ir
treatment problems.   Glove boxes should be designed to operate at  a negative
pressure (1.8 = 0.64 cm water gauge pressure) with  respect to the  room in which
they are located.  They should be equipped with differential gauoes to measure
the pressure drop and with control devices to prevent  excessive  vacuum or
pressure build-up.   Penetrations in the glove box (e.g.,  conduits, ports,
ducts, and windows)  should be sealed to prevent the release  of  radioactive
materials.
     D.  Hot Cells.   Hot cells are specialized rooms in which  large quantities
of radioactive materials are used.  The cells are normally fitted  with remote
manipulators, which  allow the manipulation of nuclides that  emit gamma rays and
high-energy beta particles without personnel receiving excessive radiation
doses to the hands,  wrists, and forearms.  Hot cells are maintained under
negative pressure to minimize the spread of radioactivity in the event of a
leak.  The exhaust should be filtered through two HEPA and charcoal filters.
8.6.2  Ventilation Systems
     Ventilation systems are an essential part of a building's  safety features.
Consequently, they should be designed to complement the building layout and
should remain functional or fail-safe during all  operations  and  all credible
accidents.
     The ventilation system must confine airborne radioactive  materials within
the appropriate areas of the building.  It should be capable of removing from
routinely occupied areas any airborne radioactive materials  resulting from
normal or accident conditions.   Further, the ventilation system should be
designed to clear all normal or accidentally generated effluents from the  air
before the air is released to the environment.
                                     8.29

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DARCOM-P 385-1
     A.  Ventilation Zones.  The ventilation system should include physically
 separate ventilation zones to prevent cross-contamination of air.   Ventilation
 Zone I  should correspond to the confinement portion of the radiation area
 (i.e.,  hoods, glove boxes, and hot cells).  Ventilation Zone II should corres-
 pond to the remainder of the radiation area and to controlled areas.  Ventila-
 tion Zone  III should correspond to uncontrolled areas.  Ventilation Z:'ne III
 is  required for buildings containing predominantly Class A laboratories that
 need office support.  (Ordinarily, Class A laboratories should be in separate
 buildings  with minimal office space; in these areas, ventilation Zone III is
 optional for  the uncontrolled areas.)
     B.  Air  Flow Patterns,  Air should flow from the ceiling to the floor of
 a laboratory  and should not flow directly across bench tops.   In general,
 laboratories  should be designed to provide draft-free conditions to keep the
 movement of particulate matter by air currents as low as possible.
     The air  flow for the whole building and for individual laboratories should
 be  from areas of low (or no) radioactivity to areas of progressively "higher
 activity.  This direction of flow ensures that material that may become air-
 borne will not contaminate other areas in excess of their permitted limits.
     C.  Pressure Differentials.  Pressure differentials should be  used to
 maintain the  desired air flow characteristics.  The exhaust system  should be
 used to keep  areas with relatively high activity levels at a negative pressure
 relative to the rest of the building.  The building itself should have a
 negative pressure relative to the outside.   In order  to maintain the proper
 pressure differentials and keep the  air flowing in  the desired  direction, the
 supply  fan delivering air to laboratories should be controlled  by interlocks
 that automatically shut off the air  supply so that  it  is  impossible to deliver
 air to  the laboratories when the exhaust  system is  shut down for any reason.
     D.  Duct Routing.  Exhaust ducts in  multistory buildings  should be  routed
 to  common  ducts, or plenums, that are easily accessible.   In addition, ducts
 should  be  labeled as to their point  of origin.  For single-story buildings,
 hoods should  be verted to the roof using  zhe least  possible  amount  of  ducting
 inside  the building.
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                                                               3ARCO.M-P  385-1
     E.  Filtration
     (1)  Type and Location.  Filters or traps for exhaust  air are  required to
ensure that release levels are kept ALARA.   Filter systems  should be  designed
for easy access, removal, contamination control,  and  in-place  testing.   In
general, exhaust filters should be placed close to the  hoods,  glove boxes, and
hot cells in order to avoid contaminating ventilation duct  systems.
     (2)  Backflow Prevention.  If the air flow through a filter were reversed,
radioactive particulates could be pulled into a laboratory,  with serious con-
sequences.   For this reason, filters that routinely become  burdened with radio-
active particulates should be protected by dampers that restrict the reverse
flow of air.  Inverse-flow dampers can be simple, weighted,  shutter-like
dampers that open passively with positive air flow.  In dampers with more
complex designs, electrical mechanisms keep the dampers open,  and springs or
pressurized air ensures their closure if the electrical supply is disrupted.
     (3)  Testing.  Filters should be designed so that  they can be tested in
place.   A DOP mist is used to test HEPA filters.   Charcoal  filters can be
tested using a gaseous halogenated-hydrocarbon refrigerant,  in accordance with
Section 12 of the American National Standards Institute's  (ANSI) Standard
N510-1975,  to ensure that bypass leakage through the absorber section is less
than 0.05S.
     (4)  Maintenance Accessibility.  Ventilation filters  and blowers require
periodic removal and replacement.  Filter systems are often contaminated at the
time of their replacement, and maintenance personnel  must  be protected against
possible inhalation of radioactive dusts, mists,  and fumes  during filter
replacement.  External exposure is also a potential problem if the filters are
loaded with radionuclides that emit gamma rays or high-energy beta particles.
The filter units should be placed so that individual  filters can be removed
easily without the need for scaffolding.  If scaffolding is required, however,
enough  free floor space should be available for the installation of the
scaffolding.
                                     8.31

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 A.RCO.M-P 3S5-1
.8.6.3  Sampling and Monitoring Equipment
     An important aspect of facility design is to provide for sampling and
measurement of the concentration of airborne radioactive materials and for
monitoring of the radiation levels in the workplace.   As discussed in
Chapter 5, sampling is the collection of air that is  then analyzed for
activity levels at a  Icter time and in a different place; monitoring, on the
other hand, 13 the continuous reading of the radiation level  in a facility by
a radiation detection instrument.  Types of sampling  and monitoring eouipment
                  I
are discussed in Chapter 2.
     All sampling and monitoring instruments should have lights that indicate
whether the instrument is turned on, in standby mode, or not operating.  These
lights, or status indicators, should be readily visible from any work area.
All monitors should be provided with both visual and. audible alarms.  The
instruments should be designed so that, if an alarm has been tripped, the
instrument must be rest manually; automatic cessation of the alarm function is
not acceptable.
     A.  Air Samplers and Monitors.  All Class A and B laboratories  should be
equipped with fixed systems for sampling and monitoring the air.  The sampling
heads should be placed where releases could occur, as well as  in  front of each
room's air exhaust.
     Areas occupied by personnel where concentrations of airborne radionuclides
may exceed the concentrations given in 10 CFR 20, Appendix B,  Table  I, should
contain a continuous-monitoring device that activates an alarm when  the air-
borne concentration exceeds 25% of the values given  in the table.
     B.  Radiation Area Monitors.  Continuously operating area monitors should
be provided to measure the ambient dose-equivalent rate wherever  that  rate may
exceed 50 mrem/hr.  An alarm on each radiation monitor should  notify workers  if
the device is not operating.  Each radiation monitor should actuate  audible and
visual alarms whenever a preset radiation limit has  been exceeded.   The instru-
ments should be capable of measuring dose-equivalent rates in  the range of
10,000 mrem/hr.  Finally, the detector portions of the monitors  should be
easily replaced and should be located where they can be  calibrated  in  place.
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                                                              DARCOM-P 3E5-I
                         Section 8.7  FACILITY  SUPPORT

     Other considerations in facility design  are  the  provision of  change  rooms,
decontamination facilities,  and separate supply and sewer  systems  for  sanitary
and process water.
8.7.1  Change Room Facilities
     Rooms in which workers  can change clothing should  be  available  and should
be designed to prevent cross-contamination.   Each worker should  have two
lockers, one for clean clothing and another  for potentially  contaminated  cloth-
ing.   Snowers should also be provided in the  change rooms.   Change rooms  may be
separate from or part of personnel  decontamination facilities.
8.7.2  Personnel and Property Decontamination Facilities
     Facilities for the decontamination of personnel  and property  should  be
available.  Decontamination  facilities for personnel  should  have showers.  The
shower drains should be separate from the sanitary sewer system  and  should
empty into a holding tank if contamination levels are expected  to  be high.
     Facilities for the decontamination of property  should be large  enough to
accommodate the largest piece of equipment.   Each facility should  include a
hood  and should have drains  that are directed to  holding tanks.
8.7.3  Water Supply and Sanitary Sewers
     Sanitary water provided in radiation areas shall be used for safety
showers and fire protection  sprinklers only.   Drinking fountains should not be
located in radiation areas.   Process water supplied  to radiation areas shall be
isolated from sanitary water systems by the  use of either  separate systems or
back-flow preventers.
     Sinks in radiation areas should not be  equipped  with  drains connected to a
sanitary sewer.   If sinks and drain lines are connected to a sanitary sewer,
they  shall be so labeled, and the discharge  of radioactive wastes to any
sanitary sewer shall be prohibited.
                                     8.33

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- .-uu.-;-.1  38 „<

                                  REFERENCES


American National Standards Institute (ANSI).   1975.   Testinc_of Nuclear Air-
  Cleaning Systems.   ANSI N510, American Society of Mechanical  Engineers,
  New York.

Fitzgerald, J. J.  1969.  Applied Radiation Protection and Control, Volume 1.
  Gordon and Breach, New York.

Industrial Ventilation - A Manual of Recommended Practice.  1980.  American
  Conference of Governmental Industrial Hygienists.Available from the
  Committee on Industrial Ventilation, P.O. Box 16153, Lansing, Michigan 48901.

Kathren, R. L., and J. M. Selby.  1980.  A Guide to Reducing Radiation Exposure
  to As Low As Reasonably Achievable (ALARAT   DOE/.EV/1E30-T5, Nat-.onal Tech-
  nical Information Service, Springfield, Virginia.

National Council on Radiation Protection and Measurements (NCRP).  1978.
  Operational Radiation Safety Program.  NCRP 59, Washington, D.C.

U.S. Code of Federal Regulations.  1982. Title 10, Part 20, "Standards for
  Protection Against Radiation."  U.S. Government  Printing Office, Washington,
  D.C.

U.S. Department of the Army, Headquarters.  Safety -  Ionizing Radiation  Protec-
  tion (Licensing, Control, Transportation, Disposal, and Radiation Safety).
  AR 385-11, Washington, D.C.

U.S. Department of the Army, Headquarters, Army Materiel  Command.  Safety -
  Radiation Protection.  DARCOM-R 385-25, Washington, D.C.
                                     8.34

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                                                                     DARCDM-P 385-i

                CHAPTER 9.  TRANSPORTATION OF RADIOACTIVE MATERIALS

Transportation of radioactive materials is governed in part by Title 49,  Code of
Federal Regulations; Title 10, Code of Federal Regulations; Army Regulations
AR 700-64, AR 385-11, and AR 55-355.
                                        9-1

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                                                              DARCOM-? 385-1

            CHAPTER 10.   MANAGEMENT OF LOW-LEVEL  RADIOACTIVE WASTE
10.1  MINIMIZING THE GENERATION OF  WASTE	10.3

10.2  COLLECTION OF WASTE	10.4

      10.2.1   Segregation of Radioactive  Waste     	     10.4

              A.  Half-Life	10.5

              B.  Biological Waste  	     10.5

              C.  Nonbiological Waste   .......     10.6

              D.  Scintillation Vials   	     10.7

      10.2.2   Containers for Collection and Temporary  Storage
              of Waste	10.8

              A.  Containers for Biological Waste 	     10.8

              B.  Containers for Nonbiological  Solid Waste  .     .     .     10.8

              C.  Containers for Nonbiological  Liquid  Waste .     .     .     10.8

10.3  FACILITIES FOR THE STORAGE OF WASTE	10.8

      10.3.1   Site Selection	10.9

      10.3.2   Control  Procedures   	     10.9

10.4  REDUCTION OF WASTE VOLUMES	1C.9

      10.4.1   Solidification  	     10.10

      10.4.2   Compaction 	     10.11

      10.4.3   Incineration    	     10.11

10.5  AUTHORIZED WASTE DISPOSAL PROCEDURES   	     10.12

      10.5.1   Requests for Disposal Instructions  .....     10.12

      10.5.2   Shipping Instructions     .......     10.13

      10.5.3   Onsite Assistance    	     10.14

REFERENCES	10.14
                                     10.1

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                                                              DARCO'-P  385-1
            CHAPTER 10.  MANAGEMENT OF LOW-LEVEL  RADIOACTIVE  WASTE
     Low-level radioactive waste is waste that contains  1)  low enough levels
of beta-gamma activity so that no speciel provisions  must  be  made for heat
removal, and 2) low enough levels of penetrating  radiation  so that minimal or
no biological shielding or remote handling is necessary  for personnel protec-
tion.  Low-level waste is generally considered to contain  less than 100 nCi of
transuranic alpha emitters (uranium, thorium, etc.)  per  gram of waste.   The
handling, storage, and disposal of low-level  radioactive waste must conform to
strict requirements imposed by the Department of  the  Army  (DA), the Nuclear
Regulatory Commission (NRC), the Department of Transportation (DOT), and the
operators of waste burial grounds.
     This chapter provides guidance for persons who  generate low-level  radio-
active waste and for those responsible for its handling, storage, and disposal.
Topics covered include the generation and collection  of  waste, facilities for
storing it, and procedures for reducing waste volumes and  obtaining DA assis-
tance in waste disposal.   Further questions about the management of low-level
waste should be directed  to HQ, ARRCOM, ATTN:  DRSAR-SF, Health Physicist,
Rock Island, Illinois 61299.  Telephone calls can also be  placed to
(309) 794-3383; FTS 367-3483; or AUTOVON 793-4942.
               Section 10.1  MINIMIZING THE GENERATION OF WASTE

     The use of radioactive material should be planned so that a minimum amount
of radioactive waste is generated.  For example, when a procedure requires the
use of radioactive material, a dry run using nonradioactive material can elimi-
nate errors that might cause contamination and create considerable waste.  The
smallest quantity of radioactive material  needed to effectively perform a task
should always be used.
     The volume of radioactive waste can be reduced •..  nonradicective and
radioactive wastes are separated and not discarded together.  Solid dry wastes
                                     10.3

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JARCOM-P 385-.
that may be contan.;rated should be surveyed  and  the  nonradioactive  portions
discarded by conventional methods.  When  a device  is  being  discarded,  any
radioactive components should be removed  if  the  device  can  be  disassembled
safely and the disassembly is authorized  by  the  NRC  license or DA permit.
Every effort should be made to decontaminate contaminated property  before  it
is disposed of.  However, the volume of waste that would be generated  by the
decontamination procedure should be considered before low-cost items  are
decontaminated (see Chapter 7).
                       Section 10.2  COLLECTION OF WASTE

     The total quantity of radioactive material disposed of into sanitary
sewage systems, the air, or nearby streams as a result of a 11  activities at an
installation must not exceed the quantities for a single licensee given in
10 CFR 20, or the quantity limitations established by applicable regulatory
agencies.  Individual users of radioactive materiel must not dispose of waste
directly by these methods unless specifically authorized by the Radiation
Protection Officer (RPO).  Instead, each user should collect any low-level
wastes according to the guidelines in this section.
     When wastes are being collected at a facility, the radioactive waste
should be separated from the nonradioactive waste.  Wastes that are taken from
a radiation area should be presumed to be radioactive unless shown to be
otherwise.  This is particularly true in hot laboratories, where paper tissue
and even writing paper may become significantly contaminated.   Radioactive
wastes should be segregated into classes of material so that all constituents
of any one batch can be dealt with in the same way.  They should be collected
in suitable containers for processing and disposal by the RPO or a designated
representative.
10.2.1  Segregation of Radioactive Waste
     Characterization of low-level radioactive waste is important for proper
waste handling and processing for final disposal.  Characterization includes
identification of the physical form of the waste,  the type and half-life  of
                                     10.4

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                                                              DARCOM-P 3E5-1
radionuclities present, the total- activity  and/or  specific activity,  and other
properties of the waste such as its  volatility, explosiveness, and  toxicity.
     Once characterized, wastes should  be  collected according  to  type  by  each
user of radioactive materials.   Procedures  for segregating  and collecting
wastes should be developed by the RPC  and  provided to  all individuals  who may
generate radioactive waste as a result  of  their work.  The  procedures  should
cover the segregation of wastes by half-life  and  by the  characteristics
described below.  The waste collected  under each  category can  be  further
separated by whether it is combustible  or  compactible.
     A.  Half-Life.  Waste containing  short-lived radionuclides  (those with a
half-life Ui/o) shorter than 30 days)  should be  collected  separately  from
waste containing long-lived radionuclides  (those  with  a  half-life longer  than
30 days).  Short-lived materiel can  usually be stored  away  from  work areas for
10 half-lives of the longest-lived radionuclide  in the material  and then  dis-
carded as nonradioactive material.  It  must be surveyed  before disposal  by
conventional methods.  Long-lived material  should be  processed for disposal as
radioactive waste.
     B.  Biological Waste.  Biological  waste, which originates primarily from
medical and research facilities, normally  undergoes decomposition by micro-
organisms, producing foul-smelling matter.  Such  material  requires freezer
storage.
     (1)  Sol id.  Solid biological waste includes radioactively contaminated
animal carcasses, fecal matter, soiled  animal bedding, and  plant by-products.
Personnel working with animals  should  be aware of radiation levels and of the
excretion routes for various radiochemicals and  drugs (National  Council on
Radiation Protection and Measurements  (NCRP)  Report  No.  48, 1976).  Animals
that are used in studies o-~ radioactive materials should not be petted or
groomed, and their carcasses should not be hand-carried if a radiation over-
exposure to the hands or body of the person carrying  them may result.   Remote
handling and storage is advised (TM 3-261).
     (2)  Liquid.  Liquid biological waste includes  radi&?ctive~.y contaminated
blood, urine, and culture media.  Because  biological  waste should be  stored
frozen, containers  should be capable of withstanding  temperature extremes
                                     10.5

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DARCOM-P 385-1
without breaking and should be filled about three-quarters  full  to  allow  for
expansion of the contents.  Polyethylene containers  are  preferred.   Personnel
dealing with liouid biological wastes should consider  not only  the  radiological
hazards and the need to provide radiation protection,  but also  the  potential
chemical and biological hazards that may be associated with the wastes.
     C.  Monbiological Waste.   Nonbiological waste is  any radioactively con-
taminated waste that, under ordinary circumstances,  does not undergo decomposi-
                  i
tion by microorganisms.
     (1)  Sol id.  Solid nonbiological waste makes up the major  portion of low-
level radioactive waste.  It includes radioactively  contaminated glassware,
protective clothing, gloves, paper, metal scraps, syringes, filters, sealed
sources, and equipment or equipment components (compasses,  meters,  electron
tubes, etc.).  Depleted uranium, either as an ore or in  metal  form, also  falls
into this category.  If a device with a solid source is  not internally or
externally contaminated, it should be handled in a manner  that  prevents  its
contamination.  For example, it should not be placed in  the same collection
container as a pair of contaminated gloves.
     (2)  Liquid.  Not all liquids are disposed of in the  same  way; therefore,
liquid nonbiological waste should be segregated into aqueous and nonaqueous
waste.  Aqueous waste—any waste in which water is the primary  solvent—
includes water used to decontaminate material or personnel, and solutions of
radioactive material used in a laboratory.  Nonequeous waste is any liquid in
which water is not the primary solvent.
     Any chemically reactive liquids should be further segregated and identi-
fied.  Organic liquids (those containing carbon compounds)  should be segre-
gated from aqueous solutions to prevent the possibility of violent reactions.
Nitric acid and alcohol, for example, if disposed of in the same vessel,  could
react together and cause an extensive spread of contamination.   Unless special
arrangements are made with the RPO, individuals who generate strongly acidic
or basic waste solutions should neutralize or dilute them enough so that they
will not cause violent chemical reactions or release strong fumes and vapors.
In the case of organic solvents, especially those that are highly volatile,
appropriate precautions should be noted on the waste container.
                                     10.6

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                                                              DARGON-P  385-1
     When the methods used to dispose of liquid wastes  include absorption  of
the liquid or ion exchange processes, the potential  for chemical  interactions
that could affect the process should be evaluated.   Types  of  liquid  wastes
that could cause adverse effects on the processing  of the  waste  include  acidic
or besic solutions; liquids containing complexing or wetting  agents; and
liquids containing certain detergents.  Precautions  must be  taken to prevent
the accidental processing of incompatible liquid wastes.
     D.  Scintillation Vials.  Small glass or plastic vials  containing  scintil-
lation fluids and low levels of redioactively labeled compounds  may  be  handled
as an entity; the contents rf the vials need not be  transferred  to a waste
container.  The vials should be packaged (preferably in their original  car-
tons) to avoid breakage, and the box should be properly labeled.
10.2.2  Containers for Collection and Temporary Storage of Waste
     Containers used for the collection and temporary storage of radioactive
waste should be made of materials that will not rust or corrode  from contact
with the wastes stored in them.  The lids of the containers  must be easy to
open so that the containers do not tip over when the lids are being  removed.
     Each container of radioactive waste should be  painted bright yellow and
marked "Caution - Radioactive Material."  It should be labeled with enough
information to permit accurate identification of the waste it contains.   This
information, which should be noted on the label at  the time the  waste is
placed in the container, should include:
 1.  the name of the waste generator
 2.  the date
 4.  the pH of a waste solution
 3.  the chemical name of the waste material
 5.  the isotope(s) contained in the waste
 6.  the activity level
 7.  any information on the" biological or chemical  hazards associated with  the
     wa s te.
     Waste containers should be checked periodically to ensure that radiation
levels ere not excessive, that outside surfaces are free of contamination,  and
                                     10.7

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that corrosion or rust is not weakening the  container.   If  e waste  container
thet is being used for the collection of wester  develops  a  high  external-
rad'.etion exposure level  or becomes externally  contaminated, it  should  not  U-
kept even temporarily at  thf- user's location,  but  should  be moved  ircmedu le :y
to the storage site for radioactive wastes.  A  container  that  is corroding  anri
losing its integrity should be placed inside a  second  container  before  being
moved.
     Individuals who generate waste should notify  the  RPO whenever  a  container
is filled and ready for removal.   The RPO should remove  the waste  and place it
in a centralized area for temporary storage  and  consolidation.   Containers
should not be moveo unless they are labeled  and  the  waste is contained  in
accordance with installation requirements.
     A.  Containers for Biological  Waste.  Solid biological waste  must  be
seeled in plastic bags and frozen.   Liquid biological  waste should  be stored
in plastic containers that can be frozen without breaking.  Glass  containers
are not acceptable (TM3-261).  Biological wastes are packed in lime for
shipping.
     B.  Containers for Nonbiological Solid  Waste.  Solid waste must be sealed
in plastic bags.  It can  be stored in a metal  waste  can  with a plastic  liner
and a lid that operates by a step-pedal.  When  the waste is to be  moved, it
must be packaged so thet  pipettes,  hypodermic  needles, and  other sharp  objects
cannot penetrate through  the plastic bag.
     C.  Containers for Nonbiological Liquid Waste.   Glass  containers should
not be used to store liquid waste.   Aoueous  waste may be kept  in polypropylene
carboys or jugs.  Nonaqueous waste (organic  solvents, acids,  and bases) may be
kept in metal solvent cans or in plastic containers  if the  liquid  will  not
dissolve the plastic.
               Section 10.3  FACILITIES FOR THE STORAGE OF WASTE

     A facility should be designated for the centralized storage of radioac-
tive wastes until  they are shipped for processing or burial.
                                     10.8

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10.3.1  Site Selection
     When a storage facility i:  being selected, wnetner  i.n  existing  structure
or a new structure, the RPO or the individual  responsible should  ensure  that
the following guidelines, in addition to those  in  Charter 8,  are  met:
 ].  The facility is close to the point  of  origin  of  t'-e waste  but away  from
     main areas of personnel tra-fic or  areas  where  routine access is
     requi red.
 ?.  The facility is weatherproof and hes adequate ventilation.
 3.  Enough storage space is provided to allow for variations in  shipping
     schedules  and, if possible, to store short-"ived materials  (those with  a
     half-life  shorter than 30 days) while  they decay.
 4.  Separate storage compartments are provided for  combustible  liquids  (for
     fire prevention).
 5.  Means of handling wastes efficiently are  provided,  to  minimize  personnel
     exposures.
 6.  The radiation dose limits for the unrestricted  area around the  facility
     will not be exceeded.
10.3.2  Control Procedures
     To keep personnel exposures to a minimum  and to protect the general public,
only individuals responsible for storing or shipping waste  should have access
to the waste storage facility.  The wastes  should be kept segregated by  type,
with higher-level waste placed far from the facility entrance to reduce  the
exposure to personnel who enter the area.  As  waste is brought into  or taken
out of storage, the amount and type of the  waste  moved,  the date, and tne name
of the user or  shipper should be entered in a  log book.   Personnel  monitoring
should be provided to ensure contamination  control.
                        Section 10.4  VOLUME REDUCTION
     Reducing the volume of low-level waste has the following benefits:
                                     10.9

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 1.  It increases the stability of the waste  form
 2.  It minimizes the possibility that radionutlides  will  be  released  to  the
     environment during interim storage,  transportation,  and  burial.
 3.  It leads to savings in transportation  and  burial  costs,  which  are
     dependent on waste volume.
 4.  It reduces the exposure of personnel  handling  the waste.
As discussed earlier, volume reduction should be accomplished primarily by
each person minimizing the amount of waste  generated.   Wastes that  have been
generated can be reduced in volume by solidification,  compaction,  and
incineration.
     Volume reduction processes can be carried  out  most economically at central
waste-consolidation facilities to which many  installations or sites ship their
radioactive wastes for treatment before final disposal.  The  use of volume
reduction equipment at an Army installation requires  an NRC license and a DA
authorization or permit.
10.4.1  Solidification
     Many burial sites require that the wastes  they handle meet certain
physical forms.  Low-level liquid wastes must be  converted to a solid that
will not leach.  Loose, dry residues from incinerators or dryers must be bound
together into a solid waste form.
     A variety of methods are used to solidify  wastes and reduce their volume.
Aqueous solutions are treated by crystallization  and dehydration.   Crystal-
lization is the removal of water, usually by  evaporation, which results in a
slurry of precipitated solids mixed with a  saturated solution.  The slurry is
then mixed with a setting agent such as cement.  Dehydration is the removal of
all the water from liquid wastes, leaving a residue of solids.  Aqueous liquids
and dry residues from incinerators and dryers can  be mixed with a binding
agent to form a solid waste.  Conventional  setting and binding agents are
cement, bitumen, glass, and urea-formaldehyde;  experimental materials include
vinyl esters, polyethylene, epoxy resins,  and an  inorganic binder.
                                     10.10

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                                                              DARCOM-P 385-1
10.4.2  Compaction
     Compaction (the removal  of excess air)  is  the  most  widely used  method  of
volume reduction for dry, nonbiological  wastes  that are  not  combustible.  Com-
paction includes compressing  the waste into  a  final  disposal  container (such
as a 208-liter drum) and baling the compressed  waste with  bends  before packag-
ing it.  Items that are currently compacted  in  the  commercial fuel  cycle
include high-efficiency particulate air (HEPA)  filters and old contaminated
drums.
     Before compaction, the waste should go  through some pretreatment.  Com-
pactible and noncompactible items should be  separated.   Hazardous  materials
(such as explosives) and materials containing  free  liquid  should be  removed.
Items that would otherwise be too large for  the compactor  can be shredded using
knife cutters or hammernrills.  For example,  equipment and  metal  can be pack-
aged as is or shredded in a hammer-mill and compacted.
     A typical compactor for  low-level waste consists of a hydraulic system
with a vertical ram, a contoured support plate, a  frame, a safety enclosure,
and automatic controls.  These drum compactors  should be located in protective
enclosures, which prevent the escape of airborne particulate matter.  A hood
or shroud around the drum opening, with a HEPA filter and an exhaust blower,
serves to control particulates.  Some drum compactors incorporate a metal
inner sleeve to protect the drum walls from the pressure of  the  ram and from
rigid metal objects.
10.4.3  Incineration
     Incineration is the removal of combustible material in  radioactive waste.
Water and air are removed at  the same time.   The types of incinerators avail-
able for radioactive-waste processing include  controlled-air incinerators,
Tluidized-bed incinerators, and rotary kilns.   Ihese systems differ in operat-
ing temperatures, waste residence times, chamber turbulence, and amount of
oxygen used.  Each incineration system requires specific methods of waste
pretreatment, feeding, ash removal, and off-gas treatment.
     All  incinerators for radioactive waste must have an off-gas systeu to
keep particulete and gaseous  effluents within  NRC,  Environmental Protection
                                     10.11

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DARCOM-P 385--
Agency (EPA), a'nd state limits.   These o^f-gas  systems  result  in  an  additional
radioactive waste stream that must be considered.
     The main advantage of incineration as a volume  reduction  process  is  the
uniform end product, ash, which  is easy to solidify  and thus minimizes  the
problems associated wizh the disposal of a wide range of materials.  The  main
disadvantage is the high initial  cost; because  a relatively  large volume  of
waste material must be generated to make the procedure  cost-effective,  incine-
ration is not economical for most sites.
                 Section 10.5  AUTHORIZED DISPOSAL  PROCEDURES

     The DA program for the disposal of low-level  radioactive waste is managed
by HQ, ARRCOM, Rock Island, Illinois.  The authority for world-wide management
of the program is assigned in AR 385-11.
     Low-level waste that cannot be disposed of locally because of local
restrictions is disposed of by land burial in Barnwell, South Carolina, or
Richland, Washington, by commercial radioactive-waste-disposal firms under
contract with HQ, ARRCOM.  Under certain conditions, waste shipments are sent
to a collecting point operated by a waste disposal  broker or the Army.  At the
collecting point, they are consolidated and ultimately disposed of.
10.5.1  Requests for Disposal Instructions
     The RPO is responsible for requesting disposal instructions from the
Commander, US Army Armament Materiel Readiness Command, ATTN:  DRSAR-DSM-D,
Rock Island, Illinois 61299.  The request can be made by letter or message.
     Requests for disposal instructions must contain the following informa-
tion:
 1.  nomenclature, national stock number, and serial numbers
 2.  physical descriptions of the items to be disposed of, including:
     a.   whether solid, liquid, or gas
     b.   the quantity per stock number and, if gas, the volume under  standard
          pressure and temperature
                                     10.12

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                                                                   M-;-  365-1
     c.   the shipping weight (pounds) and volume (measured to the nearest
          cubic foot)
     d.   '.he number of shipping containers
     e,   the shipping permit or waiver number
     f.   the transport group
     g.   the package specification
     h.   the labels used
 3.  chemical and radioisotope description, including:
     e.   the hazardous chemicals present
     b.   for liquids, the solvent present
     c.   the radioisotopes present
 4.  radioactivity and radiation measurement, including:
     a.   the millicuries of activity of each radioisotope; for special
          nuclear material, give the number of grams; for source materiel,
          list the quantity in pounds
     b.   maximum radiation dose rates (mrem/hr) at the surface and 1 meter
          from the surface of the package
     c.   the classification, basis for classification, and procedures for
          declessification
     d.   special instructions or requests for unique service, such as return
          of the containers
     e.   the name and telephone number of the responsible person to contact
          for additional  information
     f.   remarks, i' appropriate.
Requests for technical information or assistance should be submitted to the
Commander, ARRCOM, ATIN:   DRSAR-SF, Rock Island, Illinois 61299.  Telephone
requests can be made by celling (309) 794-3383/4728; FTS 367-3383/4728; or
AUTOVON 793-3383/4728.
10.5.2  Shipping Instructions
     Shipping instructions will be furnished by HQ, ARRCOM, in reply to requests
tor dispose! instructions.  Each request will be handled as a separate action,
and the instructions will include the followino:
                                     10.13

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    K  r  385-.
 1.  an ARRCOM-assioned control  number,  which  will  serve as  tne  identification
     for each request
 2.  the address of the shipping destination as  determined by  HQ,  ARRCOM;  the
     destination may be a land burial  site or  a  collection/consolidation
     point
 3.  specific marking, packaging, and  transportation  instructions.
Because safety concerns and burial  criteria change  periodically,  special
instructions will  also bt furnished.
10.5.3  Unsite Assistance
     Radioactive-waste shipments may be  audited  by  HQ,  ARRCOM, at the
shipper's installation prior to shipment.   Some  audits  require that an ARRCOM
audit team be onsite to supervise the  packaging  and' loading  of the radioactive
material.  Requests for onsite assistance  should be addressed  to Commander,  US
Army Armament Materiel Readiness Command,  ATTN:   DRSAR-SF,  Rock Island,
Illinois 61299.
                                  REFERENCES

National Council on Radiation Protection and Measurements (NCRP).  1976.
  Radiation Protection for Medical  and Allied Health Personnel.   NCRP 48,
  Washington, D.C.
U.S. Code of Federal  Regulations.  1982.  Title 10, Part 20, "Standards for
  Protection Against Radiation."  U.S. Government Printing Office,
  Washington, D.C.
U.S. Department of the Army, Headquarters.  Handling and Disposal of Unwanted
  Radioactive Materiel.  TM 3-261,  Washington, D.C.
U.S. Department of the Army, Headquarters.  Safety - Ionizing Radiation
  Protection (Licensing, Control, Transportation, Disposal. and Radiation
  Safety).AR 385-11,"Washington,  D.C.
                                     10.14

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                                                              DARCOM-P  365-1
          CHAPTER 11.   RADIATION ACCIDENTS  AND  EMERGENCY  PREPAREDNESS
11.1   THE EMERGENCY PLAN	11.5
      11.1.1  Responsibility for Emergency Planning      ....   11.6
      11.1.2  System for Classifying Emergencies    	   11.6
      11.1.3  Emergency Response Organization  ......   11.9
      11.1.4  Characterization of Installation  and  Facilities      .     .   11.13
      11.1.5  DA-Authorized and NRC-Licensed Activities  ....   11.13
      11.1.6  Emergency Plan Implementation   	   11.13
              A.   Procedures	11.13
              B.   Notification	11.14
      11.1.7  Response Actions .    .     	   11.14
              A.   Assessment Actions     	   11.15
              B.   Corrective Actions     	   11.16
              C.   Protective Actions     .    .     .     .     .     .    .11.16
      11.1.8  Facilities and Equipment   	   11.17
              A.   Emergency Control  Centers   	   11.18
              B.   Medical  Treatment  Facility   	   11.19
              C.   Assembly Areas	11.20
              D.   Communications Equipment    ......   11.20
              E.   Monitoring Equipment   .......  11.21
              F.   Aerial Monitoring  	   11.22
              G.   Dosimeters	11.22
              H.   Transportation Modes   	  11.22
      11.1.9  Offsite Agreements and Support   	  11.23
                                     11.1

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DARCDM-*5 3£b-l



      21.1.10 Re-Entry and Recovery 	  11.23



11.2  MAINTAINING A STATE OF EMERGENCY PREPAREDNESS     ....  11.24



      il.2.1  Training Staff and Emergency Response Personnel      .     .  11.24



      11.2.2  Training Members of the News Media   .....  11.26



      11.2.3  Conducting Emergency Drills     	  11.26



      11.2.4  Maintaining and  Inventorying Emergency Equipment     .     .  11.26



      11.2.5  Reviewing and Updating Plans and Procedures    .     .     .  11.27



11.3  NOTIFICATION AND REPORTING REQUIREMENTS 	  11.27



      11.3.1  Notification and Reporting Requipments:  Army  .     .     .  11.29



              A.  Notification	11.30



              B.  Reports and  Investigations  ......  11.31



      11.3.2  Notification and Reporting Requirements:  NRC  .     .     .  11.32



              A.  Notification .    .    .    .    .    .    .     .     .11.32



              B.  Reports ..........  11.32



      11.3.3  Notification and Reporting Requirements:  DOT   .     .     .  11.33



REFERENCES	11.34



APPENDIX A  - EXAMPLES OF CHECKLISTS	11.37



APPENDIX B  - RESPONSE ACTIONS  	  11.43



APPENDIX C  - EXAMPLE LISTING OF EMERGENCY  KIT EQUIPMENT .     .     .     .11.51



APPENDIX D  - EXAMPLES OF EMERGENCY ACTIONS    	  11.57
                                      11.2

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                                                               DAKCO.S-P 3EI-1
                                    TABLES

11.1  Significant Quantities of Byproduct Materic's      ....  11.7
11.2  Emergency Condition Classificetion Scheme     .....  11.10
11.3  Minimum Organizational Support for Emergency Preparedness   .     .  11.12
11.4  Instruments for Emergency Radiological  Measurements    .     .     .11.16
11.5  DA Criteria for Defining Radiation Accidents 	  11.28
11.6  NRC and DA Notification Requirements for Accidents
      Involving Licensed Materials  ........  11.28
11.7  DOT and DA Notification Requirements for Accidents Involving
      Army Motor Vehicles Carrying Licensed Materials   ....  11.29
11.8  DA Criteria for Individual  Radiation Overexposures     .     .     .  11.30
11.9  Summary of Response Actions for Individual  External Exposure     .  11.46
11.10 Actions to be Taken Within  Six Hours Following a Whole-Body
      Exposure	11.47
11.11 Actions to be Taken Following Suspected Internal
      Contamination  	  11.48
                                     11.3

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                                                               DARCOX-P 365-1
           CHAPTER  11.  RADIATION ACCIDENTS "KD EMERGENCY PREPAREDNESS
     A  radiological emergency is any sudden or unforeseen situation in ^,'hich
 damage  to  persons or property, or interruption in operations, has occurr-.-d or
 is  imminent  unless corrective actions are taken.   The severity of both an
 accident and  its effects can be decreased if procedures are followed, engi-
 neered  controls are used, and corrective end protective actions are taken.
     Planning for radiological emergencies can uncover problems that, if
 corrected, will decrease the likelihood of an accident.  Therefore, a plan for
 responding to abnormal occurrences should be developed and maintained for each
 individual operation involving radioactive materials.  Each plan will vary
 from others  accordings to the specifics of the operation.  The magnitude of
 the emergency planning needed at an installation and the notification,
 reporting, and investigative procedures required in the event of an accident
 depend  on  the potential hazards at each facility and the types of accidents
 that may occur.
     In this chapter, radiological accidents are identified and classified,
 guidance is provided on how to prepare for potential accidents by developing
 an emergency preparedness plan and how to maintain a state of emergency pre-
 paredness, and accident reporting and investigative procedures are reviewed.
 Emergency preparedness is a full-time specialty of health physics that
 requires training and experience.  This chapter is intended to introduce the
 Radiation Protection Officer (RPO) to emergency preparedness.  The assistance
 of a trained specialist should be sought for developig extensive plans and
 emergency responses.
                       Section 11.1  THE EMERGENCY PLAN

     An emergency plan is a document that details the best response to an
emergency situation, with primary cncern for protecting the health and safety
of Army and civilian personnel and the general  public.  A comprehensive plan
should contain the following key elements:
                                     11.5

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     *-P 365-1
 ;.   designation of responsibility  for  emergency  planning
 2.   assessment of potential  accidents
 3.   system for classifying emergencies
 4.   description of the  emergency  response  organization
 5.   characterization of the installation and  its facilities
 6.   description of activities  authorized by the  Department of  the  Army  (DA)
     and the Nuclear Regulatory Commission  (1,'RC)
 7.   procedures for implementing the emergency plan
 8.   response actions
 9.   description of facilities  and  equipment
10.   description of offsite agreements  and  support capabilities
11.   re-entry and recovery conditions.
The  use of a checklist such as  that presented in  Appendix  A can help ensure
that all aspects of an emergency plan have  been considered.
11.1.1  Responsibility for Emergency Planning
     The commander of each installation is  responsible for planning for and
providing training for credible emergencies (AR 385-11).   This  duty may be
delegated to an organization within the command that has  the  operational
experience and technical abilities  necessary to direct planning efforts.
Personnel involved in emergency planning must have the authority to gather
site-specific information, write procedures, and enter into  discussions with
offsite agencies.  In many cases,  the RPO and the Ionizing Radiation Control
Committee (IRCC) are the logical choices for this duty.   If  the duty is
delegated elsewhere, the RPO and the IRCC should be  involved  in at least the
radiological assessment, control,  and recovery aspects of emergency planning.

11.1.2  System for Classifying Emergencies
     Emergency plans and procedures should be developed for all facilities
where radioactive materials are handled, used, stored, or transported,
regardless of quantity.   However,  formal documented emergency plans must be

                                     11.6

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                                                              DARCO.M-P  385-1
prepared (DARCOK Disaster Control  Plans  (DCP),  Annex  E)  if  the quantities  of
radioactive materials exceed:
 1.  1 uCi of radium
 2.  the quantities listed in  Schedule B of 10  CFR 30.71
 3.  6.8 kg of source material
 A,  5 pg of special nuclear material.
Schedule B of 10 CFR 30.71 sets limits for byproduct  materials.   A portion of
Schedule B (the more common byproduct materials) has  been reproduced in
Table 11.1.

         TABLE 11.1.  Significant  Quantities of Byproduct Materials'6'
 Byproduct Materiel       Microcuries    Byproduct Material      Micrpcuries
     3H (tritium)           1,000               115Cd                 100
    14C                       100              115mCd                  10
    IP
    1CF                     1,000                 Sb                 10
    32p                        1Q              125,                   l
    35S                       100              131I                   1
    36C1                       10              133Ba                 10
    42 K                        10              133Xe                100
    54Mn                       10              135Xe                100
    59Fe                       10              137Cs                 10
    60Co                        1              144Ce                  1
    65Zn                       10              147Pm                 10
    85K                       100              148Pm                 10
    9°Sr                        0.1            197Hg                100
    9°Y                        10             197mHc                100
    99Mo                      100              198Au                100
    99Tc                       10              204T1                 10
   99mTc                      100              210Bi                  1
   109Cd                       10              210Po                  0.1
   115ln                       10
(a)  Excerpted from 10 CFR 30.71,  Schedule B.
                                     11.7

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   COM-P 385-1
     The scope of the emergency plan depends  on  the  potential  hazards  of  the
maximum credible accident and other postulated accidents.   The maximum cred-
•;ble accident is the accident that would cause the highest  radiation exposures
to onsite personnel and/or the public.   Although the maximum credible  accident
poses the greatest threat, all potential accidents should  be considered in  the
development of emergency plans.  Trie presence of small  quantities  of  radio-
active materials may require only a few procedures and  telephone  numbers, with
minimal supplies and equipment (e.g., ropes,  signs,  and survey meters).   The
presence of large quantities may require an extensive plan, many  procedures,
and facilities and equipment dedicated to an  emergency  response.
     Assistance should be obtained for emergency planning,  particularly if  the
installation does not have the resources to handle the  identified  credible
accidents.  Assistance may be available from Army emergency response  teams  or
health physics specialists in emergency preparedness.  If  local personnel
cannot identify such assistance, contact DARCOM  or the  office of  the  Surgeon
General of the United States.
     Four classes of emergency conditions that are frequently used in  the
nuclear industry to classify potential hazards — unusual event, alert,  site
emergency, and general emergency — are described  in Table 11.2 (pages  11.10-
11.11), based on NRC's NUREG 0654 (1980).  The classes  are defined in  terms of
onsite and off site consequences and projected dose commitments and exposure
rates at the boundary of the event site, which may be the  door of a laboratory
or a building, or a restricted-access gate on base.   Army  operations  would
typically encompass only the first two emergency classes:   unusual event and
alert.  If a site emergency or general emergency that might cause the release
of radioactive materials to offsite locations could  occur  at an installation,
assistance should be sought in designing and developing emergency plans and
procedures.
     The following topics should be considered in the  development of emer-
gency plans:
 1.  the kinds of radioactive materiel s potentially   released  (so the*,  respon-
     sive monitoring instrumentation can be identified)
                                     11.8

-------
                                                               DARCOX-P  385-1
 2.  the most important exposure pathways  for  these  types  of materials  (so
     thtt trie effect on the local  population can  be  determined)
 3.  a definition of the area for which  planning  should  be carried out  (celled
     the emergency planning zone (EPZ))
 4.  the potential duration of a release and the  time  available before
     exposures offsite are significant (so that  protective actions can  be
     decided upon).
     Specific conditions, both actual  and  imminent,  that require an emergency
response are called emergency action levels (EALs)  and are trie basis for
declaring an unusual event, an alert,  or a higher classification of accident.
When the EALs have been identified and documented,  the procedures, facilities,
and equipment required for a response  can  also be identified.  Thus, the EALs
can provide a framework for developing emergency procedures.
     Another useful  classification system  (Brodsky 1980) groups commonly used
radionuclides into eight groups based  on the  relative  magnitudes of their
maximum radiotoxicities.  This system  was  presented in Chapter 1 of this man-
ual.  It can be useful in determining  EALs and specifying subsequent actions.
11.1.3  Emergency Response Organization
     The coordinated efforts of several  organizations  may be required to produce
an adequate emergency response.  In the  emergency plan, one individual  must be
designated as having overall responsibility and authority for  implementing and
directing emergency procedures.  Each  support  organization and its responsibil-
ities must be identified, and persons  responsible for each group must be  iden-
tified by title or position, along with  any alternates, to assure e 24-hr/dsy
response.  All individuals assigned responsibilities must have knowledge  of
and experience in radiological emergency preparedness.
     Table 11.3 is a listing of the organizational support personnel who  must
be available at any installation and included  in any plan, with brief example
descriptions of their responsibilities.   Site  requirements may call for  £ more
complex list or may allow two or more  organizatio1 :"i functions tc be consoli-
dated.  Several duties of key response personnel cannot be delegated.   For
instance, the emergency director cannot assign subordinates  the responsibility
                                     11.9

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DARCOK-P  385-1

















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                                                              DARCOM-P 3E5-1
for declaring emergency classifications  or  recommending protective  actions.
Responsibilities that cannot be delegated must  be  identified  in  the  plan.
11.1.4  Characterization of Installation and  Facilities
     The emergency plan should include a description of  .ne principal  char-
acteristics of the installation.   Approximate populations  of  onsite  and
offsite structures should be identified. Aerial photographs  or  site maps
should be used to identify the location  of  facilities  or areas relevant  to
emergency planning.   These could include:   1) the  location of population
centers (office buildings, schools, barracks, stadiums,  personnel  housing);
2) the location of facilities that could present potential evacuation  problems
(hospitals, schools); 3) identification  of  primary routes  for bringing in
emergency equipment or for evacuating personnel or the public; 4)  location  of
emergency support facilities (fire stations,  hospitals with capability for
handling patients with radioactive contamination); and 5)  other  sites  of
potential emergency significance (hazardous chemicals, gas lines).
     Facilities in which radiological activities are  conducted should  be
concisely described.   The description should  include  confinement structures for
handling and storing  radioactive and other  hazardous  materials;  auxiliary
systems such as ventilation; radioactive waste  management; and detection and
alarm systems.
11.1.5  DA-Authorized and NRC-Licensed Activities
     Work that involves radioactive materials and  that is  authorized by the
Army and licensed by  NRC should be described  in the emergency plan.  Included
should be the location of the work; the type, form, and quantity of the
radioactive materials used; the type of waste produced; and the individuals
responsible for the activities.
11.1.6  Emergency Plan Implementation
     The emergency plan should include detailed instructions  for carrying out
emergency response actions and information  on required notifications.
     A.  Procedures.   The detailed response procedures should include the
following:
                                     11.13

-------
DARCOM-P 385-1
 1.  specific EALs and the actions planned in response  to  them
 2.  a statement of the responsibilities assigned to each  individual,  and
     which responsibilities may not be delegated
 3.  references to support documents and procedures that supplement the
     emergency plan.
The procedures should be developed to ensure that all  positions  will  be manned
and all appropriate emergency organizations will be operational  in the event
of an emergency.  '
     B.  Notification.  When an emergency class is declared,  prompt notifica-
tion of personnel is vital to response.  Methods and procedures  for 24-hr/day
notification of each organization that has an emergency response assignment
are necessary.  A site-wide notification system (i.e.,  public address or
pageboy system) is useful in alerting site personnel; however,  someone must
confirm that response groups have been notified.  A call list of key emergency
response personnel and their alternates, and of DA and NRC contacts, should be
part of the emergency plan, and one person or group should be designated to
contact them at the direction of the emergency director.  Contacts should be
completed within 15 minutes of the declaration of an emergency class.
     The methods of communication that will be used to notify onsite and off-
site personnel must also be specified in the emergency plan,  including a
description of all primary and back-up notification equipment.   Messages and
announcements that are planned and written out in advance are useful and
should be incorporated into the procedures to avoid delays and
misunderstandings.
11.1.7  Response Actions
     Emergency response actions fall into three general categories:  assess-
ment actions, corrective actions, and protective actions.  Individuals who
have emergency response assignments should be experienced in their assigned
responsibilities and should have access to procedures that stipulate what
actions should be taken.  Procedures should be well written, easy to under-
stand, and presented in a "cookbook" format, with space allotted for notes.
Appendix A contains a sample checklist of procedures to be followed  in the
                                     11.14

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                                                              DARCOX-P  385-1
event of a minor spill.  Appendix B provides  specific  response actions  end
consiaeretions for accidents involving exposure  to  'ndividuals and  for  trans-
portation accidents.
     The following sections provide a synopsis  of how  the three  categories  of
response actions should be treated in the emergency plan.
     A.  Assessment Actions.  Responding to accident situations  requires
knowing both present and impending radiological  conditions,  which can be
calculated using available information and supplemented with data obtained
from radiological surveys.  If insufficient information is available for
making calculations, survey data alone may be used  to  determine  emergency
response actions.
     For radiological surveys, instruments and  equipment capable of measuring
all anticipated conditions must be available and operational.  The type and
number of instrums'.is needed depend on how extensive the onsite  and offsite
measurements will be.  A program may be greatly simplified if only onsite
response is required.  An offsite capability requires  thorough planning over a
large area, special radiological equipment, and vehicles for transporting
personnel and equipment.
     Instruments must be capable of measuring the full range of anticipated
radiation intensities and types.  The specifications provided by vendors should
be tested, as should each instrument's response to the 50-year environmental
extremes recorded in each location.
     Op.site parameters that must be measured are dose rate, contamination count
rate, and the concentrations of radionuclides in air and effluents.  Offsite
parameters are the same except that meteorological  data  are also needed.
Examples of instrument types appropriate for making these measurements are
found in Table 11.4.  (See also Chapter 2, "Radiation Instrumentation.")
     For offsite dose assessment, simple equations must  be developed that
allow accurate calculation of integrated dose within 15  minutes of when data
are received.   A computer or desk-top calculator can be  programmed with com-
plex equations so that the insertion of required parameters  is all that is
                                     11.15

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DARCOM-P 385-1
       TABLE  11.4.  Instruments -for Emergency Radiological Measurements
	Parameter	Instrument Types	
Dose  rate                          Medium- to high-range ionization chamber
Surface count rate                 Geiger-Mueller detectors
                                   Scintillators
Concentrations of radionuclides    Air-sampling device (air pump, vacuum pump)
in air                             Analyzer:  g-as proportional counter or
                                   scintillation counter
Concentrations of radionuclides    Sampling devices {air, water, soil)
in effluents                       Analyzer:  gas proportional counter or
                                   scintillation counter
Meteorological conditions          Devices to determine wind speed and
                                   direction, temperature, and stability
                                   class

needed to run the program.  Loss of electrical power must not affect the abil-
ity to make this calculation.  The person responsible for assessment should be
guided by the emergency plan on how to apply the assessment data to obtain
projected doses.
     B.  Corrective Actions.  Efforts must be made to reduce the likelihood
that an accident will  recur.  In general, a thorough investigation is needed to
identify areas that are weak and need strengthening.  The results of the
investigation should lead to appropriate corrective actions.  If several
alternative actions are possible, the action taken should be the one that
incorporates, to the greatesT extent possible, engineered safeguards rather
than administrative guidelines.
     C.  Protective Actions.  In an accident, all radiation doses should be
kept as low as is reasonably achievable (ALARA) while the situation is brought
under control.  Limiting doses is best accomplished by limiting the release of
materials through either engineered controls or manual actions.  Because this
is not always possible, protective actions should be developed to control the
exposure of personnel  er.J the public.
                                     11.16

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                                                              DARCOM-P 385-1
     Examples of onsite actions  that should  be  considered are:
 1.  providing protective clothing and respirators  for  use by emergency
     workers
 2.  sealing windows and doors and shutting  off ventilation  systems  until
     conditions improve
 3.  removing personnel who are  not contributing to the emergency  response.
     If personnel  may need to be evacuated to offsite areas,  routes  end
methods of evacuation should be  planned and  a destination upwind from any
release should be identified.  Provision must exist for transport  vehicles,
radiological surveys of personnel  and vehicles, and offsite  decontamination.
     Emergency plans must also include ways  of accounting for onsite
personnel.  Procedures should specify
 1.  personnel assembly points
 2.  the individual(s) responsible for accountability at each point
 3.  the individual  to whom accountability status is reported
 4.  the individual  responsible  for notifying search-and-rescue teams.
As a general rule, the names of  ell missing  persons should  be determined
within 30 minutes of the declaration of an emergency.   All  personnel remaining
onsite should be continuously accounted for.
li.1.8  Facilities and Equipment
     An emergency plan and the response based on it can be  effective only if
adequate facilities  and equipment are available.  For example,  an  offsite
monitoring team would be useless if it did not have monitoring  instruments
that could measure in the range  of emergency conditions or  if it did not have
communications equipment to report back the information gathered.   The design
of facilities and the types of equipment required for effective response
depend largely upon  the maximum credible accident and  other postulated
accidents.  A variety of considerations in the design  and selection of
facilities and equipment for handling both small- and   u--ge-sc&ie accidents is
presented below.  Judgment should dictate which considerations  are appropriate
for a given installation.
                                     11.17

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:>ARCDM-P 385-1
     A.  Emergency Control Centers.   To facilitate -the  coordinetion,  direc-
tion, and evaluation of the emergency response for  site and  general  emergencies,
one facility should be designated as the emergency  control center (ECC).
Because this area would be the hub of activity in an emergency,  its  location
and design should be considered carefully.   The ECC should have  a low proba-
bility of being affected by an accident.  If a postulated accident would
result in high radiation levels in the ECC,  its location should  be changed.
     Space is a primary requirement of the  ECC. Adequate space  must be allotted
fcr each activity or group involved in the  emergency response.   Consideration
must be given not only to the number of persons involved, but also to the
space needed for chairs, tables, and monitoring and communications equipment.
The -assignment of space to groups is also important; groups  that work together
should not be on opposite sides of the room or across the hall  from each
other.
     The onsite and offsite communications  system  in the ECC is  another
primary consideration.  The system should be operational within  15 minutes
after the activation of the ECC.  The director of  each emergency response
organization must have at least one dedicated communications link between the
organization and the ECC.  The emergency director  should have several open
lines available for use.
     The facility that is set aside as the  ECC should be reserved for
emergency use only.  The emergency supplies kept there should be periodically
inventoried and replenished as needed.  Items that should be available in the
ECC (depending on the scope of the postulated accidents) include:
 1.  the documented emergency plans, procedures, and checklists for the  site
 2.  state and local emergency plans and procedures
 3.  emergency power
 4.  survey meters
 5.  air samplers
 6.  sample-counting equipment (unless adequate provisions  are  available for
     counting samples offsite)
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                                                              DARCOH-P  385-1
 7.  personnel dosimeters for e'tl  the occupants
 8.  eelibration sources
 9.  site and erea maps marked with preselected  monitoring  points,  locations
     of tnermoluminescence dosimeters (TLDs),  and  environmental  air sampling
     stations (useful maps are the U.S.  Geological  Survey 7-1/2-minute maps,
     which cover the p'ume exposure EPZ  and are  marked with cardinal  polar
     coordinates and 22-1/2-degree sectors, with the first  section  splitting
     true north)'
10.  a board for posting emergency assignments and team designations
11.  a board for posting up-to-date meteorological  conditions and estimated
     doses at given distances from the release
12.  as-built facility and building layouts
13.  first aid kit and decontamination supplies
14.  clock
15.  writing materials and note pads
16,  protective clothing
17.  dose assessment equipment such as calculators
18.  basic reference material
19.  communications equipment (telephone, radio, etc.).
     B.  Medical Treatment Facility.  Provisions must be made for either the
installation's health personnel or e local hospital to care for contaminated
individuals who are injured in an emergency.  Information may be found in
AR 40-13.   Briefly, the following needs  should be considered when a center for
handling contaminated patients is being  designed and equipped:
 1.  easy and immediate access
 2.  stretchers
 3.  first aid equipment and supplies
 4.  communication link
 5.  medical  personnel  trained in the handling of contaminated patients
                                     11.19

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DARCOM-P 385-1
 6.  operable, calibrated instruments for surveying  contaminated  patients
 7.  documented procedures for decontaminating patients
 8.  source of water and suitable decontaminants
 9.  provisions for the collection and disposal  of solid  and  liquid waste.
     C.  Assembly Areas.  Assembly areas where personnel  gather when an alert
is sounded should be able to accommodate the assigned number  of persons.
Consideration should be given to the adequcy of shielding,  ventilation, rest
rooms, communications equipment, and portable lighting for  these areas.
     D.  Communications Equipment.  Many types of communications equipment
can be used during an emergency, including alarms, pageboy  call systems,
walkie-talkies, telephones, and two-way radios.   The operation of each piece of
equipment should be checked regularly and personnel  should  be trained to use
the equipment.
     Each communications link should have a back-up and an  alternate power
source.  In addition, at least one communications system should provide uninter-
rupted service during a power failure.
     Areas or groups that should be equipped with a communications system
include:
 1.  the ECC, the emergency director, and directors of emergency response
     organizations
 2.  assembly areas and medical facilities
 3.  onsite monitoring teams
 4.  offsite monitoring teams
 5.  security personnel
 6.  the public (if applicable to postulated accidents).
     The range of communications equipment used by monitoring  teams and secu-
rity personnel must be known.  If the offsite monitoring team  uses a two-way
radio to comrrjnicate with the ECC, the radio must be able to transmit  over  the
required distance.
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                                                              DAKCO.M-P 3E5-:.
     In a general  emergency,  a  communications  system must be available  to warn
the affected public.   Sirens  mounted  on  telephone  poles, or the  local fire  or
police station, can serve this  purpose.   The public must know what  action to
take when alarms sound.
     E.  Monitoring Equipment.   Onsite and  offsite radiation-monitoring equip-
ment must be capable  of  measuring  the types and  levels of radiation expected
during a postulated accident  and must be calibrated in the postulated accident
range, using a source traceable to the National  Bureau of Standards (see
Chanter 2).  For this reason, it is suggested  that a number of  portable instru-
ments be dedicated to emergency response situations.  These instruments should
be checked routinely  for operability  and should  be calibrated annually.
     Many factors  affect the  choice of fixed and portable instruments for
emergency response.  The instruments  must be capable of  responding  in extreme
environmental conditions, such  as  high or low  temperatures or humidity.
Because many instruments do not operate  in  temperatures  below -10°F, the
manufacturer's performance specifications (which indicate the range of
operability of en  instrument) should  be  checked, and the instruments should be
tested in the field during extreme weather  conditions.
     The accessibility of fixed instrumentation  during  postulated accidents
should be assessed.  If  valuable data would be lost due  to  inaccessibility,
remote readouts should be considered. A power failure  may  also render  an
instrument or its  data inaccessible.   If a  particular  instrument's  data is
necessary for accurately assessing the  impact  of an accident,  provision should
be made so thet it will  continue to function during a  power failure.
     Fixed air monitors  can warn of airborne radiological  hazards if they  are
designed to trip an alarm that  will be heard or  seen by site  personnel.  There-
tore, these alarms should be  placed et manned  locations.
     Records should be kept for each  instrument  that: will  be  used in an emer-
gency, documenting the type of  radiation the  instrument is  designed to measure
and the maximum and minimum radiation levels  it  can detect.   The dates  of end
data from operational checks  and calibrations  should  also be  documented.  A
label indicating the  date of  the latest  operational check and calibration and
any conversion factors to be  used  in  data interpretation should be  placed on
                                     11.21

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DARCOM-P .'-85•
each "instrument    ''ne storage location of all  portable instruments and
supplies for emergency response should be documented in the emergency
procedures.
     Kits used for onsite measurement and monitoring of radiation should be
easily accessible for determining the initial  accident conditions.  The kits
should contain a high-range dose rate meter, a contamination monitor, portable
sampling devices, and light protective clothing.   Electronic equipment should
be tested periodically for operability, and the contents of the kit should be
inventoried routinely.  A breakable seal  should be placed on each kit immedi-
ately after inventory so that any intrusion into the kit can be detected.  An
inventory should be taken promptly upon the discovery of a broken seal.  A
sample listing of emergency kit equipment is provided in Appendix C.
     F.  Aerial Monitoring.  When an effluent release (the plume pathway) is
being tracked, unfavorable meteorological conditions or the passage of the
plume over inaccessible areas may hinder an accurate determination of the
plume's location.  In such cases, aerial  surveillance using helicopters or
fixed light-wing aircraft can contribute valuable information by providing
survey data over a large area.  Helicopters are best suited for emergency
radiation surveys because of their maneuverability and slower flying speeds.
A two-man crew (the pilot and someone to operate the radiation detection
equipment) would generally be needed for such aircraft.
     G.  Dosimeters.  Dosimeters that are designated for use only in emerg-
encies should be available for each member of the emergency response team.
ihese dosimeters must be capable of responding to the types and levels of  radia-
tion that would be present during postulated accidents.  Pocket ionization
chambers should be worn and checked frequently, especially by onsite and off-
site monitoring teams.  Each member of the emergency response organization
should be assigned a film badge or TLD or both to record the dose received
during the emergency.
     H.  Transportation Modes.  Vehicles must be available to transport  injured
persons to either an onsite or an offsite medical facility.  If an  ambulance
from a nearby hospital will be used, prior arrangements must be made for immedi-
ate service.  Monitoring teams need vehicles for their exclusive  use that  can
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                                                              DARCOM-P  385-1
ce-~y monitoring equipment and emergency kits  and  handle  any  environmental  or
road conditions that may be encountered.
11.1.9  Offsite Agreements and Support
     Offsite support can be invaluable in accident situations.   Personnel  at
an installation cannot always perform all the  tasks needed  to respond to an
emergency.  Areas in which offsite support may be  needed  are  fire fighting,
health physics, security, and medical aid.
     Advance agreements should be made with support organizations for their
assistance.  The agreements should specify the support  to be  provided and the
conditions under which that support will be used.
11.1.10  Re-Entry and Recovery
     During the period between the end of an emergency  and  restart of opera-
tions affected by the accident, imminent danger is not  expected  but the
potential for higher-than-normal  exposures may exist.   The  emergency plan
should provide guidance on keeping these exposures to a minimum  and ensuring
that no recovery actions would place the installation back  in an emergency
situation.
     Evacuated buildings must be  re-entered with caution  and  only after a
complete hazard assessment has been made and the emergency  director has
authorized re-entry.   The only exceptions to these conditions may be for
firefighting and search-and-rescue teams, whose activities  must  be supervised
by the health physics staff.
     The following topics relating to re-entry and recovery should be
addressed in the emergency plan:
 1.   the conditions  (e.g., exposure rates, radionuclide concentrations) under
     which rooms or  buildings may be re-entered prior to  their return to
     normal  operation
 2.   the identification of personnel to direct re-entry and' recovery
 2.   the assurance of proper  communications to keep site  personnel, response
     organizations,  and DA and NRC personnel informed of  progress in re-entry
     and recoverv.
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DARCOM-P 385-1
          Section 11.2  MAINTAINING A STATE OF EMERGENCY PREPAREDNESS

     Maintaining a state of emergency preparedness require;  the effort of
every individual within the installation.  Each person needs to understand his
or her  responsibility and how it helps ensure the safety of  the installation
and its occupants.  Emergency telephone numbers should be posted next to tele-
phones, and diagrams of evacuation routes and lists of emergency signals with
their meanings  should be posted on bulletin boards or in hallways.
Maintaining emergency preparedness includes:
  1.  training and retraining staff and emergency response personnel
  2.  conducting emergency drills
  3.  maintaining and inventorying emergency equipment, instruments, and
     supplies
  4.  reviewing  and updating plans and procedures.
11.2.1  Training Staff and Emergency Response Personnel
     All staff  members and emergency response personnel must be familiar with
the radiological emergency plan if it is to be effective.  They should  receive
training in:
  1.  safety and accident control features specific to the facility to which
     they are assigned
  2.  the emergency signals (sirens, alarms),  their meaning, and the  expected
     response
  3.  the location of emergency assembly  areas
  4.  the building layout, including emergency exits  and  evacuation routes
  5.  notification procedures and immediate  actions if they  discover  or  are
     involved in a radiation accident.
Personnel assigned emergency response duties  require additional training in
the proper execution of their duties.  A representative list  of persons or
groups  requiring this specialized training  includes:
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                                                               DARCOH-P  385-1
 1.  directors and coordinators of the plant emergency organization (see
     Table 11.3)
 2.  personnel responsible for radiological  assessment
 3.  radiation-monitoring end survey teams
 4.  radiation protection personnel
 5.  maintenance teams
 6.  security personnel
 7.  search-and-rescue teams
 8.  firefighting squads
 9.  medical personnel
10,  communications personnel
11.  staff of state and local agencies and offsite support teams (if
     appliceble).
     Formal lesson plans should be drawn up for each training session, and the
training program for each group should be documented.  Each training program
should give personnel an understanding of the emergency response plan and the
role that each group plays in its implementation.  The specific duties of each
group and how these duties are to be performed (e.g., how to use equipment,
whom to notify when, end how to treat a contaminated wound) should be
included.  Special  precautions to observe, in the performance of radiological
emergency duties should also be included in the training program (see
Appendix D).  Whenever possible, practical hands-on operation of equipment and
facilities should  be included in the training program.
     The quality of training depends to a large extent upon the quality of the
instructors.   A good instructor is professionally competent and has good com-
munication skills.   The instructor must also be thoroughly familiar with the
emergency plan and  each person's role in it.  An effective training program may
require the combined efforts of several individuals or organizations.
     Retraining is  important in maintcining a state of emergency preparedness.
Because emergency  duties are Seldom performed, they are easy to forget.  Formal
training sessions  should be held at least once a year.

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DARCOM-P 385-1
     Provision must be made for evaluating the ability of individuals  to per-
form their emergency duties.  The conditions, tasks,  and standards of  perform-
ance that form the basis for this evaluation should be documented.  Attendance
records and test scores from training sessions should also be documented.
11.2.2  Training Members of the News Media
     When emergency planning includes offsite locations, training should be
offered to individuals from the local news media.  Newspersons should  be
                 i
trained in basic radiation protection practices and associated terminology.
During an accident, one location should be designated as the media center, and
all newspersons should be directed to that area upon arrival at the installa-
tion.  The public relations spokesperson from the installation should  be
responsible for providing the media with up-to-date information, to help avoid
conflicting stories and general confusion among the reporters and to help
maintain credibility with the public.
11.2.3  Conducting Emergency Drills
     Emergency plans should be tested annually through the use of emergency
drills (AR 385-11).  Drills jog memories, lead to the application of skills
learned in training sessions, and keep interest in emergency response duties
high.  Drills also allow problem areas to be identified and corrected under
controlled rather than accident conditions.  In a full-scale dri-11, all onsite
and offsite participants respond to a simulated severe accident.  Smaller-
scale drills involving specific response organizations should be  held every
6  months.
11.2.4  Maintaining and Inventorying Emergency Equipment
     lo maintain a state of emergency preparedness, a schedule  for maintaining
equipment and supplies should be developed and followed.  The inventory of kits
and supplies should be checked periodically  for completeness.   This check  should
include operating and calibrating all instruments.  The maintenance procedures
should specify the corrective actions to be  taken  promptly  when deficiencies
are found during these checks.
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                                                              DARCOM-P  3t5-i
11.2.5  Reviewing and Updating Plans and Procedures
     When conditions chenge within an installation,  emergency  plans and  pro-
cedures may need to be changed to meet the new conditions.   The  extent of the
updating needed may range from changing a name on  a  call  list  to reassessing
potential accidents if a new radiological function is  defined.   To ensure the
adequacy and effectiveness of emergency preparedness,  provisions should  be
made for a periodic review and update of the radiological  emergency plan.  A
full-scale review should be conducted annually by  a  committee  designated for
this purpose in the emergency plan.  This committee  would ensure that:
 1.  the emergency plan and procedures are current
 2.  training sessions and drills have been conducted  on  schedule, test scores
     and drill  deficiencies have been documented,  and  corrective actions have
     been taken
 3.  the emergency plan addresses the postulated accidents.
An individual or a committee should also be assigned to make necessary changes
in call lists or equipment inventories as they occur.   The name of the person
or persons responsible for such changes should be  documented in the emergency
plan.
             Section 11.3  NOTIFICATION AND REPORTING REQUIREMENTS

     The DA criteria for defining radiation accidents are based on individual
exposures, effluent releases, damage to property, and loss or theft of radio-
acti.'° material and are given in Table 11.5.  Both Army personnel and civilian
licensing agencies must be notified when accidents that meet these criteria
occur.  Tables 11.6 and 11.7 list how soon notification is required for
different accident levels, as set forth by DA (AR 385-40), NRC (10 CFR 20),
and the Department of Transportation (DOT) (49 CFR 171).  Other requirements
for notification and for investigations and reports are given below for the
three groups.
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DARCOM-P 385-1

     TABLE  11.5.  DA Criteria for Defining Radiation Accidents (AR 385-40)

     Tvoe of Accident                           Criteria
 Individual Exposure
Effluent Releases

Damage to Property

Loss or Theft of Radio-
active Material
              1.  External  exposure:
                    Exposure  greater  than  limits  in
                    10 CFR 20
              2.  Internal  exposure:
                    Airborne  concentrations  in  a  restricted
                    area,  or  234y,  235u,  238y  concentra-
                    tions  greater  than  limits  in  10  CFR  20,
                    Appendix  E,  Table 1,  Column 1
              3.  Fatality, lost-time  injury,  restricted-duty
                 work

              Greater than 500 times  the  limits in 10 CFR 20,
              Appendix B,  Table  II  (averaged  over 24 hours)

              1.  Cost is  S300.00 or more
              2.  Loss of  facility  operation  for 1 day or more

              Quantity that may  result  in substantial hazard
              to  personnel in unrestricted areas
   TABLE 11.6.
Notification

Immediate
NRC AND DA Notification Requirements  for Accidents Involving
Licensed Materials^5)
Within
24 hours
    Individual  Exposure

  Whole body (head,  trunk,
  active blood-forming
  organs, lens  of eye,
  gonads) _>25 rem

  Skin j>150 rem

  Extremities _^375 rem

  Whole body (head,  trunk,
  active blood-forming
  organs, lens  of eye,
  gonads) j^5 rem

  Skin ^30 rem

  Extremities >75 rem
  Release
>5000 x amount
listed in
10 CFR 20,
Appendix B,
Table II,
averaged over
24 hours
>500 x amount
listed in
10 CFR 20,
Appendix B,
Table II,
averaged over
24 hours
  Damage
to Property

>S200,000

Loss of
_>1 week
of facility
operation
>S2,000

Loss of
±1 day
of facility
operatton
(a) Exce-pted from 10 CFR 20 and AR 385-40.
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                                                              DARCOM-I  385-1
   TABLE 11.7.   DOT and DA Notification Requirements  for Accidents  Involving
                Army Motor Vehicles  Carrying  Licensed Materials'2)
   Notification         Individual  Exposure           Damage  to  Property
   As soon es       Any event that  presents a       >S50,000
   practicable       hazard to personnel at  the
                    site                           Fire, breakage,  slippage,
                                                   or suspected  radioactive
                    Fatality or lost-time           contamination
                    injury
   (a) Excerpted from 49 CFR 171 and AR 385-40.

11.3.1  Notification and Reporting Requirements:   Army
     The criteria indicating what constitutes  radiation  accidents  are further
subdivided into four DA classifications based  on  the  degree  of  damage caused
by the accident.  These four general classifications  are used for  all Army
accidents except aircraft mishaps.
 1.  Class A accident
     a.    property damage, injury, or occupational  illness  costing $200,000 or
          more
     b.    fatality as result of Army operations
     c.    fatal  injury of off-duty Army military  personnel.
 2.  Class B accident
     a.    property damage, injury, or occupational  illness  costing between
          $50,000 and $200,000.
 3.  Class C accident
     e.    property damage costing between $300 and $50,000
     b.    loss of one or more workdays due to  injury  or  occupational illness.
 4.  Class D accident
     a.    property damage less than $300
     b.    one or more days of restricted v/ork  activity due  to  injury or
          occupational  illness
     c.    nonfatal  case without loss of workdays.
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DARCOM-P 385-1

Although immediate emergency actions and notification  do  not  depend  on  these

classifications, recording, reporting, and Investigation  requirements do.

     In addition to the accident criteria for individual  exposures  specified

in Table 11.5 and those described by Classes A,  B,  C,  and D above,  AR 40-14

defines three types of radiation overexposures to individuals.   These classes

are summarized in Table 11.8, and the reporting  requirements  are specified in

Section B below.

     A.  Notification.  The following Army personnel  must be  notified by tele-

phone or electrical means immediately or within  24  hours  of an  accident (see

Table 11.6) (this notification applies to Type III  individual  overexposures  in

Table 11.8):

 1.  the affected major Army commander or his representative

 2.  the licensee
        TABLE 11.8.  DA Criteria for Individual Radiation Overexposures
                     (AR 40-14)
       Body Part
Whole body, head
  and trunk, active
  blood-forming organs,
  gonads, lens of eye

Skin of whole body,
  forearms, cornea
  of eye

Hands and wrists,
  feet and ankles

Other organs (bone,
  thyroid, tissue,
  organ systems)
      Type I
   Overexposure

>400 mrem/nxr '
<1.25 mrem/qtr
>3 rem/mo but
<7.5 rem/qtr
>6 rem/mo but
<18.75 rem/qtr

>1 rem/mo but
<5 rem/qtr
        Type II
	    Overexposure

but         (b)
            (b)



            (b)


            (b)
   Type III
 Overexposure

>5 rem/yr or
>1.25 rem/qtr
>30 rem/yr or
>7.5 rem/qtr
>75 rem/yr or
>18.75 rem/qtr

>15 rem/yr or
>5 rem/qtr
(a) mo = calendar monzh; qtr = calendar quarter; yr = calendar year.
(b) Dose rate exceeds the quarterly rate for a Type I Overexposure but is less
    than the annual rate for a Type III Overexposure.
                                     11.30

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                                                              DAKCOM-P  365-1
 3. "HDQA (DAPE-HRS), AUTOVON 225-7291;  DASG-PSP,  AUTOVON  227-2795
 4.  HQ L-'ARCOM (DRCSF-P), AUTOVON 284-9340
 5.  the Chief of Engineers (DAEN-rEZ-N),  AUTOVON1  354-5501,  if the accident
     occurs et a reactor facility.
     B.  Reports and Investigations.   The  initial  report must  contain tne
following information:
"This is a Radiological Accident Report, RCS:DD-SD(AR)1168."
 1.  the date of the event
 2.  the radiation-producing device or source involved,  including its national
     stock number, nomenclature, and  radiation characteristics and parameters
 3.  a description of the event, including the cause,  the  name and social
     security number of each person exposed,  estimated exposures and dose
     rates, contamination levels, facilities  affected, and actions taken
 4.  any action taken to prevent a  recurrence
 5.  recommendations on how to avoid  similar  accidents at  other installations
     possessing similar material
 6.  a specific contact (name, address,  telephone  number)
 7.  a statement of when appropriate  DA, NRC, and  DOT offices were notified.
     Class A, B, and C accidents must be documented and a  report  (DA Form 285)
must be submitted to the U.S. Army Safety Center in Fort Rucker, Alabama,
within 30 days of the accident.  All  Class A  accidents require a formal  board
of investigation.  This board is appointed by the  commander to whom the radio-
active materials license has been issued.   Class B and C accidents are investi-
gated by the local commander.  Reports of these investigations should be
forwarded through channels to HQDA (DAPE-HRS, DASG-PSP), Washington, DC 20310
and to Commander, DARCOM (DRCSF-P), 5001 Eisenhower Avenue, Alexandria, VA
22333, within 90 days of the accident.  The requirements of the Privacy Act of
1974 must be taken into account whenever an individual is  identified.
     An informal investigation of Type I individual overexposures  (see
Table 11.8) is conducted by the immediate commander.  The commander must
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DARCOM-P 385-1
conduct a formal investigation of Type II and Type III  overexposures  and  for-
ward a report of tie investigation and of the corrective actions  through
command channels to HQDA (DASG-PSP), Washington,  DC 20310.
     The investigation of a radiation accident can establish its  cause  and
identify corrective and protective actions that will  prevent the  recurrence of
the accident.  The investigating individual  or group should:
 1.  collect and preserve evidence
 2.  interview witnesses
 3.  prepare diagrams of the accident scene
 4.  re-enact the accident if appropriate.
     When trying to establish the cause of an accident, the investigator(s)
should consider possible defects in a component's basic design or construc-
tion.  If a component is faulty, it should be identified in the investigative
report by name, model number, manufacturer,  and name-plate data.   Other possible
causes of an accident that should be considered are human error or misjudgment,
incomplete or incorrect procedures, or the absence of procedures.
11.3.2  Notification and Reporting Requirements:   NRC
     Either NRC or an agreement state^3' licenses Army installations  to use
radioactive materials.
     A.  Notification.  If the license is from NRC, the director of the
appropriate NRC Inspection and Enforcement Regional Office  (see 10 CFR 20,
Appendix D) must be notified of an accident.  If the license is from an
agreement state, the director of the branch of state government issuing the
license must be notified.  Notification time shall be as described in
Table 11.6.
     B.  Reports.  A formal written report must be sent within 30 days of any
accident to the appropriate NRC regional office listed in 10 CFR 20,  Appen-
dix B.  A copy of this report should be submitted to the Director of Inspec-
tion and Enforcement, USNRC, Washington, DC 10555.
(a) An agreement state is any state with which NRC has entered into an effec-
    tive agreement under Section 274 b. of the Atomic Energy Act of 1954, as
    amended (73 Stat. 689).
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                                                              DARCOM-P  385-1
     Reports of tneft on loss of licensed materiel  or  of  individual  overexpo-
sures (Type III) should include the following:
 1.  6 description of the licensed material  involved,  including  the  kind,  quan-
     tity, and chemical and physical  form
 2.  a description of the circumstances under which the  loss,  theft, or  over-
     exposure occurred
 3.  a statement of the disposition or probable disposition  of the licensed
     materiel involved
 4.  quantitative radiation exposures to individuals and the extent of possible
     hazard to persons in unrestricted areas
 5.  actions that have been or will be taken to recover  lost or stolen
     material
 6.  procedures or measures that have been or will  be adopted to prevent a
     recurrence of the loss, theft, or overexposure.
     After filing the written report, the licensee  shall also report any
substantive additional information on the accident  within 30 days after the
licensee learns of such information.
     In reports filed with NRC, the names of individuals who may have been
exposed to radiation shall be stated in a separate  part of the report,
including for each individual exposed the person's  name, social security
•number, and date of birth, and an estimate of the individual's exposure.  The
requirements of the Privacy Act must be taken into  account whenever an
individual is identified.
11.3.3  Notification and Reporting Requirements:  DOT
     Each carrier must notify DOT at the earliest practicable moment after c
transportation accident specified in Table 11.7.  Notification should be given
by telephone ((800)442-8802) and should include the following information:
 1.  the name and phone number of the individual reporting the accident
 2.  the nans snd address of the carrier represented by the individual
     reporting the accident
                                     11.33

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     N' P Jfi5-l

 2.   *he date, time,  location,  and nature  of  the accident

 4   the classification, name,  and quantity of  radioactive materials  involved

 5.   the extent of injuries, if any,  and whether a  continuing  danger  to  life

     exists at the accident scene.

A written report must be submitted to DOT  in  duplicate within  15  days of the

accident.
                                  REFERENCES


Brodsky, A.   1980.   "Determining Industrial Hygiene  Requirements  for Installa-
  tions Using Radioactive Materials."  Health  Phys.  38:1155-1171.

International Commission on Radiological  Protection  (ICRP).   1978.   The
  Principles and General Procedures for Handling  Emergency  and Accidental
  Exposures  of Workers.  ICRP 28, Pergamon Press,  Oxford.

Privacy Act, 1974.   U.S. Code. Title 5, Section  552a.

U.S. Code of Federal  Regulations.  1982.   Title  10,  Part  20,  "Standards for
  Protection Against Radiation."  U.S.  Government Printing  Office,
  Washington, 5.C.

U.S. Code of Federal  Regulations.  1982.   Title  10,  Part  30,  "Rules of General
  Applicability to  Domestic Licensing of Byproduct Material."  U.S. Government
  Printing Office,  Washington, D.C.

U.S. Code of Federal  Regulations.  1980.   Title  49,  Part  171, "Hazardous
  Materials  Regulations - General Information, Regulations  and Definitions."
  U.S. Government Printing Office, Washington, D.C.

U.S. Department of  the Army, Headquarters. Medical  Support - Nuclear/Chemical
  Accidents  and Incidents.  AR 40-13, Washington, D.C.

U.S. Department of  the Army, Headquarters. Safety - Accident Reporting and
  Records.  AR 385-40, Washington, D.C.

U. S. Department of the Army, Headquarters.   Safety  - Ionizing Radiation
  Protection (Licensing, Control, Transportation, Disposal, and Radiation
  Safety).AR 385-11,"Washington, D.C.

U.S. Espartment of  the Army, Headquarters, Army Materiel  Command.  DAP.CC"-'.
  Disaster Control  Plans, Annex E, "Radiological  Accident/Incident Control."
                                     11.34

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                                                               DAKCO.M-P 385-

U,S. Department of the Army and Defense Logistics  Agency.   Medical Services
  Control and Recording Procedures for Exposure to lonirinc F\eC'i:: lor anc
  Radioactive Materials.AR 40-K, DLAR 1000.26,  Washington,  D.C

U.S. Nuclear Regulatory Commission and Federal  Emergency Management Agency.
  i960.   Criteria for Preparation and Evaluation of Radiological  Emergency
  Response Plans.  NUREG-0654, FEMA-REP-1,  U.S. Government Printing Office,
  Washington, L.C.
                                     11.35

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                                                               DARCOM-P 355-1
                                  APPENDIX A

                            EXAMPLES OF CHECKLISTS
A.I   EMERGENCY PLAN CHECKLIST
Emergency Planning
	   Hes one person or organization been assigned the responsibility for
      emergency planning?
	   Does this person or organization possess the authority to accomplish
      the task?
	   Have the maximum credible accident and several of the most probable
      accidents been determined?
Emergency Classification
	   Is the emergency classification system consistent with potential hazards
      at the installation?
	   Have the existing or imminent conditions for each class been defined?
	   Are definitions of radiation range continuous but distinct for each class
      (no gaps or overlap in definitions)?
Emergency Organizations
	   Have all emergency response organizations been identified?
	   Has each organization been assigned its emergency responsibilities?
	   Has each key person within the organizations been assigned a
      responsibility?
	   Have enough people been assigned responsibilities so that the emergency
      plan can be carried out completely and efficiently?
	   Do all  individuals have sufficient training to carry out their
      responsibilities?
                                   11.37-A

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r.mergency Facr'uv and Equipment Identification
	  Are aV emergency response facilities identified  ar"~  fully described?
	  Have ell emergency resocr.se resources and equipment been identified?
	  Have all onsite and nearsite impediments to the response been identified,
     along with realistic suggestions on ways to minimize  their effects?
Emergency Plan Implementation
	  Do procedures exist for implementing the emergency plan?
	  Are personnel assignments and methods of implementation clear?
	  Do implementing procedures ensure that all organizations are manned at
     the "alert" stage?
	  Do procedures ensure that all support organizations will be notified
     promptly of emergency situations?
	  Are emergency action levels defined?
Emergency Response
	  Are all response procedures functional and easy to understand?
	  Is the  installation capable of assessing all possible radiological
     conditions that may exist onsite and offsite as a result of its
     operations?
	  Have corrective actions to mitigate an accident been identified?
	  Are recommendations for protective action established?
	  Are they consistent with the recommendations of offsite agencies?
Emergency Facilities and Ecuipment
	  Have all emergency response facilities and areas been described in  the
     plan?
	  Is the  ECC expected to be habitable through most accident situations?
	  Have all tools, assessment equipment, protective equipment, and other
     support equipment used in emergency response been described  in the  plan
     or in a procedure referenced in the plan?

                                   11.38-A

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                                                               DARCOM-P 3E5-1
 Agreements with Offsite Support Groups
 	  Have all  necessary agreements  been  made for offsite  support  and
      cooperation?
 	  Are the agreements specific and  the agencies  reliable?
 Re-entry and Recovery
 	  Is re-entry of evacuated  buildings  controlled?
 	  Have radiological  conditions been established under  which  buildings  may
      be re-entered for  return  to unrestricted use?
 	  Are key positions  in  the  recovery organization  identified  and  have the
      responsibilities associated with those  positions  been assigned?
 	  Is a communications system in  place?

 A.2  KINOR SPILL CHECKLIST
      In the event of  a  minor spill  of radioactive materials,  the  following
 checklist should be used.
 Immediate Actions
 	  Alert everyone in  the immediate vicinity of the spill.
 	  Have everyone  leave the room and assemble in a  nearby area such  as a
      hellwe;'.   Allow  no one to  leave the area without  a radiation survey.  If
      the  spilled material  is highly toxic, evacuate  the building  to an
      assembly area.   No re-entry should  be attempted without  health physics
      supervision.
 	  Call  for health  physics assistance.
 	  If the material  does  not present a  hazard through toxicity or  high dose
      rates, attempt to stop the  leak and contain the contaminant  with absorbent
      pads  or other barriers.  Try to minimize personnel contamination and
      exposure.
	  If large quantities of gaseous or highly volatile materials  have been
      released, promptly shut down all heating,  ventilation, and air condition-
      ing operations to prevent the material  fro/;? spreading.
                                   U  33-A

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DARCOM-P 385-1
	  Seal off the area with signs, rope, or locked  doors  unf'l  health physics
     assistance arrives.
Recovery of the Spill Area
	  Have health physics personnel supervise all  recovery and re-entry
     activities.
	  Ensure that all persons involved in the accident or  in recovery proce-
     dures are surveyed and decontaminated, if necessary, before release.
	  Try to determine the types and quantities of radioisotopes involved so
     that appropriate protection is used upon re-entry.
	  Establish a step-off pad at the entrance to the affected rooms.
     Enter the room with appropriate protective clothing and devices, includ-
     ing dosimeters.  Ensure that release of the material  is halted and that
     cleanup can be performed without personnel receiving unacceptable doses
     (evaluate the radiological hazards).
     Decontaminate the area, being careful not to spread contamination over an
     area larger than necessary.
     Collect contaminated waste in plastic bags as it is generated, for later
     disposal.
     Make a final survey of the room before it is released for use.
     Have dosimeters processed promptly.
                                  11.40-A

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                                                               DARCO.K'-P  .?..
                                  APPENDIX-B-

                               RESPONSE  ACTIONS
B.I  EXPOSURES TO INDIVIDUALS
     The magnitude of abnormal  radiation  exposures  is  not  always  apparent imme-
diately following an accident.   Radiation protection,  medice" ,  and administra-
tive decisions will  be based on a combination  of  all available  data.   However,
the immediate care of an injured individual  is of prime  importance.   Initially,
any severe physical  injuries (e.g.,  burns,  cuts,  or trauma)  are likely to be
more important than possible radiation injuries.   Therefore, the  extent of the
injuries and the mobility of the patient  should be  assessed  immediately, and
first aid and lifesaving actions should be performed.   Specific actions to be
taken if contamination of a wound or the  skin  accompanies  the physical injury
ere discussed below.  (See also Chap''    7.)
     In order to identify the response actions appropriate for individual radi-
ation exposures, it is useful to define three  categories:   external  exposure,
internal contamination, and external contamination.
External Exposure
     The level of action needed to respond to  an  external  exposure depends on
the magnitude of the dose received.   The  individual should be removed from the
work environment and an accurate assessment of exposure should be made.  Action
end investigation levels are defined in AR 40-14.  A summary of response
actions to various doses received by an individual  is  outlined in Table  11.9,
based on Publication 28 of the International Commission on Radiological  Pro-
tection (1978).
     An accurate dose estimation becomes  more  important as the dose gets
higher and can be accomplished through a  combination of clinical, biological,
biochemical, and physical assessments of the exposed individual.  The  informa-
tion pro1, ided by personal dosirr/rters. reconstruction of the event, end  icenti-
fication of radiation fields can be used to assess the dose.   In  the  case of
exposure to neutrons, activation products in or on the body  (e.g., in  the
                                    ll.Al-B

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\\kCOM-P 385-1
            TABLE 11.9   Summary of Response  Actions  for  Individual
                         External  Exposure
       Dose         	Response Actions	
      5-10 rem      Administrative actions,  investigation
                    Physical dose measurements
     10-25 rem      More detailed administrative  investigation
                    Assessment of possible biological  consequences
                    Physician brought in to assess  the need for and the
                    extent and nature of clinical,  biological,  or bio-
                    chemical examinations
       >25 rem      Same as above, plus an examination by  the physician

blood, on the hair, or on metal objects such as coin.s  or watch  bands)  can also
aid in .this assessment.  Observable clinical symptoms  such as nausea and
vomiting would appear in approximately 10* of individuals  exposed to 75  to
125 rem.
     Priorities for treatment, and response actions for individuals subjected
to whole-body exposures, are given in Table 11.10.
Internal Contamination
     If an intake is suspected, first aid should be given  immediately,  the
nature and degree of contamination should be determined, and therapy procedures
should be started under the direction of a physician.
     The initial indications for therapy include the first dose assessment and
the results of nose blows and of monitoring for skin contamination, contami-
nated wounds, and, if appropriate, air and surface  contamination.  Examples of
types of therapy to consider are:  1) isotopic dilution of an ingested radio-
active substance by the administration of a stable  isotope (e.g., administra-
tion of stable iodine, as sodium iodide or potassium iodide, to reduce the
deposition of radioiodine in the thyroid gland); 2) acceleration of excretion
through the administration of a laxative to minimize gastrointestional  absorp-
tion; and 3) adnrir.istratic;  of irritants or expectorants to minimize respiratory
absorption.  Actions to be taken following a suspected internal contamination
are presented in Table 11.11.
                                   11.42-B

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                                                       UARCOX-P 385-1

TABLE 11.10.   Actions  to  be  Taken  Within  Six Hours  Following  i
              Whole-Body  Exposure

  Medicrl  Management

       Administer lifesaving treatment
       Check  for external  contamination
       Remove clothing and wash  contaminated areas
       Give mild sedative  for nausea  and  vomiting

  Clinical  Observation

      ' Collect dosimetric  data
       Interrogate patient about accident and  relay information
         to dosimetry  team
       Make tentative  prognosis  based on  above findings

  Biological  Investigations

       Take and keep  urine samples
       Take blood samples  for immediate cell counts,  biochemical
         analysis, lymphocyte culture, and chromosomal  analysis

  Dosimetric  Studies

       Process all personal  dosimeters from exposed individual
         and  bystanders
       Check  installed recording equipment in  vicinity  of accident
       If neutron exposure is suspected,  measure  induced activity
         using coins  the  exposed person was carrying
       Make first assessment of  likely type, quantity,  and distri-
         bution of radiation, and  inform  physician
       Interrogate bystanders

  Administrative Actions

       Perform detailed inquiry  into  the  circumstances  of the
         accident
  (a)  Excerpted  from ICRP  28.
                            11.43-B

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UARCOM-P 385-]
        TABLE 11.11.  Actions to be Taken Following Suspected Internal
                      Contamination

          Medical Management

               Preliminary therapy (under the direction of a physician)
               - isotopic dilution
               - expectorants
               - laxatives
               - chelating agents

          Biological Investigations

               Take swabs from the nose or mouth
               Perform whole-body count
               Collect urine and fecal samples
               Take blood sample

          Dosimetric Studies

               Confirm intake
               Check installed air monitors
               Make direct measurements using an external or wound
                 probe and an organ scanner
               Perform radiochemical assay of urine, fecal, and
                 blood samples

          Administrative Actions

               Perform detailed inquiry into the circumstances of' the
                 accident
                                    11.44-B

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                                                               DARCOH-P 365-1
 External  Contamination
      External  contamination  can  involve  both en external dose and internal
 contamination.   First aid  (including decontamination procedures) should be
 given immediately,  and the dose  received and the extent of contamination
 should be assessed  promptly.
      The  individual  should be decontaminated as effectively as possible before
 being taken  to the  hospital.  Chapter 7  desci t>es personnel decontamination
 procedures  in  detail.  A few simple procedures are mentioned here.  Skin areas
 are  decontaminated  by washing them with  soap and large amounts of water.  The
 contaminated individual can often do this.  Measurements of residual contamina-
 tion should  be taken after each  washing.  However, this treatment should ceese
 before skin  abrasions appear.  The eyes, nose, and mouth can be decontaminated
 by flushing  them with large quantities of water.  Contaminated wounds  should
 immediately  be washed with large quantities of water, and bleeding  should be
 encouraged.  The use of a chelating agent is recommended.  All of the  pro-
 cedures except for  the washing of skin areas require the supervision of
 medical personnel (see Chap-.   7).

 B.2   TRANSPORTATION ACCIDENTS
      Accidents that occur during the shipping of radioactive materials may
 require the  involvement of state and local authorities, and/or the  DOT.
 Appropriate  responses to the accident include the following actions:
  1.   Administer first aid to seriously injured persons and summon a rescue
      squad.
  2.   Confine contamination to the local  area; an exclusion area may be
      established.
  3.   Locate  people  along the shipping route who may have been exposed  or
      contaminated.
      Federal interagency radiological assistance can be obtained  by calling
 tr.--  Joint Nuclear Accident Coordinating  Center at Kirtland Air  Force  Base
 (Commercial  (505)264-8279 or 'AUTOVON 964-6279).
     The nearest Army facility may also be called upon for assistance. Table
£-1  of AB  385-11 lists  Army addresses and emergency telephone numbers.
                                   11.45-B

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                                                             DARCOM-P  365-1
                                  APPENDIX C
                  EXAMPLE LISTING OF EMERGENCY  KIT  EQUIPMENT
                   Items
                                                       Date  checked_
                                                       Checked  by
Quantity
Box 1
Box 2
Protective clothing
     Coveralls
     Neoprene qloves
     Disposable cloves
     Head covers
     Resoirator cartridoe
           Chemical
           Participate
     Masking tape
Posting equipment
     Radiation rope
     Radiation sions
     Radiatior, labels
     Radiation tape
     Masking tapie_
     Twine
                                    11.47-C

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DARCOM-P 3E5-1
                     Items
Quantity
Box 1
Box 2
 Tools
      Scissors
      Tones  (46  or.)
       Extension
      Channel-lock  pliers
      Screwdriver
      Raci-.r  liaht1
       Knife
 Surveying  and  sampling  supplies
      Cotton  swabs
      Disposable  bottles
      Large  plastic bottles
      Scintillation vials
      Air-samolino filters
      Air-samcling  cartridges
       Plastic bags
            Larae
            Small
 Decontamination aids
      Deteroent
      Cleanser
      Gauze  pads
                                     11.48-C

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                                                              DARCO.M-P  3S5-1
                   Items                           Ouantitv      Box  I      Box  2
Miscellaneous
     Adhesive tape
     Pencils
     Notepads
     Butcher paper
     Stopwatch
     Extra batteries
Readily Available Equipment
     Survey meters
          lonizetion chamber
          Geiger-Mueller counter
     Air samplers
     Aloha detector
     Fast- and slow-neutron meters
     High-range pocket dosimeter
     Spare film badges
     Small fire extinouishe
     Portable power source
     First eid kit
                                    1J.49-C

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                                                              DARCON-P 365-1
                                  APPENDIX D

                         EXAMPLES OF EMERGENCY ACTION'S
D.I  AMBULANCE OR RESCUE SQUAD PERSONNEL
Guidelines for handling patients contaminated with radioactive materials:
 1.  Give lifesaving emergency assistance if needed.   '
 2.  If a health physicist is immediately available,  have him or her ride with
     the patient in the transport vehicle.
 3.  Cover the stretcher and pillow with an open blanket; wrap the patient in
     the blanket to limit the spread of contamination.
 4.  Call the appropriate hospital  by radio or telephone and provide available
     information.
 5.  Save all materials suspected of being radioactively contaminated in
     plastic bags or containers labelled with patient's  name, date, and time.
 6.  Ensure that rescue squad personnel and equipment are monitored upon
     arrival at the hospital.

D.2  HOSPITAL EMERGENCY ROOM PERSONNEL
Upon notification of the imminent arrival of a contaminated patient, the
following actions should be tatcen:
 1.  Notify responsible staff physician, hospital administrator, and health
     physicist.
(a) Note   f-'edicc":  treatment takes precedence over personnel decontamination
    and/or contamination control.
                                    11.51-D

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DARCOK-P
  2. Take-precautions to prevent the spread of contamination:
     a.    Prepare a separate space, usir.g an isolation  room  or  cubicle  if
           available.
     b.    Cover the floor with absorbent paper.
     c.    Mark and close off the area.
     d.    Prepare to shut off air circulation system,  if  dust  is  involved.
  3. Obtain appropriate survey meter.
  4. Put on protective clothing.

 Upon arrival of the patient:
  1. If patient is seriously injured, give emergency  lifesaving  assistance
     immediately.
  2. Have health physicist check patient for contamination using survey meter.
     Record patient's name, date, time,  location  and  extent of  contamination,
     and radiation measurements.
  3. If external contamination is involved, save  all  clothing and bedding from
     ambulance, all metal objects (jewelry, belt  buckles), and  all blood,
     urine, stool, and vomitus, and label with patient's name,  date, and time.
     Store in plastic bags or containers marked "Radioactive -  Do Not
     Discard."
  4. Begin decontamination procedures (if patient's medical status permits) by
     cleansing and scrubbing the area of highest  contamination  first, using soap
     and warm water; showering may be necessary.   Resurvey and  record measure-
     ment after each washing or showering.  If a  wound is  involved, use self-
     adhering disposable surgical drape to cover  it,  then  cleanse neighboring
     skin surfaces and seal with surgical drape.   Remove the wound covering and
     irrigate the wound with sterile water, catching  the water in a basin marked
     "Radioactive - Do Not Discard."
  5. Save physicians', nurses', arc attendants' srrub or protective clothing.
     Follow monitoring and decontamination procedures.
                                     11.52-D

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                                                              DARCOX-P  385-.
D.3  FIREMEN
Special  precautions must be taken in fignting a fire involving radioactive
materials:
 1. Identify and isolate the hazard.
 2. Contact e health physicist for guidance and assistance.
 3. Stay upwind from the fire.
 4. Wear self-contained breathing apparatus and full protective clothing.
 5. Limit time spent in hazard area to shortest possible time.
 6. Avoid contact with leaking or damaged packages.
 7. Fight fire from as far away as possible.
 8. Move undamaged packages out of the fire zone if this can be done with no
    risk.
Additional information may be obtained from the National Fire Protection
Association (NFPA 801-Recommended Fire Protection Practice for Facilities
Handling Radioactive Materials).
                                    11.53-D

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                                                              DARCO.M-P  385-1

                             CHAPTER 12.   TRAINING
12.1 TRAINING FOR RADIATION WORKERS     	    12.3

     12.1.1  Frequency of Training 	    12.4

     12.1.2  Course Content   	    12.4

             A.  Radiation Biology and the Risk from
                 Occupational Exposure  .......    12.4

             B.  Radiation and Radioactive Material     ....    12.4

             C.  Measurement end Control of Radiation Exposure
                 and Radioactive Material     	    12.6

             D.  Radiation Protection Program     .....    12.6

             E.  Emergency Preparedness .     .     .     .     .     .     .12.6

     12.1.3  Use of Mockup Facilities   	    12.7

     12.1.4  Evaluation of Trainee Performance    	    12.7

     12.1.5  Documentation of Training  	    12.8

12.2 INSTRUCTION TO WOMEN OF REPRODUCTIVE CAPACITY     ....    12.8

     12.2.1  Recommended Prenatal Occupational  Exposure Limit    .     .    12.9

     12.2.2  Requirements     	    12.9

     12.2.3  Rationale for Limit   	    12.10

12.3 INSTRUCTION IN THE USE OF RESPIRATORS	12.11

     12.3.1  Extent of Training    	    12.11

     12.3.2  Contents of Training Program    	    12.12

     12.3.3  Drills 	    12.13

12.4 TRAINING FOR MANAGERS    	   -12.13

     12.4.1  Frequency of Training ........   12.13

     12.4.2  Contents of Training Program    ......   12.13
                                     12.1

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DARCCM-P 385-x

 12.5 TRAINING FOR THE RADIATION-P-ROTECTIOK' STAFF	12.14

 REFERENCES	12.14
                                     TABLES
  12.1 Appropriate Subjects  for  a Radiation  Protection
      Training  Program     ..........    12.5
                                       12.2

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                                                             DAKCOM-P  3S5-1

                             CHAPTER 12.   TRAINING
     The Radiation Protection Officer (RPO)  is  responsible  for  conveying  to
all staff members policies and procedures relating  to  radiation safety.   The
extent and breadth of the training needed varies  significantly  with job  require-
ments and responsibilities.  For a clerk, a  brief description of the working  •
environment, an explanation of protective measures  to  be  taken  in case of an
emergency, and an assurance of personal  safety  may  be  sufficient.   For a  han-
dler of radioactive weste, extensive formal  training  is  required.
     Information on the radiation safety policy should be presented during the
new staff member's orientation.  At that time,  a  general  introduction tc  the
radiation hazards associated with the work should be  given.   Radiation hazards
and related safety programs should be presented not as unique or special
entities, but rather as part of the overall  program for occupational health.
Written material on these topics can be an invaluable  resource  for distribu-
tion to new employees.
     This chapter describes the training that should  be presented to radiation
workers, women of reproductive capacity, users  of respirators,  managers,  and
radiation protection personnel.
                 Section 12.1  TRAINING FOR RADIATION WORKERS

     The term "radiation worker" is synonymous with the term "occupationelly
exposed individual."  A radiation worker is an individual whose work is per-
formed in a radiation area or a controlled area and who might be exposed to
more than 5* of the basic radiation protection standard listed in Chapter 3,
Table 3.2, of this manual (see also AR'40-14) as a result of duties in these
areas.  Radiation worker training should be extended to all individuals who
work in radiation areas or controlled areas even if they do not work directly
with r-adicactive material.  For example, fire fighters, security forces,
emergency response personnel, janitors, and night guards who may need to enter
a radiation or controlled area during the course of their work should receive
                                     12.3

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l'AACOM-P  363-1
 radiation training,  es  should  those  assigned  to work  full  time  m  these  areas.
 Required instruction *or workers  is  detailed  in 10  CFR  19.12.
 12.1.1  Frequency of Training
      Individuals should receive training  before entering  or  beginning  work  in
 a radiation or controlled area.   They  should  be retrained annually or  whenever
 policies or procedures  are changed.
 12.1.2  Course Content
                 i
      The training program should  include  the  subjects  listed in  Table  12.1  and
 discussed below.  The topics emphasized will  vary with  the needs  of each
 individual  or group  being trained.   Each  individual's  work assignment  and  the
 standing operating procedures  (SOPs) covering the assignment should be care-
 fully reviewed to determine the scope  of  training needed.  Appropriate refer-
 ence documents covering essential  facts,  requirements,  regulations, procedures,
 and plant organization  should  be  given to each individual.
      A.   Radiation Biology and the Risk from  Occupational  Exposure.  Persons
 who work in or near  radiation  and controlled  areas  or make decisions about
 work in  those areas  should be  taught enough about radiation  effects to appre-
 ciate the importance of keeping exposures as  low as is  reasonably achievable
 (ALARA).  These individuals should be  informed of the level  of radiation dose
 anticipated in their work area and the risk associated with  such a dose level.
 Appropriate topics could include  dose-effect  relationships for internal  and
 external radiation and  the collective-dose concept  of risk (individual and
 group) as it applies to the ALARA philosophy.
      B.   Radiation and  Radioactive Material.   Types of radiation and their
 characteristics should  be discussed  to the extent necessary  to explain the
 nature of the material  people  work with.   Types and forms of radioactive
 materiel should be detailed so that  staff members understand proper control
 procedures.  Sources and origins  of  radioactive material  and radiation onsite
 should be identified, as should the  signs and labels used to mark this
 material.
                                      12.4

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                                                             DARCOM-P 3£3-i

TABLE 12.1.   Appropriate Subjects for a Radiation  P-otection Training Program

1.  Radiation Biology and the Risk from Occupational  Exposure

    a.   Dose-effect relationship
         (i)  External  radiation
         (2)  Internal  radiation
    b.   Collective-dose concept
         (1)  Group total man-rem risk
         (2)  Individual dose risk

2.  Radiation and Radioactive Materiel
    a.   Types of radiation and their characteristics
    b.   Types and forms of radioactive materials
    c.   Sources (origins) of radioactive materials and radiations onsite
    d.   Source identification
                                »
3.  Measurement and Control of Radiation Exposure  and Radioactive Material

    a.   Dosimetry
    b.   Maximization of distance between people and radiation sources
    c.   Shielding
    d.   Detection and  control of contamination, and decontamination
    e.   Radiation measurement and survey instruments
    f.   Area and air monitoring
    g.   Personnel monitoring
         (1)  Internal
         (2)  External

4.  Radiation Protection Program

    e.   Radiation protection standards, guides, and limits
    b.   ALARA program
    c.   Responsibilities of individuals
    d.   Radiation areas at the site
    e.   Signs and labels
    f.   Control of radiation areas
    g.   Investigation  and reporting of abnormal exposures
    h.   Radiation survej's--purpose and methods
    i.   Protective apparel
    j.   Respirators and their use
    k.   Rules and procedures, including standing operating procedures
    1.   Professional  guidance and assistance
    m.   Control and removal of contamination and contaminated equipment

5.  Emergency Preparedness

    a.   Plant safety and accident control features
    b.   Signals and alarms
    c.   Evacuation routes and procedures
    d.   Assembly points
    e.   Communications
    f.   Emergency equipment
    g.   First aid and  treatment of contaminated wounds
                                    12.5

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DARCOM-P 385-.
      C.  Measurement and Control  of Raclation Exposure  and  Radioactive Mate-
 rial .   Each radiation worker should be informed that  radiation  and radioactive
 materials can be measured at levels significantly below radiation protection
 standards and controlled by means of suitable design  and procedural  techniques.
 Radiation workers should understand the elements of radiation measurement and
 control well 'enough to participate in an effective radiation  protection pro-
 gram consistent with the ALARA philosophy.   Emphasis  should be  on 1) the
 sources of radiation, 2) contamination control, 3) the  use  of  time,  distance,
 and shielding to, reduce doses, 4) SOPs, and 5) the proper use  of dosimeters.
 The importance of administrative  and engineered controls and  the performance
 of work in accordance with carefully planned procedures should  be stressed.
      0.  Radiation Protection Program.  Personnel should understand the nature
 and scope of "the radiation protection program, including pertinent portions of
 regulations, site rules for radiation protection, and safe  operating procedures.
 Emphasis should be placed on the  ALARA philosophy, its  objectives, and its
 implementation within the framework of the tasks to  be  performed.  The respon-
 sibility of the radiation protection staff in implementing  ALARA goals, and
 the responsibilities of the individual staff member  within  the ALARA program,
 should be understood.
      At the completion of the training program, radiation workers should under-
 stand that personnel outside radiation and controlled areas should not be
 significantly affected by activities in these areas  that involve radioactive
 materials or radiation.  The meaning and importance  of posted instructions,
 including radiation warning signs and tags, and the  importance of following
 instructions should also be understood.
      E.  Emergency Preparedness.   Staff members should know the  appropriate
 response to alarms and signals.  They should be familiar with the details  of
 emergency procedures and preparations so they will know what is  expected of
 them and from whom they can expect guidance in an emergency.  They  should  know
 the locations of emergency facilities and equipment  as well as emergency
 escape routes and safe assembly points.  Preparations for possible  emergencies
 should be emphasized; such emergencies should  include accidents  Involving
                                      12.6

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                                                              DARCOM-P 385-1
severe personal contamination, contaminated wounds,  and  localized fires in
radiation and controlled areas.
12.1.3  Use of MOCNUP Facilities
     The use of equipment or facility mockups allows individuals to practice
procedures in a realistic setting before they perform the procedures using
radioactive materials or enter areas where a potential  for exposure to
radioactive contamination exists.  This type of training is especially
valuable for repair and maintenance tasks that could result in high doses to
personnel in relatively short periods of time.  Another valueble application
is in research laboratories where radioisotopes are used.
12.1.4  Evaluation of Trainee Performance
     Each radiation worker's knowledge, competency, and understanding of the
radiation safety aspects of specific jobs should be-evaluated.  The evaluation
may consist of only a written or oral test, but should,  in most cases, include
a written test, an oral test, and a "practical" or on-the-job performance
test.  The questions asked and the responses given in all examinations should
be documented.  Requalification testing should be conducted in conjunction
with refresher training.
     High test grades (i.e., 80» or higher) should be required because each
person's training covers radiation protection information relevant to the
person's needs and safety in the work environment.  Radiation workers should
be reinstructed and retested in any areas in which their knowledge is shown to
be deficient.
     Tests should cover ell the information presented during training but
should emphasize tne day-to-day radiation protection practices relevant to
each person's job.   As experience is gained, test questions should reflect
the radiation protection problems actually experienced onsite.
     Practical  or on-the-job tests should stress knowledge and proper job
performance.   A person may know what to do but be unable to do it promptly
when faced with a situation demanding immediate end effective action.  In
preparing a  test, consideration should be given to individual job responsi-
bilities,  training  received, and radiation protection experience.
                                     12.7

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DARCOM-P 385-1

      Tests should be designed to:
  1.  measure the person's ability to recognize and cope  with  radiation
      hazards that may be encountered on the job
  2.  stress preparedness for work in radiation and controlled areas
  3.  assess the individual's knowledge of and attitude  toward his or her
      rights and obligations regarding radiation protection
  4.  assess the individual's understanding of control  procedures.
                  i
 12.1.5  Documentation of Training
      Records that describe the content of training courses,  such as course
 outlines, syllabuses, brochures, video tapes, texts,  and tests,  should be
 maintained.  These records serve as a basis for determining  the  depth and
 scope of training given in each subject area.  Trainee-specific  training
 records, which provide a complete history of each person's training experi-
 ences, should also be maintained.  A complete description of information to be
 included in the training records is given- in Chapter 13, "Recordkeeping."
      A staff member who has been trained at one site and is  later to be
 employed at a different site should receive a statement of training received.
 This statement will allow the person responsible for training at the second
 site to take the staff member's previous training into account and thereby
 avoid needless repetition of training.  The statement should clearly and
 explicitly describe all training received and should identify non-plant-
 specific training segments that may be applicable to work in the new
 position.
           Section 12.2  INSTRUCTION TO WOMEN OF REPRODUCTIVE CAPACITY

      A special situation arises when an occupationally exposed woman is preg-
 nant.  Exposure of the woman's abdomen to penetrating radiation from either
 external or internal sources would also expose the embryo or fetus.  A number
 of studies have indicated that the embryo or fetus is more radiosensitive than
                                      12.8

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                                                              DARCOM-P 385-1
an adult, particularly curing the first  3  months  after  conception,  when a
woman may be unaware of her pregnancy.
12.2.1  Recommended prenetal  Occupational  Exposure Limit
     The National  Council  on  Radiation Protection and Measurements  (NCRP)
recommends in its  Publication 53 (1977)  that,  because the  unborn  are more
sensitive to radiation than adults,  their  radiation  dose from  occupational
exposure of the mother should not exceed 0.5  rem.  The  International Commis-
sion on Radiological Protection (ICRP) recommends in its Publication 9 (1965)
that the occupational  radiation exposure of all women of reproductive capacity
be received gradually, in  small increments, so that  an  unborn  baby  would be
unlikely to receive more than 0.5 rem in the  first 2 months  after conception,
when a woman may not be aware that she is  pregnant.
12.2.2  Requirements
     All individuals who work in a restricted area must be instructed as to
the risks associated with  radiation  exposure  (Nuclear Regulatory  Commission
(NRC) Regulatory Guide 8.13 (1975)).   This instruction  should  include informa-
tion on the risks  to the unborn.  Women  should be encouraged  to  inform the
RPO of a pregnancy.  Every effort should be made  to  limit  the  dose to an
embryo-fetus to 0.5 rem during the entire  gestation  period.   The  mother's
exposure should be as  uniformly distributed over  time as  is  practicable.
     The establishment of  differential occupational  exposure limits for men
and women can raise a  number  of social and legal  questions.   All  alternatives
should be considered before the situation  arises.  Options include the
following:
 1.  The dose to the unborn child can and  should  be  reduced  by a) decreasing
     the time the  woman spends in radiation areas, and/or  b) increasing the
     distance between  the  woman and  the  source of radiation, and/or
     c) shielding  the  abdominal area  (the  use of  lead aprons could be
     considered),
 2.  The woman  can be  reassigned to  an area or job  involving less radiation
     expcsurt.
 3.  The woman  can be  reassigned to  a nonradlatlon  position.

                                     12.9

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DARCOM-P 385-1
 All available options should be discussed with the  expectant  mother.   It  is
 important that a decision be reached quickly,  as  the  unborn child  is  most
 radiosensitive during the first 3 months of pregnancy.
 12.2.3  Rationale for Limit
      The radiosensitivity of cells (their susceptibility to damage by radia-
 tion)  is directly related to their degree of differentiation, that is,  to the
 extent to which they have developed distinct and  identifiable functions.
 Kidney cells, for example, have a function different  from that of  cells of the
 eye.   Because most cell differentiation takes  place in newly  forming  and
 growing beings, embryos are more radiosensitive than  fetuses, fetuses more
 radiosensitive than children, and children more radiosensitive then adults.
 This principle has long been a factor in the development of  radiation protec-
 tion standards, as exemplified by the difference  in the exposure limits for
 minors and adults:  the occupational radiation exposure of anyone  under the
 age of 18 cannot exceed 10% of the limits for adult workers.
      The development of a baby is usually divided into three  stages:   ovum,
 embryo, and fetus.  An ovum becomes an embryo about 7 days after fertiliza-
 tion;  the embryo stage lasts approximately 8 weeks; and the  fetal  stage is the
 time remaining until birth.  The particular effect of radiation, and its
 severity, depend on the stage of development at which exposure occurs.   An
 unborn child is more sensitive to radiation during the embryonic stage than in
 the earlier or later stages of development.  During this period, the organs
 are being formed and the cellular organization cf the embryo is changing
 rapidly.  Cells become specialized and start processes leading to the
 development, in a fixed sequence, of specific tissues.  Consequently, the
 effect of radiation varies from day to day, and different degrees and kinds of
 organ  malformations are produced depending on exactly when the exposure
 occurs.
      During the earlier or ovum stage, relatively few cells  are present, and
 the most common effect of exposure to radiation is chromosomal injury leading
 to cell death.  During the later or fetal period, most organs have already
 been formed, and malformations from radiation exposure are less common and
                                      12.10

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                                                              DAKco::-? '-,L:>-i
less severe.  The major radiation effect during this period is re_ducec orowtr,,
which may persist throughout life.
     The genetic end cancer risks per unit of radiation cose fror ir,-i,-_e"c
exposure also exceed those from adult exposure.  The undeveloped ovum cells in
the female fetus ere actively dividing and are nearly as sensitive as the
male fetus's imn.cture sperm cells.   The most sensitive period for generic
damage in both sexes is probably the lest 6 months before birth.
     The leukemia risk from in-utero exposure has been estimated as beinc 10
times greater than that for adults  who get the same dose.  The follow-up
period for solid tumors, which have a longer latency period than leukemia, has
probably not been long enough to allow a good estimate of the total risk for
other cancers caused by in-utero exposures.  The absolute risk of getting
fatal cancer, other than leukemia,  in the first 10 years of lifr from in-utero
exposure, however, has been estimated as five times the risk that an adult has
of getting cancer within 10 years of receiving the same exposure.  For all of
these reasons, the occupational radiation exposure of pregnant women should be
limited.
              Section 12.3  INSTRUCTION IN THE USE OF RESPIRATORS

     Training in the use of respirators should be given by a qualifier  end
experienced instructor, such as a health physicist, industrial h.ycjisr.is'., or
safety engineer.  The instructor must have a thorough knowledge of  the  f.cp'!icc-
tion and use of respirators and of the hazards associated with rscno?ci1ve
airborne contaminants.  He or she also must have had considerable ex.^r'iei-.r-:
in the practical selection and use of respirators for protection apainst
radioactive airborne contaminants.
12.3.1  Extent of Training
     The instructor should develop an adequate training program ••?..<•,£•:.  on  the
hazards that may be encountered and the types of respirators to be  v-vr,.
Training must be given not only to the persons who will perform work  usv/ij  the
respirators but also to those who will direct the work.  Especially '.vnere
                                     12.11

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DARCOM-P 385-1
 respirators are used only occasionally, staff members  should be retrained
 often enough so that a high degree of proficiency is  retained when respiratory
 equipment is actually used.
 12.3.2  Contents of Training Program
      Training in the use of any respirator must cover  at least the following
 topics:
  1.  the nature of the airborne contaminants against  which the wearer is to be
      protected, including their physical properties,  maximum permissible
      concentrations, physiological action, toxicity,  and means of detection
  2.  the construction, operating principles, and limitations of the respirator
      and why the respirator is the proper type for the particular purpose
  3.  the reasons for using the respirator and why more positive control of
      airborne contamination is not immediately feasible, including recognition
      that every reasonable effort is being made to reduce or eliminate the
      need for respirators
  4.  procedures for ensuring that the respirator is in proper working
      condition
  5.  how to fit the respirator properly and how to check the adequacy of the
      fit
  6.  the proper use and maintenance of the respirator
  7.  application of available cartridges and canisters for air-purifying
      respirators
  8.  what emergency action to take if the respirator malfunctions
  9.  radiation and contamination hazards, and other protective equipment that
      mey be used with respirators
 10.  classroom and field training in recognizing and coping with  emergency
      situations
 11.  other -oecial trair.ing as needed for special purpose's.
                                      12.12

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                                                              DARCO.N-P 3E5-1
12.3.2  Drills
     Training should include actual  use of respirators  under simulated condi-
tions of exposure so that the wearers develop a  sense  of confidence in their
ability to use the devices properly.   A qualified  observer should review with
the trainees their performance in these drills.
                      Section 12.4  TRAIN'ING FOR MANAGERS

     Managers need to be knowledgeable in all  radiation safety policies and
procedures and to understand the ALARA philosophy.   They should know who the
members of the radiation protection staff are  and how to contact them.
12.4.1  Frequency of Training
     Managers should be offered training when  they move into a position which
requires that they oversee work with radioactive materials.   This training can
often be done on a one-to-one basis.  Retraining should be provided whenever a
change in policy is made.  A presentation at a regularly scheduled staff
meeting is a convenient way to provide retraining.
12.4.2  Contents of Training Program
     Training for managers should include the  following topics:
 1.  basic radiation safety and radiation biology;  sufficient detail should
     be provided to allow an understanding of  the ALARA program
 2.  site-specific radiation program
 3.  responsibility of the manager
 4.  responsibility of staff members
 5.  emergency preparedness.
                                     12.13

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DAHCO>2-P 385-1

            Section 12.5  TRAINING FOR THE RADIATION PROTECTION STAFF


      The responsibilities of the RPO and the radiation protection staff were

 detailed in Chapter 3.   Members of the radiation protection staff need

 training that will prepare them to meet those responsibilities and to maintain

 proficiency in their duties.  Contact DARCOM Headquarters for assistant in

 identifying appropriate short courses.
                                   REFERENCES
 International Commission on Radiological Protection (1CRP).  1965.  Recommenda
   tions of the International Commission on Radiological Protection.  ICRP 9,
   Pergamon Press, Oxford.
 National Council on Radiation Protection and Measurements (NCRP).  1977.
   Radiation Dose Limit for Embryo and Fetus in Occupational^ Exposed Women.
   NCRP 53, Washington, D.C.

 U.S. Code of Federal Regulations.  1982.  Title 10, Part 19, "Notices,
   Instructions and Reports to Workers; Inspections."  U.S. Government Printing
   Office, Washington, D.C.

 U.S. Department of the Army and Defense Logistics Agency.  Medical Services -
   Control and Recording Procedures for Exposure to Ionizing Radiation and
   Radioactive Materials.  AR 40-14, Washington, D.C.

 U.S. Nuclear Regulatory Commission (NRC).  1975.  "Instruction Concerning
   Prenatal Radiation Exposure."  Regulatory Guide 8.13, Washington, D.C.
                                      12.14

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                                                              DARCOM-F 385-1

                          CHAPTER 13.   RECORDKEEPIN5




13.1 RADIATION RECORDS FILES 	    13.6

     13.1.1  Personnel File	13.7

             A.   Identification of the Individual      ....    13.7

             B.   Training Records ........    13.7
               t
             C.   Project/Task Listing   	    13.7

             D.   External-Exposure Records  ......    13.8

             E.   Internal-Exposure Records  ......    13.9

             F.   Radiation Exposure Received During Prior Employment  .    13.11

             G.   Exposure Received by  Individuals  at Other
                 Installations During  Current Employment   .     .     .    13.11.

             H.   Simultaneous Employment at Another Facility    .     .    13.12

             I.   Exposure Evaluation   .......    13.12

             0.   Unusual Exposures     .......    13.12

             K.   Transfer of Records   .    .    .    .    .     .     .    13.13

     13.1.2  Radiation Protection Program File   	    13.13

             A.   Licenses and Authorizations     	    13.13

             B.   Radiation Protection  Policies and Standards     .     .    13.13

             C.   Documents of the Ionizing Radiation Control
                 Committee   .    ,    .    .    .    .    .     .     .    13.14

             D.   Procedures for Obtaining and Evaluating Data on
                 Individual Exposures   	    13.14

             E.   Inspections and Appraisals      .    .    .     .     .    13.14

             F.   Changes in Procedures and Methods    .    .     .     .    13.15

     13.1.3  Project File	13.15

             A.   General Records	13.15
                                     13.1

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DARCOM-P 385- :
              B.   .iidncnng Operating Procedures   .     .     .     .     .     13.15
      13.1.4  Radiation Work Area File	13.16
              A.   General Records	23.16
              B.   Radiation and Contamination Surveys  	     13.16
              C.   Area Monitoring Records	13.17
              D.   Airborne-Radioactivity Monitoring  Records .     .     .     13.18
                 i
      13.1.5  Instrumentation and Dosimeters File 	     13.16
              A.   Capabilities of Dosimeters and Instruments     .     .     13.18
              B.   Calibration and Maintenance     	     13.19
              C.   Inventory Records     	     13.19
      13.1.6  Radioactive-Material Inventory File 	     13.20
              A.   Sealed and Unsealed Sources and Radioactive
                  Commodities	13.20
              B.   Environmental Samples 	     13.21
      13.1.7  Waste Management File     	     13.22
      13.1.8  Transportation File  	     13.22
      13.1.9  Accidents/Incidents File  	     13.23
      13.1.10 Training File	     13.23
      13.1.11 Quality Assurance File    	     13.24
 13.2  RECORDS FILING SYSTEM  	     13.24
 13.3  RECORDS RETENTION AND STORAGE    	    13.25
      13.3.1  Types of Records Retention     	    13.25
              A.   Hard Copy	13.25
              B.   Computer Records 	    13.26
              C.   Microform	13.25
              D.   Combinations     ........    13.26
                                      13.2

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                                                              DARCOM-P  385-1

     13.3.2  Retention Period •    	     13.26

     13.3.3  Storage Precautions  	     13.27

REFERENCES	13.27

APPENDIX A - SAMPLE RECORDS FORKS 	     13.29

APPENDIX B - OCCUPATIONAL RADIATION EXPOSURE FORKS    .     .     .     .     13.37

APPENDIX C - CROSS-REFERENCE SYSTEM FOR FILES, AND
             FLOW CHARTS FOR PROBLEK SOLVING     	     13.43



                                    TABLES

13.1 Cross-Reference System  	     13.45
                                     13.3

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                                                             DARCOM-P  385-1
                          CHAPTER 13.   RECORDKEEPING
     Good recordkeeping is essential  in the radiation  work  environment.   Accu-
rate records and a filing system that incorporates  extensive  cross-referencing
can help the Radiation Protection Officer (RPO)  and the  installation  commander
achieve the following:
 1.  plan an individual's occupational  exposure,  keeping in mind the  ALARA
     philosophy (maintaining radiation  exposures  as low  as  is reasonably
     achievable)
 2.  demonstrate good management practices in the handling  of radioactive
     sources
 3.  demonstrate compliance with government regulations  and the site's Nuclear
     Regulatory Commission (NRC) license
 4.  evaluate the effectiveness of the  radiation  protection and quality assur-
     ance programs
 5.  trace the cause of a trend of elevated doses
 6.  document, for both legal and medical purposes, the  exact conditions under
     which an individual  received a particular radiation dose (i.e.,  what the
     radiation source was, its activity or probable concentration, and when and
     how the individual was exposed).
This chapter describes the content and  form of the radiation  work records that
must be maintained in accordance with the requirements of the Department of the
Army (DA) and the following parts of the U.S. Code of Federal Regulations:
Title 10, Parts 19 and 20, and Title 29, Parts 570.57 end 1910.96.  A method of
organizing these records  into a filin^  system that would provide easy access to
all records pertaining to an individual, a specific project,  a radioactive
source, a radiation work  area, or a particular radiation-measuring instrument
is  also described.   The chapter closes  with a section on retention and storage
of  records.
                                     13.5

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DARCOM-P 385-1

                      Section 13.1   RADIATION RECORDS  FILES

      A well-managed radiation protection program requires a  substantial  number
 of records.  Many of these records have been described  in previous  chapters.
 In this section, a summary of the  required records  will  be  provided.   For the
 purpose of this manual, the records have been organized into the  following
 series of files:
  1.  personnel fi'le
  2.  radiation protection program  file
  3.  project file
  4.  radiation work area file
  5.  instrumentation and dosimeters file
  6.  radioactive-material inventory file
  7.  waste management file
  8.  transportation file
  9.  accidents/incidents file
 10.  training file
 11.  quality assurance file.
 A system for cross-referencing these files is provided in the next section.  A
 records filing system for radiation safety files is also given in AR 340-18-6.
 Each RPO should evaluate records requirements to determine  what kind of filing
 system is most appropriate.
      Reference will be made throughout this chapter to "suspense" files.  These
 are files used for procedures that are repeated regularly (e.g.,  weekly,
 monthly, quarterly, or yearly).  Data sheets for the particular procedure are
 filed under the week, month, quarter, or year in which the  procedure must be
 performed next.  Suspense files can take the form of card catalogs, spiral note-
 oooks, or file folders, end are appropriate for scheduling  routine procedures
 such as leak tests of sealed sources, contamination surveys of radiation areas,
 instrument calibration tests, training and retraining sessions, and bioessays.
                                      13.6

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                                                              DARCOM-P 385-1
13.1.1  Personnel File
     Complete and up-to-date personnel  files provide a means of 1) assessing a
radiation worker's'training needs for specific projects or job changes,  and
2) tracking the history of the individual's exposures and of any doses received.
Occupational exposure records must be kept as part of each individual's  health
record or civilian employee medical  file.   Each personnel file must include a
signed Privacy Act statement (AR 40-14).
     A.  Identification of the Individual.  An individual's social security
number should be used for identification  on all records.  If another number is
used to identify the individual, this number should be cross-referenced  to the
social security number.  If an individual  who may work with radiation does not
have a social security number, he or she  should be instructed to get one.  The
birth date end sex of the individual should also appear on all personnel  records
es another means of identification.   In this chapter, "identification of the
individual" will mean the person's name,  social security number, birth date,
and sex.
     B.  Training Records.  Participation by a radiation worker in formal and
on-the-job training sessions should be documented to indicate the individual's
qualification to perform radiation-related tasks.  The training records-should
include:
 1.  identification of the individual
 2.  title and date of the training program
 3.  identification of the instructor and training location
 4.  e performance reting for each segment of training or each training  program
     satisfactorily completed:  a numerical or letter grade and/or e written
     evaluation.
A suggested format for training records is shown in Appendix  A.
     C.  Project/Task Listing.  To facilitate tracing an  individual's exposure
histo"} et a given insiallction, a listing cf ell the projects or tasks  on
which the individual  has worked should be included in his or  her  personnel  file.
A useful concept is  the assignment of a key word descriptor to each project.
Key word descriptors  are one- or two-wore descriptions of the focus of a

                                     13.7

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DAKCOK-P 3i5-.
project, for example, weapons testing, gaseous effluents,  radioimmunoassay.
Thev can be used to locate all other projects of the same  type for either
cr.sit'.- or intsrsite comparisons.  A record sheet for listing projects and
tasks '.'ctlc include —
    1.  identification of the individual
    2.  title and number of the project or task
    2.  key word descriptor for the project or task
    4.  the dates on which the individual began and ended participation  in
the project or task
    5.  standing operating procedures (SOPs) for the project or task.
A sample record sheet is shown in appendix A.
        D.  External-Exposure Records.  Two Department of Defense  (DD)  forms
are used to record an individual's occupational radiation exposure history—
DD Form 1952  (Dosimeter Application and Record of Occupational Radiation
Exposure) and DD Form 1141 (Record of Occupational Exposure to Ionizing
Radiation).   (Both forms are reproduced in app B.  See AR 40-14 for  details
on the  information summarized here.)  DD Form  1952 identifies the
individual's  employment status, gives dosimetry information for the
individual's  current job (e.g., the type of exposure involved and  the  dosi-
meters  and bioassays required in connection with the work), and lists  the
names and addresses of previous employers for  whom  the individual  worked
with  radiation, with the dates of employment.  A new DD Form  1952  is
initiated each time the individual is reassigned, and  the previous exposure
history is transferred to the new form.
    DD  Form 1141 includes the individual's  identity, a summary of exposures
from previous jobs, and a month-by-month record of  the individual's  dose
from  the current assignment and accumulated  lifetime dose  from exposures to
the whole body or skin of the whole body.  The installation or  location at
which each exposure occurred  is also  noted on  the  form.     A  separate  DD
Form  1141
   The inclusion of  the  title  and number  of  the  project  on which the
   individual received each monthly  exposure  would  facilitate  cross-
   referencing of  this information with  that  in  other files.
                                    13.8

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                                                             DARCOM-P  385-1
is kept to record exposures  to parts  of  the  body other than the whole body  or
skin of the whole body (e.g.,  the thyroid, head and  neck, or  fingers).   An
alternative to the use of DD Form 1141  is  the  use  of the automated  dosimetry
records prepared by the Army's Central  Dosimetry Record Repository.  Whichever
record is used, it should include exposures  received by the individual  from
outside (non-Army) work and  from medical  sources.
     Department of Defense Form 1952  is  kept in the  individual's  health record
or (for civilian'employees)  medical  file.  Department of Defense  Form  1141  (or
the automated dosimetry records) can  be  kept either  in the  individual's person-
nel file or in his or her health record  or medical file.   If  DD  Form 1141  is
kept in the personnel file,  a  chargeout  record noting the  location  of  the  form
must be placed in the health record  or  medical file.
     E.  Internal-Exposure Records.   Internal-exposure records can  include
bioassay data, the interpretation of  bioassay data,  whole-body-counter
records, and airborne-radioactivity  measurements.  All internal-exposure
records can be maintained either in  the  individual's personnel file or in  the
health record or medical  file.  If they  are  kept  in  the personnel file, a
chargeout record noting their location  must  be placed in the  health record or
medical file.
     (1)  Records of Bioassay Data.   An  individual's internal radiation expo-
sure is determined from bioassay studies.  Records of these studies should
include the following information (American  National Standards  Institute (ANSI)
Standard N13.6-1972):
 1.  identification of the individual
 2.  purpose of the sample and, if applicable, date  of  suspected intake, work
     area, and project number and title
 3.  collection period for the sample and date submitted
 4.  type of sample and size of aliquot
 5.  type of radioactivity (e.g., alpha, beta)
 6.  gross and net activity observed  and counting  time
 7.  identity of radionuclide, when  required
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DARCQM-P 385-1

  8.  cross-reference to calibration and control  data  and  confidence limits  (in
      the instrumentation and dosimeters file)
  9.  cross-reference to identity and efficiency  of analysis  equipment and
      radiochemical analysis procedure (in the  instrumentation  and dosimeters
      file)
 10.  identification of the laboratory technician(s) performing the analysis.
      (2)  Records' of Bioassay Interpretation.   In addition to  items 1 through
 10 ebove, records relating to interpretation of  the data  from  a bioassay study
 should be kept and should include:
  1.  a listing of the bioassay data used in the  interpretation, and the iden-
      tity of the radionuclide
  2.  reference to the method of interpretation
  3.  assumptions used in arriving at the conclusion,  including the known or
      assumed date of exposure
  4.  conclusion as to the magnitude and location of the body burden, expressed
      in microcuries of the specific radionuclide
  5.  identification of the individual making the conclusion.
      (3)  Whole-Body-Counter Data.  Whole-body-counter data provide an assess-
 ment of internally deposited radionuclides.  Records of an individual's whole-
 body count should include:
  1.  identification of the individual
  2.  date, time, and purpose of the count and, if applicable,  date and time of
      suspected intake
  3.  quantitative dara (e.g., length and type of count, counts per channel, keV
      per channel, energy range over which counts were made)
  4.  cross-reference to procedure, calibration factors, periodic background and
      resolution checks, and confidence levels (in the instrumentation and
      dosimeters file)
  5.  description of or reference to calculational procedure
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                                                              DAKCOM-P  385-i
 6.  identity and location of the radionuclide and magnitude of the  body
     burden
 7.  identification of the individual  making the conclusion.
A sample record sheet is shown in Appendix A.
     (4)  Airborne-Radioactivity Measurements.  If airborne-radioactivity meas-
urements and exposure times indicate that an individual  has received an internal
exposure via inhalation, the following information should be recorded:
 1.  identification of the individual
 2.  period(s) covered by the measurements
 3.  basis for exposure estimate
 4.  concentration of airborne radioactive material, length of exposure, and
     estimated breathing rate
 5.  reference to any documentation of the factors in item 4
 6.  estimated internal exposure
 7.  identification of the investigator.
     F.  Radiation Exposure Received During Prior Employment.   To ensure that
the information on DD Forms 1952 and 1141 is complete, the RPO should have each
new staff member complete and sign a questionnaire indicating whether any pre-
vious employment (civilian or military) may have involved internal or external
exposure to radiation, with the names  and addresses of former employers where
any exposure may have occurred.  Previous employers who are contacted for
information should be requested to use the individual's social security number
when providing information, to ensure  the correct identity of the individual.
     The following information on each previous exposure should be obtained and
kept in the personnel file:
 1.   the period(s) of employment and the identification of the employer
 2.   the nature and magnituc-j of the exposure, bcth internal and external, and
     the period of exposure.
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DARCOX-P 385-i
     G.  Exposure Received by Individuals at Other Installations  During 'Current
Employment.  The radiation exposure received by an individual  at  another
installation during an official  visit or special assignment  should be main-
tained  in the personnel file.  A special film dosimeter may  need  to be assigned
for the visit.
     H.  Simultaneous Employment at Another Facility.   Individuals should
report when radiation exposure is being incurred at two facilities as a result
cf simultaneous employment by two firms or government  agencies.
     I.  Exposure Evaluation.  The RPO should review and evaluate DD Form 1141
(or the Automated Dosimetry Records) and the results of any  bioassays on  a
quarterly basis and note the date of the review on DD Form 1141.   If action is
necessary to limit an individual's exposure, the RPO must notify the individ-
ual, the individual's commander and supervisor, and the responsible medical
officer.
     J.  Unusual Exposures.  Any accident/incident that involves a radiation
worker  (such as an exposure in excess of permissible limits, the use of special
exposure limits, or an exposure that results in the withdrawal of the  individ-
ual from a work position—see Chapter 11) must  be described and recorded.  The
extent  of the information recorded will depend  upon the type of accident/
incident but should include:
  1.  identification of the individual
  2.  time, date, and location of the accident/incident
  3.  description of the accident/incident
  4.  results of the event (e.g., the exposure  received by the individual
     involved, the extent and nature of skin contamination, and any  confisca-
     tion of personal property)
  5.  probable cause of•the accident/incident
  6.  action taken at the time of the event
  7.  reference to or summaries of subsequent action taken to  prevent recurrence
     of the accident/incident
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                                                             fARCOM-P  385-1
 8.  reference to or summaries  of supporting  data used to determine  the above
     items, such es radiation surveys,  film dosimeter studies, air sample
     assays, and photographs
 9.  identification of the investigator(s).
A sample form is shown in Appendix A.
     K.   Transfer of Records.  When a  radiation  worker transfers  to  another
assignment or organization, all  chargeout  records for DD  Form 1141  (or  the
automated dcsimetry records) and for biotssay records must  be removed from the
individual's health record or medical  file  and replaced with the  original forms
and records.  The health record or medical  file, containing complete and  accu-
rate originals of DD Form 1952,  DD Form 1141  or  the  automated dosimetry
records,'and bioassay records,  is sent  to  the gaining organization  to which  the
individual has been assigned.  A copy  of each document should be  retained at
the original installation, with the address of the gaining  organization noted
on the copy of DD Form 1141 to  ensure  that  any additional dosimetry  information
received after the transfer is  forwarded to the  gaining organization.
13.1.2  Radiation Protection Program File
     A record of the installation's radiation protection  policy  and procedures
should be maintained to allow the RPO  and  his or her supervisor  to  continually
evaluate and update the program.  In addition, records  should  be readily  avail-
able to demonstrate to auditors and inspectors the adequacy of  the  program.
     A.   Licenses and Authorizations.   All  documents related  to  licenses  and
authorizations to procure and use radioactive materials  should  be maintained.
These documents may include DA permits  and  authorizations;  NRC  license  applica-
tions, licenses, and amendments; and authorizations  to  store,  transfer, ship,
or dispose of radioactive materials.
     B.   Radiation Protection Policies  and  Standards.   Policies  and standards
established for the overall conduct of radiation work at  the installation
should be documented.  These records should include:
 i.  scope and organization of the radiation  protection  program
 2.  training and experience of the individuals on the  radiation protection
     f J- *\ — £
     S to i i
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DARCOM-F '£'•--
 3.  orientation and training requirements for individuals  who will  perform
     radiation work
 4.  specifications for the frequency and techniques to be  used in measuring
     the radiation exposure received by individuals
 5.  control procedures for radiation work, such as permissible levels of
     radiation and contamination in work areas, as well as  posting and labeling
     requirements
 6.  plans and procedures for radiation emergencies, including the type and
     frequency of training drills
 7.  criteria for the investigation of unusual radiation occurrences
 8.  reporting and records requirements
 9.  regulations, standards, procedures, and higher-headquarters instructions,
     along with effective dates for each.
     C.  Documents of the lonizino Radiation Control Committee.  Documents
relating to the meetings and decisions of the Ionizing Radiation Control Com-
mittee  (IRCC) should be kept.  This information should include reports on  IRCC
reviews of applications for approval to use sources of ionizing radiation.  The
records should note whether each application was approved or disapproved,  the
conditions under which each source was approved for use, and the qualifications
of the users.
     D.  Procedures for Obtaining and Evaluating Data on Individual Exposures.
The procedures used to obtain, process, and evaluate data for  individuals'
external and internal exposure records should be recorded.   Records of the
methods used to obtain an individual's exposure should refer to pertinent  pub-
lished documents or reports and should show the period of applicability of the
methods used.
     E.  Inspections and Appraisals.  Documents related to compliance inspec-
tions performed by DA and civilian licensing agencies  should be maintained.
These records should include notifications of inspection, inspection  report?,
and documents related to follow-up corrective actions.
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                                                              DARCOM-P  3E5-1
     A neelth physics appraisal  provides  an evaluation of the  overall adequacy
and effectiveness of the radiation  protection  program.  Appraisals may  be
performed by a team of outsiae experts  and/or  the  installation RPO (see  Chap-
ter 15).   All of the documents related  to the  appraisal of the radiation
protection program should be maintained and should  include appraise! notifica-
tions, findings, and corrective actions.
     F.  Changes in Procedures and  Methods.  Substantial revisions of proce-
dures, methocs of evaluation, or policies should be  recorded.  When  pertinent,
the reasons for such changes should also  be recorded.
13.1.3  Project File
     Each project or task should be fully documented.  A title and an identifi-
cation number should be assigned to a  project  before it is begun, and project
records should be ^iled by the project  identification  number.
     A.  General Records.  All documents  relating  to a project should include
the project's title and identification  number, key  word descriptor(s) relevant
to the project, and the name of the principal  investigator.   A list  of  key word
descriptors available for assignment to a program  should also be  kept in the
project files.  The records for each project should include:
 1.  a complete description of the  project with its start and completion dates
 2.  a complete listing of all radioactive materials used for the project,
     including for each source its  activity, the date the activity was  deter-
     mined, and its half-life
 3.  a complete listing of all instrumentation used in the  project,  includvK
     for each instrument its identification number (serial  or inventory num-
     ber), company, model number, and  storage  location
 4.  the principal  investigator, and a  list identifying  all  project  workers end
     the dates on which each individual began  and  ended  work on  the  project.
Sample project forms are shown in Appendix A.
     B.  Standing Operating Procedures.  Specific  procedures performed  in con-
nection with a project are described in SOPs.   The SOP is  a  locally developed
form completed by the area supervisor  and countersigned  by  the RPO prior to
the start of work.   The SOP should  include:

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DARCCtt-P 385-1
  1.   the title and number of the project
  2.   effective date of the procedure
  3.   identity of personnel and/or the organization authorized to perform the
      work
  4.   location of the work
  5.   potential radiation hazards and specific procedures, instructions, and
      precautions ,to be observed
  6.   equipment and dosimetry requirements
  7.   protective-clothing and equipment  requirements
  8.   descriptions of conditions that would terminate or suspend work in
      progress
  9.   identity of the individual approving the procedure.
 A copy  of  each SOP initiated for a project should be included in the records
 for  the project and kept in the project file.
 13.1.4  Radiation Work Area File
      Documentation of work area conditions is necessary to ensure that good
 housekeeping procedures are followed and that,  in the event of an accident/
 incident,  the radiation source could be quickly characterized and doses to
 personnel  in the area estimated with reasonable accuracy.
      A.  General Records.  Any investigation of a radiation accident/incident
 requires that substantial supportive data be available.  The radiation work
 area file  should therefore include for  each laboratory or work area:
  1.   its location and a map showing the layout  of the area
  2.   a  description of the uses of the laboratory or area and its facilities
      (e.g., hoods, glove boxes, permanently installed equipment)
  3.   the titles and numbers of projects carried out in the area, with  the  iden-
      tity  of the principal investigator for each.
      B.  Radiation and Contamination Surveys.   Surveys are conducted to assess
 the  condition of a particular work area.  Survey records should  include:
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                                                              DAKCOM-P  3£5-i
 1.  dete and time of the survey
 2.  location of the survey, that  is,  building  and  room  (sketches may be
     included)
 3.  specific location or object surveyed  (sketches  may  be  included)
 4.  purpose of the survey (e.g.,  leak test  of  sealed  source,  routine survey
     for contamination on floors and  other surfaces, or  survey to establish
     dose rates to personnel)
 5.  identification (type and serial  number)  of the  particular radiation  detec-
     tion instruments used to perform the  survey
 6.  measurement results (e.g.,  dose  rates and  contamination  levels), and
     housekeeping conditions observed
 7.  conclusions and recommendations
 8.  identification of the individual  performing the survey.
     C.  Area Monitoring Records.   Chart  recordings  of radiation area monitors
should identify:
 1.  period covered by the chart (beginning  and ending dates  and times)
 2.  location of the detector and  the area monitored
 3.  a clear relationship between  chart divisions and  the exposure  or exposure
     rate units
 4.  identity of the scale or range of operation
 5.  notations of source checks  and calibrations performed
 6.  identification of the individuals operating the equipment.
     Additional information for  continuous air  monitors should include:
 1.  type of instrument (e.g., fixed  filter  or  moving  tape)
 2.  tape and chart speed
 3.  specific relationship between the chart divisions and the concer.trction cf
     the airborne radioactive material, which depends  on the tape speed and
     flow rate of a moving filter  unit, or on the flow rate of a fixed filter
     unit.
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DARCOM-P 3BI-*
     D.  Airborne-Radioactivity Monitoring  Records.   If airborne  radioactive
material is monitored, the following information  should be  recorded:
 1.  date and time of sampling
 2.  general location of the air-sampling station (building and  room)
 3.  specific location at which the air sample  was  collected
 4.  purpose of sample collected (e.g., routine air sampling or  air  sample  for
     special evaluation)
 5.  type of sample collection equipment used (e.g, filter, impact,  or evacuated
     ionization chamber)
 6.  collection efficiency of sampling system
 7.  flow rate, duration of sampling, and total volume of air sampled
 8.  identification of sample analysis equipment used
 9.  counting data:  time count was taken,  background, source count, gross
     count, net count, duration of count
10.  reference to calculated correction factors such as backscatter, self-
     absorption, and efficiency of analytical equipment
11.  calculated concentration of airborne radioactive material
12.  identity of the air contaminant, if determined
13.  identification of the individual performing the analysis.
13.1.5  Instrumentation and Dosimeters File
     If the limitations of an instrument have not been determined and the
instrument has not been calibrated, the information that it provides about
radiation levels in work areas is useless.   Therefore, records documenting the
availability, limitations, and calibration of all radiation-measuring instru-
ments and dosimeters should be kept in an instrumentation and dosimeters
file.
     A.  Capabilities of Dosirr.eters and Instruments.  The following information
on the capability of equipment should be recorded:
                                     13.18

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                                                              DARCOM-P  385-1
 1.  identification, description,  and functional  specifications  of the  individ-
     ually worn dosimeters and the other radiation  measurement instruments
     used in the radiation protection program
 2.  dcte and results of any acceptance or performance  tests  that show  the  sen-
     sitivity, ranee, and energy dependence of the  instruments
 3.  specie  studies documenting bases for use, efficiency, correction  factors,
     and interpretation of date.
     B.  Calibration and Maintenance.  Procedures,  criteria,  and schedules  for
calibration and maintenance of radiation measurement  instruments and dosimeters
are of value in demonstrating the instruments' dependability  and reliability.
Routine survey instruments should be calibrated every 90 days unless subject  to
extreme environmental conditions,  hard usage, or corrosive environments.   In
these cases, more frequent calibration is required. (ANSI N323-1978).  Contin-
gency instruments should be calibrated every 240 days.   A suspense file can be
used for this purpose.  The records system should include:
 1.  procedures used for the calibration of the individually  worn dosimeters
     and other radiation measurement instruments
 2.  descriptions of the calibration sources and any data showing intercompari-
     sons with sources from other laboratories
 3.  data on the frequency of calibrations
 4.  date and results of the calibration tests, including the identification of
     the individual performing the test
 5.  maintenance history of individual radiation measurement  instruments.
     C.  Inventory Records.  In addition, the following information should be
documented for each radiation-measuring instrument and dosimeter:
 1.  identification:  type, company, inventory number
 2.  manufacturer's specifications
 3.  titles and numbers of projects for which the instrument has been used
 4.  person to whom the dosimeter is assigned, and documents  used to record
     issuance and retrieval of dosimeters.
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DARCOK-? 385-1
13;-l-.6  Radioactive-Materiel Inventory File
     The identity, form, activity, and location of each  radioactive  source must
be documented to ensure good housekeeping  procedures  and  provide  a  quick  indi-
cation of a lost source.  Information on the  form and  activity  of a  source can
also be used to indicate radiation doses to personnel  in  the  area (in  adcition
to personnel-monit&ring devices), particularly for cases  where  radioactive
material was inhaled or ingested.
     A.  Sealed and Unsealed Sources and Radioactive  Commodities.  As  soon as a
radioactive source or commodity is received,  a file containing  items 1 through
4 below should be set up.   Subsequent information that should be  kept  in  this
file includes items 5 through 8 below.  Items 9 through  11  should be included
for radioactive commodities:
 1.   name of shipper, and DA authorization and NRC -license  of shipper
 2.   packing papers that identify the source, the amount and  activity  of  the
     source, and the date received
 3.   designated storage location (a subsequent change in storage  location, or
     transfer or disposal  of the source, should also  be indicated, with the
     date of the change)
 4.   department the source is assigned to, and the  responsible individja'!
 5.   locations and dates of use, identity  of  involved personnel and (for
     unsealed sources) quantity used and quantity  remaining
 6.   titles and numbers of projects in which  the source or commodity was  used
 7.   leak test records:  date, identity of person  performing  the  test, tech-
     nique used, counting instrument used  (with  its  inventory or serial number),
     and test results (in dpm, which may be converted to the  appropriate  curie
     unit)
 8.   disposal details - how, when, and where  the sources or commodities were
     disposed of
 9.   research, development, and test summary
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                                                             DARCOM-P 3E5-1
10.   associated technical  bulletins
11.   system safety sources.
     A suspense file may be  established  to  schedule  leak  testing  of  seeled
sources.   A sealed-source inventory  should  list  all  sealed  sources available  at
the  facility, their activity as  specified on  the packing  papers,  with  the date
of receipt, and their storage location.  This  list should be  kept in th-
inventory file and updated whenever  these conditions  change,  for  example, when
the  storage location is changed,  or  the  source is transferred to  another
department or disposed of.
     Because unsealed sources present both  external-contamination hazards and
the  possibility of internal  exposure through  inhalation,  ingestion,  or entry
through a wound, it is essential  to  know how  much material  is available at  any
time in a particular location.   An unseeled-source  inventory  should  therefore
include a list of ell unsealed  sources available at  the site, the quantity  and
activity of each on the date of  its  receipt,  the storage  location of each,  and
the  quantity and activity remaining  on the  date  of  any change in  a source's
location.  The total depletion  of an unsealed source should be indicated  on the
inventory.
     B.  Environmente 1 Samples.   Environmental samples (e.g., air, water,
soil, vegetation, and game)  are  often used  to characterize  the impact of  a
particular operation on the  environment. The samples themselves  should be
labeled (a numbering system is  frequently used)  and  the records of these
samples should include:
 1.   label identification number
 2.   type of sample (water,  vegetation,  etc.)
 3.   where the sample was obtained
 4.   counting results
 5.   instrument used for counting
 6.   any actions taken as a  result of a  high  reading
 7.   disposal details - how, when, and where  the sample was disposed of.
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DARCOM-P 385-1

13.1.7  Waste Management F'Tle
     Radioactive waste may include sealed or unsealed  radioactive sources;  con-
taminated equipment, clothing, and supply items;  and biological  organs.   Chap-
ter 10 provides guidance for the handling, storage,  end disposal  of low-level
radioactive waste.  In general, the following items  should be  documented for
radiaoctive waste:
  1.  assigned identification number(s)
                 i
  2.  physical description of the waste:  solid, liquid, or gas,  quantity,
     shipping weight and volume, number of containers, shipping  permit number,
     transport group, package specification and labels used
  3.  chemical and radioisotope description:  hazardous chemicals, solvent
     present  (liquid), radioisotopes present
  4.  radioactivity and radiation measurements:  activity, maximum dose rates at
     surface  and 1 meter, classification
  5.  identification of previous responsible department or individual and stor-
     age location
  6.  disposal details - how, when, and where the material will be disposed of
  7.  identification of responsible individual(s).
13.1.8  Transportation File
     Any movement of radioactive material onsite or offsite requires careful
planning by the shipper and the receiver.  Specific documents must accompany
the material, and records of all movements must be kept.  Shipping procedures,
records, and  packaging requirements are discussed in Chapter 9.   The shipping
documents and records described there  include:
  1.  consignee license
  2.  bill of  lading
  3.  description of material on shipping papers
  4.  shipper's certification
  5.  specific instructions for exclusive-use shipments
  6.  survey records
  7.  records  showing compliance with package design-and-performance standards.
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                                                             DARCO>!-P 385-1
13.1.9  Accidents/Incidents File
     Complete records of radiation accidents/incidents  are  necessary  for  efter-
the-fact documentation of the event.   The  following  information  about  each  acci-
dent/incident should be recorded:
 1.  date, time, and location
 2.  description
 3.  results of the event (e.g., the  exposure  received  by  the  individual(s)
     involved, the extent and nature  of skin contamination,  and  any confisca-
     tion of personal property)
 4.  probable cause
 5.  action taken at the time of the  event
 6.  reference to or summaries of subsequent action  taken  to prevent
     recurrence
 7.  reference to or summaries of supporting data used  to  determine the  above
     items, such as radiation surveys, film dosimeter studies,  air sample
     analyses, and photographs
 8.  identification of the investigator(s).
13.1.10  Training File
     The RPO or the training supervisor should maintain a  file that  includes
the following information for each course  thet is given:
 1.  date and location of course
 2,  identity of instructor(s)
 3   description of course content, including  course outline, syllabus,  and
     other descriptive information
 4   identification of individuals in attendance (name, social security number,
     birth date, sex)
 5.  result! of examinations.
A suspense file can be set up to schedule  training or retraining sessions.
                                     13.23

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DARCOM-P 385--

13.1.11  Quality Assurance File
     Quality assurance programs are described  in  Chapter  14.   Recoras  cf  each
program element should be maintained,  including documents  related  to:
 1.  facility design
 2.  procurement
 3.  organization of the program
 4.  control of purchased material, equipment,  services,  and  special  processes
 5.  inspections 'and tests
 6.  control of measurement and test equipment
 7.  handling, storage, and shipping procedures for material  and equipment
 8.  nonconformance and corrective actions.
                      Section 13.2  RECORDS FILING SYSTEM

     A recordkeeping system that incorporates the capacity for extensive cross-
referencing among files can be invaluable in answering questions and solving
problems related to an individual's radiation dose.   The II files in which the
records just described should be kept—personnel  file, radiation protection
program file, project file, radiation work area file, instrumentation and
dosimeters file, radioactive-material inventory file, waste management file,
transportation file, accidents/incidents file, training file,  and quality
assurance file—contain some overlapping information that would permit an
individual's work and exposure history to be traced and the conditions under
which the individual received any dose to be reconstructed quickly and
accurately.
     Through a cross-reference system such as that shown in Appendix C, persons
who were involved in a project, whether as principal investigator, calibrator
of instruments, or radiation surveyor, can be identified and could be called on
to assist in the evaluation of exposure trends or the investigation of occur-
rences.  The two flow charts in Appendix C illustrate how this system could be
used to solve specific problems.  The repetition of some data in more than one
                                     13.24

-------
                                                             DARCOM-P 385-1

file permits the investigator to  track  down  information by moving from one file
to several others,  es necessary.
                  Section 13.3  RECORDS  RETENTION AND STORAGE

13.3.1  Types of Records Retention
     Records can'be kept es hard copy (paper),  on a  computer disc  or  tape, or
or microfilm or microfiche.     The  main considerations  in  choosing which
method of retention to use ere:
 1.  the storage space needed for the number  of records  generated
 2.  the ease of accessibility to the stored  information that  each type  of
     record provides
 3.  the admissibility of each type  of record as evidence in a court  of  law.
The initial expenses of establishing each type of system should also  be  con-
sidered in relation to the long-term gains of the system, but  an extensive
cost-effectiveness study is beyond the scope  of this manual.   Each form  of
record is discussed below in relation to storage needs,  accessibility, and
legal  status.
     A-  jjardCopy.  The American National Standards Institute recommends  in
its publication ANSI N13.6-1972  that dose records for every individual occupa-
tionally exposed to radiation be kept until  10 years after the individual's
death  (if the date of death is known) or until the  individual  would have
rearh^d the age of 75 (if the date of death is not  known).   Records should  be
kept this long for both scientific purposes (to permit studies of the long-term
effects of radiation) and legal  reasons.  An  extensive records system for a
large  program, if kept in hard-copy  form, could involve considerable paper and
space.  Easy access to such a system would require  an excellent centralized
(0  Kicrofi ',rr. is a fine-grained, high-resolution photographic film corteininc
     en image greatly reduced in size from the original.  Microfiche "is a
     sheet of microfilm containing multiple microimages in a grid pattern.
     The term microform is used to refer to any storage form that uses
     microimages.

                                     13.25

-------
DARCOM-P 385-1
filing system.  For a small  installation with  relatively few  recorcs,  hard-copy
files would be practicable.   Moreover,  in terms  of  legal applications,  hard
copy is often the preferred  method of records  presentation; in  a  court  of  lew,
evidence on original  hard copy is difficult  to dispute.
     B.  Computer Records.  If computer storage  is  used, space  must  be  allotted
for the computer itself, for a terminal, and for storage of the discs  or
tapes.
     There is a 'great deal of controversy over the  admissibility  of  a  computer
printout as evidence  in a court of law.  It  is difficult to guarantee  that a
program or number has not been tampered with,  and the  data  records  cannot  be
signed as a way of verifying a record or a change in a  record.  To  stand up as
legal evidence, computer entries would have  to be verified  upon entry,  and
access to the computer would have to be strictly controlled.
     C.  Microform.  Microfilm and microfiche  do not take up  much space, and  a
good filing system would allow easy access to  records  in these  forms.   Micro-
film has the legal status of an original document if it has been  made  in com-
pliance with the law.  '
     D.  Combinations.   The  use of a combination of record  retention systems
would provide flexibility and make use of the  advantages of each  system.   A
computer system could be used to provide day-to-day access  to all types of
records, and hard copy or microform could be kept for  legal evidence.
     The filing system described in this chapter assumes the  use  of hard  copy;
however, the concepts discussed could be easily  incorporated  into a  computer  or
microform file.
13.3.2  Retention Period
     The minimum retention period for all the  records  described in this chapter
is 5 years (ANSI 13.6-1972;  AR 385-11).  However, because  records relating to
personnel exposure have both scientific and  legal implications, the following
records for each individual  should be kept until the  individual would have
(a)  See the following U.S. Code sections:   44 U.S.C.  3312, 44 U.S.C. 2112,
                                     13.26

-------
                                                             DARCOhi-P  3E5-1
reached the age of 75  (if the cete  of deeth  is not known) or until  10 years
efrer his known deeth  (ANSI  M3.6-1972):
 1.  records of interne!  and external exposures
 2.  calibration date  associated with evaluation of the  individual's exposure
 3.  records of procedures and metnods used  to interpret and evaluate the  indi-
     vidual 's exposure
 4.  records describing unusual  occurrences  in which the individual was
     involved.
13.3.3  Storage Precautions
     The effort involved in  keeping good  records would  be wasted  if they were
lost because of fire or theft.  To  prevent such  a  loss,  the  following  sugges-
tions are presented:
 1.  Keep duplicate copies of all vital records  in an area  remote from  the
     original documents.
 2.  Use a standard records  vault to minimize  the  possibility  of  a fire start-
     ing in the vault or entering it from outside  (National  Fire  Protection
     Association 1980).
 3.  Consider microfilm for records storage after  consulting applicable state
     laws concerning the legal admissibility of  microfilm.


                                  REFERENCES

American National Standards  Institute  (ANSI).   1972.   Practice for Occupational
  Radiation Exposure Records Systems.  ANSI N13.6-1966  (R 1972),  New York.
American National Standards  Institute  (ANSI).   1978.   Radiation Protection
  Instrumentation Test and Calibration.  ANSI  N323,  New York.
National Fire Protection Association.  1980.  Standard  for the Protection of
  Records.  Publication 232, Boston.
U.S. Code.  Title 44,  Section 2112,  "Legal Status  of Reproductions; Official
  Seal."
                                     13.27

-------
DARCOM-P 385-1

U.S. Code.  Title 44, Section '3312, "Photographs or Micro Photographs Con-
  sidered as Originals:  Certified Reproductions Admissible as Evidence."

U.S. Code of Federal Regulations.  1982.  Title 10, Part 19, "Notices, Instruc-
  tion and Reports to Workers; inspections."  U.S.  Government Printing Office,
  Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 10, Part 20, "Standards for
  Protection Against Radiation."  U.S. Government Printing Office, Washington,
  D.C.

U.S. Code of Federal Regulations.  1982.  Title 29, Part 570, "Child Labor
  Regulations, Orders, and Statements of Interpretation."  U.S.  Government
  Printing Office, Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 29, Part 1910, "Occupational
  Safety and Health Standards."  U.S. Government Printing Office, Washington,
  D.C.

U.S Department of the Army, Headquarters.  "Maintenance and Disposition of
  General Personnel Management and Safety Functional Files."  In Army
  Functional File System.  AR 340-18-6, Washington, D.C.

U.S. Department of the Army, Headquarters.  Safety - Ionizing Radiation
  Protection (Licensing, Control, Transportation, Disposal, and Radiation
  Safety).AR 385-11, Washington, D.C.

U.S. Department of the Army and Defense Logistics Agency.  Medical Services -
  Control and Recording Procedures for Exposure to Ionizing Radiation and
  Radioactive Materials.  AR 40-14, DLAR 1000.28, Washington, D.C.
                                     13.28

-------
                                    DARCOM-P  385-1
         APPENDIX A

    SAMPLE RECORDS FORMS

Personnel File:
  Project Sheet
  Training Record
  Whole-Body-Counter Record
  Radiation Occurrence Record

Project File:
  Project Characterization
  Project Personnel List
  Key Word Descriptors
          13.29-A

-------
DARCOM-P 385-1
Name
    Project
      No.
 1.

 2.

 3.

 4.
Name
                       PROJECT SHEET FOR PERSONNEL FILE
 1.

 2.

 3.

 4.
Project
 Title
PROJECT SHEET

SS*	

  Key Word
  Descriptor
                             Birth date
Start
            TRAINING RECORD

             SS#             Birth date
               Sex
         Date
         Course Title
ind
                                      Sex
               Date     Instructor/Location     Test Score
                                   13.30-A

-------
                                                              DARCON-P 365-1
                 WHOLE-BODY-COUKTER RECORD FOR PERSONNEL FILE
Name
WHOLE-BODY-COUNTER RESULTS

       SSf             Birth date
Sex
Date of Measurement
Purpose
	 routine
	 suspected intake: see below*
                 Type of Measurement (checl'):
                 	whole bod}'
                 	 lung
                 	 thyroid
Radionucl
(check)
ide
Count Rate Activity Body Burden
(cpm) (dpm or uCi) (dpm or uCi)
2/ilAm
226Ra
222Rn
235
"3u
234Tu <^
< T*A
ifOV*^
Instrument used (name, c
            number, identification number)
Calculetional Method:  cpm to dpm or uCi
* Date of suspected intake:
  Location of suspected intake:
  Project number, title, principal investigator:
                                  13.31-A

-------
DARCO.M-P 385-1
Name
                RADIATION OCCURRENCE RECORD FOR PERSONNEL  FILE


                          RADIATION OCCURRENCE REPORT

                                 SS?             Birth date
                                                                       Sex
Occurrence Time end Date

Occurrence Reported by
                                                 Building end Location
Air Sample  ID Number
                            Dosimeter ID Number
Other supporting data (description and location)
                                                       Survey ID Number
Occurrence Description:
     Probable Cause:
     Initial Actions:
     Subsequent Actions to Prevent Recurrence:
Radiation Exposure Data
(check) _ a _ 8
                                                           Dose
                                                                           (rem)
(check) _ _  Skin contamination
_ Hospital ization
Describe:
                                       In-vernal deposition
                                                                  First  aid
                                        Investigated by
                                                       Date
                                  13.32-A

-------
                                                              DARCOM-P 3&5-1

                   PROJECT CHARACTERIZATION' FOR PROJECT FILE
                           PROJECT CHARACTERIZATION

Project No.	    Key Word Descriptor _
Title	  Principe!  Investigator

Start Date                    Ending Date	
Location of Work:


Description:
Instrumentation (I.D. Number, company, mode!  number,•storage location)
Radioactive Materials
                    Sealed or ^-  ^MVV ~"          Activity and
Identity   ID No.   Unsee.1>< TfcalaBir  Radiation(s)      Date       Half-life
HP Support
   Dosimetry required
   Monitoring required
   Protective equipment required_

   Special  instructions	
                                  13.33-A

-------
DARCOM-P 385-1





                    PROJECT PERSONNEL LIST FOR PROJECT FILE






                           PROJECT PERSONNEL RECORD




Project No.	 Key word descriptor_
Title	Principal investigator_
                                                                   Date
  Name of Radiation Worker       Social Security Number     (Start)
                     KEY WORD DESCRIPTORS-^^.
                                     £5TRIPTOR LIST



Key Word Descriptor       \^. "	Project Number and Title
                                   13.34-A

-------
                                         D/vRCOM-P 365-1
             APPENDIX B

OCCUPATIONAL RADIATION EXPOSURE FORMS

   Department of Defense Form 1952
   Department of Defense Form
               13.35-B

-------
DARCOM-P 385-1
                 DOSIMETER ALLIGATION AND RECORD Of OCCUPATIONAL RADIATION EXPOSURE
                     f ufibiv of ryfn *U information
                                                        S*r f~-iuar** Art Siattment on rrv*r*t
     FUl-L NAME ti^Ml. /V*l. it**


     JARVIS, Whitney K.
                                                                      SOCIAL HCURITY Hi.

                                                                       777-07-300C
     DUTY XCTlOx /£•.»!_ »«m. l»n.«u.;

      Research  Laooracory
                              t. JOt TITLE

                                 Chemitt
                                          e. Ol/T> »HONI

                                             263-1SK
              f AY a
   Cl VUIAN

     CS-12
                                 i- HAVf YOU WOMN A DOSIMETER ISSUED IY
                                  THK COMMAND IN TMC 'AST
                                               ENC
                                          . DATE Of KADlATlON
                                          nrrmiDDi

                                              81-05-01
tO. DUTN ET ATUS

     [^^E RMANENT

 LJTRANSIENT i »HOLE-»ODY
                                                                                    J'INCER
     IIOASSAVS RfOUlRtD
   _'Y«
              COUNT
                DNO
                    THYROID

                     OVES
                                            URINALYIlt
                                                                       FREQUENCY
                                                       LjMflN
                                 o/vx cxrn ro* mtis n THROUGH IP
   It. DOSIMETERISI ISSUED
       81-05-02
I                              It. DD 'ORM(S) 1141 INITIATED
                                  81-05-03
                                                               17. DOSIMETERISI DISCONTINUED
   11. LAST DOSIMCTER1S) RETURNED   It. LOCATOR CARD TO HEALTH
                                   RECORD  81.05-03
                                                               20. DD FORMISI II" TO MEDICAL NCCORDS
                                       OCCUPATIONAL
        NOTZ: Thii UCUOD only «ppb« u> toe indmdiuJ wbo tat wo^k^aAnth ndution-producani otTiea or ndioiaotopo
   t£ i permanent nitiu.   Lin only tbo»* cmploycn (or wnom you workre «htb ndutioo.
      NAME Or EMPLOYER
                                   ADDRESS
                                  Faji. rffy. »mtt,
                                                           VRr  »*o
                                                                                         ^o nor
                                                                                         Ui IM>
   Nuclear Services,
   Inc
                          Shickshinny, PA
                             78
                                                              06
                                                                        80
                                                                            Oi
   Rosewacer   Univer-
   sitv
Portland, OR
                                                       80
                                                                        El
                                                           TOTAL EXPOSURE DATA
   DD  /Vo-v  1952
                                     EDITION OF 1 SEP 74 It OBSOLETE..
                                                  13.36-B

-------
                                                                                            DARCO!-!-?  3£:-l
                                          PRIVACY ACT STATEMENT
                               DATA REQUIRED BY THE PRIVACY ACT OF 1974
                                                 (6 USC 552*1

1. TJ'| i J Of FORM: Do«imeier Application and Record of Occupational Rad»lior, Expoturt.

2. PRESCRIBING DIRECTIVE:  AR 40-14 and DLAR 4)45.24.

8. AUTHORITY:  I USC SOl-DepanmenuJ Refutation.  10 USC 1071. Mtoicml and Dental Care. rMrpoa«*: 41 USC 2072.
2093, 2091. 2111, 2133, 2334, 2201fb). and 2201 1 o). The authority lor aolicmnr tne aociaj »rcunty Dumber u 10 CFR 20.
44 UEC 3101-R*cord Maaaeement by Afeney Ketdt, General Duliet

4. PRINCIPAL PUrU*OSE(S):  To Mtabliab qualification  of penonnel monitonnf and document prrnoui expoaurr butory.
Tot inlormalion u u>«d ih the evaluation of ruk of expoturr to loruzinf radiation or radioactive material*. The oau p«rrruu
n>unin?1\L! eompanaon of'both current (iinonHem) and lonf'ierra ezpo«urr  10 lonjunf ndmion ur ndioactivc xnatcnal.
Dau OB your tzpotun to ionJZjn[  ndianoo or r»dio«cn»t maicnaU u «»»iltblt to you  upon rrquai.

5. ROUTINE USES   Tbt infornutior mir be lucd u> proridc dau to other FrdenJ afencio. academic irutitutioru.  and oon-
|ovtmmentaJ afcncie*. auch at the  Sa. jna) CounciJ on RadiatiOD hrotection anc Meacuremeni and tbc NationaJ Knearcb
Council, uirolTed in moniionn(ir appropriate aulboritie* in the event the information indicate* • TiolauOD or potenuaJ violation
of Uv aAd in the courae of an adcuojatntive or judicial
6. MANDATORY OR VOLUKTARY DISCLOSURE AND EFFECT OK £NDATDUAL NOT PROVIDING INFORMATION:
It ii Tolunury that you fumiih the requeaud information, includint tocial aecunty Dumber; however, tbt irutailatioo or acuv-
\ty mint maintain a completed DD Form 1141 on each individual occupationally cxpoaed to ionixie; radiation or radioactive
material mi required by 10 CFR 20. 28 CFR 19)0.96 and AR 40-34 /DLAR 4345.24.  U information u not iurnubed. individ-
ual miy not become 1 radiation worker. Toe aocval accuhty number u uaed to aaiure that the Army /Agency bu accurate
iceatlfier not aubject to the coincidence of aimilar name* or birtbdi'.e* amon{ the Ulf e number of penoni on whom expoaure

-------
DARCOM-P 3£5-1
RECORD OF OCCUPATIONAL EXPOSURE TO IONIZING RADIATION
rof :-.'TF: :r:o-: SEF Kt\-rps! or sw£rr
07* JARV-S, WHIT!,TV 1C.
•L ACt »-t"t
WHOLE =5rv
AC T 1 V - »
i
Previous Exposure-
Adr-in Dose -
APG-EA, HD
do
do
do
do
do
dc
do
do
do
do
Tort PlunXett
do
do
do
CO
do
do
Fort Smitr., CA



1. Nuclear Services
2. Rosewater Ur.iver
No fil= badge re
KR - none reported;
Has wrist badge No.

rROM TC
1 C
Aucoe ArrcE
Aprbi Aprcr
2May69 4 June 9
6Jun69 '60un69
5jun69 4Jul69
5Jul69 i7Aua69
8Aug69 l6Sep69
6Se?69 !6Sepc9
7Se?69 4Oct69
5O=t69 4Nov69
5Nov69 i 6Dec69
Film Badoe Ser^-:
6Dec69 !6Dec69
2 Jan 70 ! 3Feb70
4Feb70 ; 3Kar70
4Mar70 2Asr70
22Ma.r70,'22Mar70
2Apr70 l4Kay70
5May7C i3Jur.70
4Jun70 !2Jui70
Auc70 !0\il71
i
I

> »CC *.. » I C w * ' » * «•**»• *• *
777-07-3000 TDK
DOSE txis "E»iCC
unoJ. I'.r It. "pruAfMT ••
C 	 -,01^
» It >1 12
::.-. X.12? I.T oc.107
- 05.000
NF. iQO.OOO NC 00.000
i'uarterlv Review by RPO
00.003 00.010 NU 00.010
NR '.00.078 KU 100.078
Ci^n* irtc-3 ' NU ! 00. 4 16
Quarterly Review oy RPO
NR 100.064 N'J 00.064
NR 00.075 NU 00.075
00.016 100.070 NU 00.070
ce Discontinued 6 Dec 69
Quarterly Review by RPO
NR IOC. 000 ' 00.000! 00.000
NR 100.178 • 00.062! 00.240
00.052 J02. 504 00. 126! 02.630
Quarterly Review by RPC
Relieved Fror Duties'
Ir.volvinc Exposure to RADS
00.017 JOO. 100 ! 00.0431 00.143
Exposure Received i
i
SAMPLE i
i i
•- • • - t c '
',...;•..' V.,'.
15 Apr 4:
ACCunui. »TEO OOSC
'O' »L
1]
• f Khdl ft-
1IH-1II
14
OC.10"
05.1C7 45.000
05.107
45.000
-
05.117
05.195 '
05.611 •
45.000
45.000
45.000
-
C5.675
05.750
C5.E20
45.000
45.000
45.000
_
_
05.S20
06.060
08.690
45.000
45.000
45.000
-
OE.690
06.690
06.S23
08. £22

50.000
50.000
50.000
55.000



INITIA^
•C "IO-
CN * » -
li
•CED
CED
CED
JER
CED
CED
CED
JER
CED
W* W
WLW
WLK
JER
RKO
RKO
RKC
KJK
RKC
R.KO
RKO
ec.



, Inc., Shicicshir.ny , PA 3. Acair. Dose «= . ~ "eir' . = 00.416 rem
xnor* ^ii s
sity, Portland, OR 4. Alleged overexposure.
:crds (AR 40-14). 5. Pending investigation IAW AR 40-5.
NU - not used
086.
TO BE RETAINED PERMANENTLY IN INDIVIDUAL'S MEDICA^ RECORD
                                   13.38-B

-------
                                                           DARCOK-P 355-1
RECORD OF OCCUPATIONAL EXPOSURE TO IONIZING RADIATION'
'Of iKirffcrii'*: ;ci KI^IPH . ' ixrr-
-^-»t« 'Ko^tr* »o» *'o- ''*"." '*'*.''
086 JARVIS, ""KITtCY N. 777-07-300C TDK. li Apr <:
»L ACt *MERC
CX»OSU«t OCCURRED
ACTIVITY
f
Previous Exposure"
AdJiir. Dose ;
APG-EA, KL
do
do
do
do
do
do
do
do
do
do
Fort Plurucett
do
cc
cc
do
CO
do
Fort Smith, CA



It REh4A*«s (Connnuf pn *tfit
I. wrist Eecora {W£
2. Nuclear Services
3. Rosewater Univer
No filr badge re
NR - none report
»e«ioo
Of CX'OHIRC
rnow TO
!l>*r-*
T • t
Auooc .AproE
Acrot 'Acr69
3May59 i-iO-j.169
6Jun69 I60\in69
SJiyi6& l4Ju!69
5Jul69 ;"JAucr69
6Au?69 l6Sep69
8Seo6S IBSeja69
7Sep69 !4Oct6S
SOc-69 i«Novc9
5Nov69 l€Decc9
File Badge Servi
6Dee69 l6Dec69
2Jan70 13Feb70
4Feb70 :3Kar70
4K5.r70 i2Asr70
22>4ar70 !a2Mar70
3Anr70 i4Mav70
5Mav70 l3Jun70

rs*nj , ».«A» j »r«toc
1 1C >' ' 12
OC.2G-;
75.000
NK OO.OOS irj 00.009
{Quarterly Re\"iev by RPO
00.007 D0.01E NU 00.016
• NR D0.15S NU 100.159
l ~lir. .
Badoe.' Lost11 NU . 106.250
1
(JvlErtp-rlv Rp\ri««v hy RPr
NR bC. 143 : NU 100.143
NR DC. 162 NU ioc.162
00.032 DO. 150 NU '00.150
ce Discontinued fc Dec i 69
Quart£trly Review bv RPO
NR DO. 015 '• NU 100.015
NR bo. 420 ' NU 100.420
00.140 h.E.1255 NU 16.125
i
Quarterly Review by K>0
Relieved From' Duties'
Involving Exposure te KAD
00.025 DC. 200 : ra |oO. 200
1 j NO r^irc. ataoe worr. or
Auc70 iJu!71 | Exposure Receivec i

i
,
.
SAMPLE j
!
»CCU-U. »TCi DOSC
P C MMI ft-
TOIAU »»^t
LircTiMt t i.t'r-'»-c
J'N-J»i
1) l«
OC.204 »«'A
7£.ao^ ::«
75. 213 NA
HA
75.231 ^"*
75.390 "^
B1.640 N^
NA
B1.7B3 NA
81.945 KA
62.0oe *"'1
NA
NA
62.110 NAV
62.530 • t;;--
100.655 ; NA
. ! NA
100.655 • !;"
IOC. 655 ! N«
100. B5 5 | K?>
10C.E55 i N?>
!
i
:
ion*/ •«••[ ii jrmc»F»»ry) . / -; rer
L«ro-^ n/Zi 4 ^f*T^i^ T>-i<:«> » *c*. ,., ->T;ri
.HIT,At
f T W»O*«
« » . ». c
t~ •-. .
1 J
-T-r,
~D
CED
JtR
crD
CED
CED
JSK
CSD
WLW
wrv
KLK
JTF
?^;c
RKO
RKC
V TW
RJCO
RJCC
RKC
GKL




, Inc., Shiclcshinny , PA 5. Accidental Expos urfc. *~ Case dccurtentec
siry, Portland, OR IAW AR 40-5.
cords (AR 40-14) 6. Necessary to avoid exceeding quarterly
edj NU - not used. lirr.it
TO 5E RETAINED PEF.^AK=KTLY IN INDIVIDUAL'S MEDICAL ".ECCP.D
DD/JIV.,1141
                                  13.39-B

-------
                                         DARCO.N-P 3E5-1
             APPENDIX C

CROSS-REFERENCE SYSTEM FOR FILES, AND
   FLOW CHARTS FOR PROBLEM SOLVING
               13.41-C

-------
DARCOM-P 385-1
                      TABLE 13.-1.  Cress-Reference System
                 Records
       Cross-Reference
Personnel File

  Identification of the Radiation Worker

  Radiation Exposure Received During Prior
  Employment

  Exposure Received by Individuals at
  Other Installations During Current
  Employment

  Simultaneous Employment at Another
  Facility

  Training Records

  Project/Task Listing

  External-Exposure Records
  Internal-Exposure Records


  Exposure Evaluation


  Unusual  Exposures

Radiation  Protection Program File

  Licenses and Authorizations

  Radiation Protection Policies and
  Standards

  Procedures and Methods for Interpretation
  and Evaluation of Individual  Exposure Date

  Inspections and Appraisals

 • Changes  in Procedures and Methods
Training File

Project File

Instrumentation and Dosimeters
File

Instrumentation and Dosimeters
File

Radiation Protection Program
File

Accident/Incident File
                                    13.42-C

-------
                                                                  M-P 365-1
                           TABLE  13.1.   (continued)
                 Records
      Oe: :-Reference
Project File

  General Records
    -  project descripton—tiates,  location
    -  radioactive materials list

    -  instrumentation list

    -  principal investigator/project
       workers

  Standing Operating Procedures

Radiation Work Area File

  General Records
    -  location/map
    -  work area uses/equipment  and  instru-
       ments
    -  projects in the area
    -  project principal  investigator

  Radiation and Contamination Surveys
    -  date, time, location, purpose
    -  instrument identification

    -  measurement results

    -  individual(s) performing  survey

  Area  Monitoring Records
    -  date, location
    -  instrument type,  call oration

    -  source check records

    -  individual(s) operating equipment

  Airborne-Radiation Monitoring  Records
    -  date, time, location, purpose
    -  identity of sampling equipment

    -  collection efficiency

    -  counting data
    -  calculated correction factors, con-
       centrations, and  efficiency of equip-
       ment
Radiation Work Area File
Radioactive-Material Inventory
  File
Instrumentation and Dosimeters
  File
Personnel File
Instrumentation and Dosimeters
  File
Project File
Personnel File
Instrumentation and Dosimeters
  File
Radiation Protection Program
  File
Personnel File
Instrumentation and Dosimeters
  File
Radioactive Material Inventory
  File
Personnel File
Instrumentation and Dosimeters
  File
Instrumentation and Dosimeters
  File

Radiation Protection Program
  File
                                   13.43-C

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         385-1
                           TABLE 13.1.   (continued)
                 Records
Radiation Work Area File (continued)

    -  identity of eir contaminant
    -  individual  performing analysis

Instrumentation and Dosimeters  File

  Capabilities of  Dosimeters and Instruments

  Calibration and  Maintenance
  Inventory Records

Radioactive-Material  Inventory File

  Sealed Sources
    -  packing papers
    -  storage and use locations
    -  responsible department/individual
    -  projects
    -  project personnel
    -  leak test records:
       -  instrument

       -  individual
       -  results

    -  disposal history
       inventory

  Unsealed Sources
    -  packing papers
    -  storage and use locations
    -  responsible department/individual
    -  dates  of use, quantity
    -  projects
    -  project personnel
    -  disposal history
    -  inventory

  Environmental Samples
    -  •!dentifi:aiion  number,  sample typs
    -  location
    -  counting results
                                   13.44-C
       Cross-Reference
Personnel  File
Personnel  File
Radioactive-Material Inventory
  File

Project File
Radiation Work Area File
Personnel File
Project File
Personnel File

Instrumentation and Dosimeters
  File
Personnel File
Radiation Protection Program
  File
Waste Management File
Radiation Work Area File
Personnel File

Project File
Personnel File
Waste Management File
Radiation Protection Program
  File

-------
                                                              DARCOM-P  365-1
                           TABLE 13.1.   (continued)
                 Records
       Cross-Reference
 Radioactive Materiel  Inventory File
 (continued)

     -   counting  instrument

     -   disposal  history

 Waste Management File

  General Records
     -   assigned  identification number

     -   physical  description
     -   chemical  and radioisotope description
     -   radioactivity and radiation measure-
        ments
     -   previously responsible
        department/individual(s)
     -   storage location
     -   disposal  details:
        -  how, when, where
        -  responsible individual

 Transportation File

  Radioactive-Material Shipments


 Accidents/Incidents File

  General Records
     -   date and time
     -   location
     -   description, cause
     -   involved individual(s)
     -   corrective/protective actions
     -   supporting data:
        -  survey, sample results
        -  instruments, dosimeters

     -   investigator(s)

Training File

  General Records
     -  date
    -  instructor/attendees
    -  description
Instrumentation and Dosimeters
  File
Waste Management File
Radioactive-Material Inventory
  File
Personnel File
Radiation Work Area File
Personnel File
Radioactive-Material Inventory
  File
Radiation Work Area File

Personnel File
Radiation Work Area File
Instrumentation and Dosimeters
  File
Personnel File
Personnel File
                                   13.45-C

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J.AKCOM-P 385-1
                 FLOW CHART 1.  Occupational Exposure History
Problem:  Recreate staff member's working conditions and verify exposure  from
July through December 1980.
   PERSONNEL FILE

 1. Worker ID
 2. ^rejects worked
    on
 3. DD Form 114)
 4. Exposure records
    between July and
    December 1980
- PROJECT FILE
1. Project title
and number
2. Principal
investigator
3. Participants
and dates of
invol vement
4. Location of
5. Radioactive
materials
used I
RADIOACTIVE-
MATERIAL INVEN-
TORY FILE
Unsealed Sources
1. Source ID
2. Amount and
activity of
source between
July and
December 1980
Sealed Sources
1. Source ID
2. Leak Test
3. Person per-

RADIATION WORK
— APFfl FTI F

1. Laboratory or
work area
2. Project title
3. Survey records
between July and
December 1980
4. Person
performing
surveys
5. Survey
instruments


INSTRUMENTA-
TION AND
FILE
1. Instrument
ID
2. Calibration
records
3. Person
performing
calibration
                            forming  leak
                            test
                                    13.46-C

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                                                             DAKCOy-P 385-1

                    FLOW CHART  2.   Project Characterization

Problem:   Confirm or refute allegations  of misuse of radioactive meterisls

during a  specific project that  could  have resulted  in overexposures.

                           PROJECT  FILE


1.
2.
3.

PFP^nNNFl FT! F -rf

Worker ID
DO Form 1141
for specified
dates
Exposure
received
1. Project title
and number
2. Principal
investigators
3Pfl t*t i r i nant «;
and dates of
involvement
5. Radioactive
materials
t
RADIOACTIVE-
MATER1AL INVENTORY
FILE

RADIATION WORK
- ARFA FILE

1. Project title
and number
2. Survey records
and oerson per-
forming the
survey
                         Unsealed Sources

                         1.  Source ID
                         2.  Amount and
                            activity of
                            source present
                            for specified
                            dates

                         Seeled Sources

                         1.  Source ID
                         2.  Leak test
                            records
                         3.  Personnel per-
                            forming test
-  INSTRUMENTA-
   TION AND
   DOSIMETERS
   FILE	

 1.  Instrument
    ID  •
 2.  Calibration
    records
 3.  Person  per-
    forming
    calibration
                                   13.47-C

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                                                             DARCOM-P  365-1

                    CHAPTER 14.   QUALITY  ASSURANCE  PROGRAM




14.1   QUALITY ASSURANCE AND QUALITY  CONTROL   	    14.3

14.2   DEFINITIONS   .                    	14.4

14.3   IMPLEMENTATION OF QUALITY  ASSURANCE    	    14.5

      14.3.1  Who Needs a Quality Assurance  Program    ....    14.5

      14.3.2  How Extensive a Program Should Be    	    14.6

      14.3.3  Who Determines the Extent of the  Program ....    14.6

14.4   ELEMENTS OF A QUALITY ASSURANCE PROGRAM     	    14.7

      14.4.1  Organization of the Quality Assurance Program .     .     .    14.7

      14.4.2  Preparation and Documentation  of  the  Quality
              Assurance Program     	    14.8

      14.4.3  Control  of Facility Design      	    14.9

              A.  Designs for Facilities	14.9

              B.  Independent Analysis of Designs  	    14.10

              C.  Design Verification   	   14.10

              D.  Design Changes and Documentation      .     .     .     .14.11

      14.4.4  Control  of Procurement Documents     	   14.11

      1^.4.5  Instructions, Procedures, and Drawings   .     .     .     .14.11

      14.4.6  Document Control     ........   14.12

      14.4.7  Control  of Purchased Material, Equipment, and Services  .   14.12

      14.4.8  Material Identification Control     	   14.12

      14.4.9  Control  of Special Processes   ......   14.13

      14.4.10 Control  of Inspections and Tests     	   14.13

      14.4.11 Cont.-ol  of Measuring and Test Equipment  .     .     .    .14.13

      14.4.12 Handling, Storage, and Shipment     	   14.14
                                     14.1

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DARCOM-P 385-1




      14.4.13 Inspection, Test, and Operating Status   ....    K.U



      14.4.14 Nonconformance and Corrective Action     .     .      .     .14.11



      14.4.15 Quality Assurance Records 	    14.15



      14.4.16 Audits     	    14.15



REFERENCES	14.16
                                     14.2

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                                                              DARCOM-P 385-1
                    CHAPTER 14.	QUALITY ASSURANCE PROGRAM
     The  purpose of a radiation protection program is  to provide control  in
 the storage, handling, and use of radioactive material  and radiation-generating
 machines,  so as to minimize the hazard to personnel  and the general  public.
 Personnel  responsible for the radiation protection program must implement
 established regulations and meet the requirements of the facility license.
 They are  also responsible for ensuring that the radiation protection program
 accomplishes its purpose.  Consequently, a surveillance plan is needed to
 verify that activities are conducted as desired and that regulations are met.
 A quality  assurance program provides £ means of controlling the radiation pro-
 tection program and verifying that it is meeting the purposes for which it was
 established.  It allows those responsible for a program or a facility to
 ensure that the quality required for safe and reliable operation is  achieved.
     This  chapter provides a review of the elements of quality assurance and
 how they  are incorporated into a radiation protection program.  Special terms
 are defined near the beginning of the chapter, followed by a discussion of how
 a quality  assurance program is implemented—when a program is needed and how
 extensive  it should be.  The elements of a quality assurance program, including
 the purpose of each element and the activities it involves, are then reviewed.
              Section 14.1  QUALITY ASSURANCE AND QUALITY CONTROL

     Quality assurance is sometimes confused with quality control.  Quality
assurance is all of the planned and systematic actions needed to provide
adequate confidence that a structure, system, or component will perform
satisfactorily in service.  In other words, quality assurance is a planned
program for verifying that each part of the radiation protection program .is
being carried out adequately and that the total program meets its purpose.
It is the application of systematic management principles, such as pUnm'nc,
documenting, auditing, and verifying.  Quality control is the quality
assurance actions that relate specifically to the physical measurement of an
                                     14.3

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DARCOM-P 385-1

item, and it provides a means of controlling the quality of the  item to
predetermined requirements.   Quality control  is a part of  quality  assurance.
     While quality performance is the responsibility of each individual,  a
planned quality assurance program provides a method of 1) ensuring  that all
the elements necessary for adequate radiation protection have been  considered,
and 2) verifying their implementation.  The installation commander  designates
who is responsible for the quality assurance program, for example,  a Quality
Assurance Office,'Plans and Programs Office, or Program Evaluation  Office.
                           Section 14.2  DEFINITIONS

     Some terms have a specific meaning when used in quality assurance pro-
grams.  The terms used in this chapter are defined below.
  1-  Quality assurance - All of the planned and systematic actions needed to
     provide adequate confidence that a structure, system, or component will
     perform satisfactorily in service.
  2.  Quality control - The quality assurance actions that control the physical
     measurements of an item in accordance with predetermined requirements.
  3.  Analysis - The examination of a complex problem by separating it into  its
     fundamental elements.
  4.  Appraisal - The evaluation of the worth, significance, or status of a
     program or item.
  5.  Audit - A formal, documented examination of an activity or program to
     verify compliance with established requirements.
  6.  Evaluation - The determination of the worth of something by  careful
     appraisal and study.
  7.  Inspection - Examination or measurement to verify whether an item or
     activity conforms to specified requirements.
  8.  Surveillance - Monitoring or observation to verify whether an item or
     activity conforms to specified requirements.
                                      14.4

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                                                             DARCO.M-P 385-1
     Test - The determination of the capability  of an  item to meet  specified
     requirements by subjecting the iterr. to £  set of physical, cnemical,
     environmental,  or operating conditions.
               Section U.3  IMPLEMEKTATJON OF  QUALITY ASSURANCE

     Because different operations involve different degrees of  risk,  not  all
facilities or operations require the same degree  of quality assurance applica-
tion.  The level  of control and assurance necessary ft-  a  specific  facility or
operation depends upon the importance and complexity  of  the operation and its
effect on the safety of the facility, its personnel,  and the  public.   For
example, the quality assurance program for an instrument calibration  facility
requires rigorous control  and documentation to  ensure that instrument measure-
ments are accurate and reproducible; control  of radiation  sources  to  ensure
that they are traceable to nationally recognized  standards; and records to
ensure that regulations on the quality of instrument  calibration  and  frequency
are met.  In contrast,'an  operation involving the use of a commercial device
with an internal, sealed radioactive source may require  only  a  periodic
inventory to verify the location of the device, and a routine wipe  survey" to
ensure that the source is  intact and not leaking.
14.3.1  Who Needs a Quality Assurance Program
     A quality assurance program should be developed  for facilities or loca-
tions where the following  take place:
 1.  radioactive  materiel  is received, stored,  handled,  or used
 2.  radiation-generating  machines are operated
 3.  personnel radiation dosimetry is evaluated
 4.  radiation detection or measurement equipment is  procured,  received,
     repaired, calibrated, or used
 5.  facilities or equipment that will be used  for these operations  ere
     designed, constructed, or modified.
                                     14.5

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LARCOM-P 385-1
     Each facility or operation should have,  as  a  minimum,  a  written quality
assurance program that defines the extent and content  of the  program, the
records required, ana the audit activities needed  to verify implementation of
the program.
14.3.2  How Extensive a Program Should Be
     Not every quality assurance program requires  additional  staff or a
rigorous effort.  The extent of the quality assurance  program needed for a
facility or operation should be determined from  a  thorough  evaluation of the
activities to be conducted, their potential effect on  the safety of plant
personnel and the public, and the requirements of  applicable  regulations and
licenses.  The quality assurance program should  provide documented, verifiable
evidence to support the reliability and effectiveness  of the  radiation safety
program, and compliance with regulatory and 1 icense'requirements.
     Radiation protection actions for which a quality  assurance program should
be developed include, but are not limited to:
 1.  dose evaluation for all personnel who work  at the facility and for all
     visitors
 2.  receipt, inventory, shipping, and disposal  of radioactive material
 3.  radiation and contamination surveys
 4.  detection, measurement, and evaluation of airborne radioactivity
 5.  procurement, receipt, maintenance, repair,  and calibration of
     radiation detection and measurement equipment
 6.  personnel qualification, training, and retraining
 7.  radioactive-effluent releases and environmental  monitoring
 8.  facility design and modification
 9.  abnormal occurrences and investigations of  them.
14.3.3  Who Determines the Extent of 'the Program
     The extent of a quality assurance program should  be determined by the
manager responsible for the overall performance  and safety of a facility or
operation.  In most instances, this responsibility is  assigned one  level above
                                     14.6

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                                                             DARCOM-P  385-1
the person responsible for the facility's radiation  orotection  program.   Tie
person responsible should use the general guidelines provided above  to  develop
a quality assurance program es extensive as is  needed  to  assure  adequate
radiation protection.  In some instances, parts of the  quality  assurance
program may be established by regulatory requirements  or  by  the  recommendation
of the Ionizing Radiation Control Committee.
             Section 14.4  ELEMENTS OF A QUALITY  ASSURANCE  PROGRAM

     A quality assurance program is composed of numerous  elements.   Each one
is intended to provide surveillance of a major aspect  of  the  radiation protec-
tion program.  All elements may not be needed in  a  particular facility or
operation, but each should be considered when the quality assurance program
for that facility is being established.  How far  each  element of the quality
assurance program is developed depends upon the degree of control  required.
14.4.1  Organization of .the Quality Assurance Program
     A defined organizational structure should be established for the quality
assurance program to ensure the effective management of quality assurance"
activities.  The organizational structure, functional  responsibilities, levels
of authority, and lines of communication for operations affecting the quality
of the radiation protection program should be written  down.  All individuals
in the program should know what their jobs are, what authority they have to
accomplish their work, and to whom they should report  problems so that correc-
tive action will be taken.
     The person or organization responsible for developing  and implementing
the quality assurance program should be specified.   This person or organiza-
tion should have sufficient authority, access to work  areas,  and organizational
freedom to 1) identify quality problems; 2) recommend  or provide solutions to
quality problems through designated channels; 3)  verify that the solutions
have oeen implemented; and 4} ensure that any further  processing, delivery,
installation, or operation is controlled until the problem has been corrected.
The person or organization should have direct access to responsible management
at a level where appropriate action can be taken.

                                     14.7

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UARCOM-P 385-1
     The organizational structure should be designed and  indivic^al  responsi-
bilities should be assigred so that quality is achieved and  maintained by
those responsible for a job and is verified by persons or organizations not
directly responsible for the job.  In some instances,  it  may not be  possible
or practical to assign a separate organization or person  to  verify the
achievement of quality.  In those instances, the quality  assurance responsi-
bilities should be written carefully to ensure that they  do  not conflict with
the job responsibilities of the individual assigned to carry out multiple
duties.  It may be appropriate to have an outside organization provide quality
assurance.  In all cases, the individual's or organization's responsibility
and authority should be clearly defined and documented.
14.4.2  Preparation and Documentation of the Quality Assurance Program
     The quality assurance program should be documented  as a means of defining
the program, providing a basis for review, and ensuring  continuity.   Adequate
planning is needed before the quality assurance document  is  written to ensure
that all necessary elements have been included.  The program should:
 1.  provide control over operations affecting the quality of the radiation
     protection program, to whatever extent is consistent with the importance
     of those operations
 2.  Identify the operations, processes, and equipment to which the program
     applies
 3.  include consideration of the technical aspects of quality assurance
     actions
 4.  be established es early as possible consistent with the schedule for
     accomplishing the operation
 5.  provide for any special controls, processes, test equipment, tools, and
     skills needed to attain the required quality and for necessary
     verification of quality
 6.  provide for the training of personnel performing operations that affect
     quality, to ensure that they can do the job adequately
                                     14.8

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                                                             DARCO.M-P  385-1
 7.  provide for regular management assessment  of  the  adequacy  and  effective
     implementation of the Duality assurance  program.
     Not ell operations in a facility or  program  require  formal  quality assur-
ance consideration.  Those operations that  are  important  for  adequate  radiation
protection and/or that must be performed  consistently  should  be included in
the quality assurance program plan.  The  program  should  specify the qualifica-
tions, training, and skills required of quality assurance personnel;  the type,
frequency, and method of audits, inspections,  and  tests  for the assurance of
quality; and the system for reporting, correction,  and follow-up on any unsatis-
factory condition that may be identified.   If an  extensive quality  assurance
program is necessary, additional details  on planning the  program may oe found
in the American National Standards Institute's  (ANSI's)  standards and in the
Nuclear Regulatory Commission's (NRC's) regulatory guides.  (See the bibliog-
raphy at the end of this manual.)
14.4.3  Control of Facility Design
     It is particularly important that radiation  protection requirements be
included in the design of new facilities  or the modification of existing facil-
ities in which radioactive material will  be stored, handled,  or processed or
in which radiation-generating machines will be operated.   Engineered features
for controlling radiation and contamination are most cost-effective, and some
are only feasible, when included in the original  design and construction or  in
a major modification of a facility.  The  design for facilities should there-
fore be defined, controlled, and verified to ensure that radiation protection
requirements are met, that the design is  approved by appropriate authorities,
and that construction meets the design specifications.
     A.  Designs for Facilities.  Appropriate design bases, performance  require-
ments, regulatory requirements, and codes and standards should be  identified
and documented, and their selection reviewed and  approved.  For  radiation pro-
tection purposes, the ventilation criteria, shielding provisions,  equipment
reliability and maintenance, personnel traffic patterns and occupancy zones,
and waste-handling systems must be reviewed specifically for how well they
will protect personnel and keep radiation doses as low as is reasonably  achiev-
able (ALARA).   If designs are changed, the changes and the reason  for the
                                     14.9

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DARCOM-P 385-1

changes should be identified, approved,  and documented  to  ensure  that altera-
tions that could affect radiation protection are  adequately  reviewed.   The
purpose of quality assurance in this process is to  verify  that  these  steps  are
taken and that reviews and approvals are completed  by  appropriate personnel.
     The organization responsible for the design  should document  its  actions
in enough detail so that the design process can be  carried out  and it is
possible to verify that the design meets requirements.   The  design of the
facility, and the' materials, equipment,  and processes  that are  essential to
radiation protection and exposure control, should be selected and reviewed  for
suitacility of application.
     B.  Independent Analysis of Designs.  It may be advisable  to provide for
an independent analysis of the design of facilities and equipment, to ensure
that all factors have been considered and that the  'resulting designs  are
correct.  The independent analysis should be performed by  individuals who are
technically qualified in the subject and independent of the original  designers.
These persons may vary from electrical experts, who ensure that electrical
load-carrying capacities are adequate, to health  physicists, who ensure that
shielding factors for shielding casks are correct.
     The analysis should be documented in enough  detail so that a person
technically qualified in the subject can review the analysis and verify the
findings.  Documentation of an analysis should include the purpose of the
analysis, pertinent sources of data and supporting information, and review and
approval.  Here again, the quality assurance function is to verify that
analyses, reviews, and approvals required as part of the quality assurance
program have been completed.
     C.  Design Verification.  Designs for important facilities should be
verified to ensure that the design was performed  correctly and that the final
product as provided for in the design will perform the function described  in
the design criteria.
     Designs should be verified by competent personnel other than those who
drew up the original design.  The design verification results should be
documented and the verifier identified.   The extent of the design verification
                                     14.10

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                                                              DARCO.M-P 38 5-1
 required depends on the importance to safety  of  the  item  under consideration,
 and verification methods may include analyses, simple  reviews, alternate
 calculations, and/or qualification testing.
     D.  Design Chances and Documentation.   Once c design has  been approved,
 any changes, including field changes, should  be  controlled in  the same manner
 as the or-iginal design.
     Tne design documentation and records,  which provide  evidence that the
 facility was designed and the design was verified as  required, should be  gen-
 erated and maintained in accordance with documented  procedures.
 14.4.4  Control of Procurement Documents
     A fourth function of a quality assurance program is  to ensure that docu-
 ments generated for the procurement of items  or  services  include enough infor-
 mation (applicable design bases, technical  requirements,  specifications,
 drawings, instructions, etc.) so that the items  being procured will be ade-
 quate in quality; they must fit, work properly,  end  do the job required.
     The procurement documents should identify the means  (tests, inspections,
 documentation) that the purchaser will use  to determine the acceptability of
 the items.  If certain aspects of acceptability  cannot be determined at .this
 point, the procurement documents should specify  the  quality assurance require-
 ments necessary in the supplier's plant.
     Procurement documents and changes to them should be  reviewed and approved
 by the purchaser tc ensure that they are clear and detailed enough so that the
 supplier can provide the items or services  that  meet the  specified requirements,
     Depending upon the type and use of the item or  service being procured, it
 may be necessary to include in procurement  documents the  requirement that
 suppliers also have and implement a documented quality assurance program.
 14.4.5  Instructions, Procedures, and Drawings
     Operations that affect the quality of  the radiation protection program
must be reproducible, and complex operations  should  be performed in accordance
with documented instructions, procedures, or drawings as appropriate to  the
circumstances,  to ensure consistent and adequate performance.   Such operations
include tests,  equipment control, calibration of instruments,  and surveys.

                                     14.11

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DAKCOM-F 38 >-i
14.4.6  Document Control
     Documents tnat specify quality  requirements  or  prescribe  operations
affecting quality should be prepared,  issued,  and changed  in a  controlled
manner to ensure that correct documents  are  being used.  These  documents,
including changes to them, should be reviewed  for adequacy and  approved for
release by authorized personnel.
     The document control system should  identify  which  documents  are  to be
cortrolled; who is responsible for preparing,  rev-iewing, approving, and
issuing them; how their adequacy, completeness, and  correctness is  to be
ensured prior to issuance; and the methods of  ensuring  that documents in  use
are current and that outdated or inappropriate documents are removed  from use.
14.4.7  Control of Purchased Material, Equipment, and Services
     The procurement of material, equipment, and  services  should be controlled
to ensure conformance with the requirements  specified in  the procurement
documents.  Procurement operations should be planned and  documented and  should
include the preparation and review of procurement documents and control  of
changes to them (see item 14.4.4 above); selection of procurement sources;  the
evaluation of bids and the award of a contract;  purchaser  control of supplier
performance, if warranted by the circumstances;  any necessary verification
actions, including surveillance, inspection, or audit of the supplier; plans
for controlling and disposing of material,  equipment, or services that do not
meet requirements; methods of correcting problems occurring in the procurement
process; acceptance of material, equipment,  or services;  and the quality
assurance records needed.  Most purchased material, equipment, and services
should be inspected when they are received from the supplier to ensure that
they meet the requirements of the procurement  documents and the purpose for
which they were purchased.
14.4.8  Material Identification Control
     Controls should be established to ensure that only correct and accepted
items are used or installed.  Identification should be maintained either on
the items or in documents traceable to the items.
                                     14.12

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                                                              DARCOM-P 385-1
      Items  with  a  limited calendar or operating life should be identified and
 controlled  to  prevent their use after their life has expired.   For instance,
 batteries,  some  adhesives, rubber products, chemicals, end radioactive sources
 mey  degrade  in storage as well as in use and may need to be controlled to
 ensure  their effectiveness when needed.
 14.4.9   Control  of Special Processes
      Measures  should be established and documented to ensure that special
 processes such as welding, heat treating, cleaning, nondestructive examina-
 tions,  and  analytical evaluations are carried out by qualified personnel and
 under controlled conditions, in accordance with applicci-le codes, standards,
 and  specifications, and other special requirements.  The qualifications of
 personnel performing special processes should comply-with the requirements of
 applicable  codes and standards. If no such codes or standards exist, the
 requirements for personnel qualifications should be defined and documented.
 14.4.10 Control of Inspections and Tests
      Inspections and tests to verify that an item or operation conforms to
 specified requirements should be planned and documented.  The characteristics
 to be inspected or tested, the methods of inspection or testing, and the cri-
 teria for evaluating the results and documenting whether the item or operation
 is acceptable should be identified.
      Records of inspections and tests should include the identity of the item
 or operation involved, the date, the name of the inspector, the type of
 inspection or test given, and the results.
 14,4.11  Control  of Measuring and Test Equipment
     Tools,  gauges, instruments, and otner measuring and test equipment used
 for operations  effecting quality should be controlled to ensure that they meet
 the defined  specifications,  are used as designed, and provide the necessary
quality of measurement and test data.  Meesuring and testing equipment should
be of the type, range, accuracy, and tolerance needed to accomplish the
function intended.   At prescribed intervals, or before its use. or whenever
 its accuracy is suspect,  measuring and test equipment should be calibrated and
adjusted against  certified equipment that has known valid relationships to
nationally recognized  standards.
                                     14.13

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DARCOM-P 385-1
     Devices that are out of adjustment  should  be  tagged or  segregated -and  not
used until they r.sve ! een recalibrated.   Equipment  should  be  ;.roper1y handled
and stored to maintain its accuracy.   Records  should  be kept  end equipment
should be suitably marked to incicate its calibration status.
14.4.12  Handling, Storage, and Shipment.
     To prevent damage or deterioration  of material and equipment,  measures
should be established for their handling, storage,  shipping,  cleaning,  and
preservation in accordance with work  and inspection instructions.   When
necessary for particular products,  special  protective environments  such as  an
inert-gas atmosphere, specific temperature levels,  absorbent  material,  and
shielding should be specified and provided.
     Instructions for marking and labeling items  for  packaging,  shipping,  hand-
ling, and storage should be established.  Any  need for special  environments or
controls should be indicated on the label.
14.4.13  Inspection, Test, and Operating Status
     Measures for identifying the inspection and  test status  of equipment
should be established and documented.  The status should  be  known  throughout
the manufacturing, installation, and  operation of the equipment.   The  inspec-
tion and test status should be maintained through the use  of status indicators
such as physical location, tags, markings, stamps, or inspection and test
records.  Only items that have passed the required inspections  should  be
installed or operated.
     Procedures should be developed to ensure  that operations are  conducted in
accordance with applicable documented instructions and procedures  and  that
items perform satisfactorily in service.  Measures such as tagging should  also
be used to indicate the operating status of systems and components, and to
prevent any inadvertent, unplanned use of equipment.
     The emergency response capability of the  facility (personnel  and
equipment) should be included in this program.  Annual testing of emergency
response mist be conducted and adequate performance verified.
                                     14.14

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                                                             DARCOM-P  385-1
14.4.14  Nonconfonnance and Corrective Action
     Controls should be established to ensure that  failures,  malfunctions,  and
defects in equipment and nonconformances to procedures  and  processes  are
promptly identified and corrected.   In the case of  a  significant  problem,  the
controls should ensure that the cause of the problem  is determined and  that
corrective action is taken.  The problem,  its cause,  and the  corrective
actions needed should be documented and reported to appropriate  levels  of
management.  Follow-up action should be taken to verify implementation  of
corrective action.
     Items that do not conform to requirements should be controlled to  prevent
their inadvertent use.  Control provisions should include identifying and
disposing of the items and notifying affected organizations.
14.4.15  Quality Assurance Records
     Records should be kept identifying operations  that affect  quality  and
showing that regulatory and license requirements have been  met.   The records
should be legible, identifiable, and retrievable, and should  be  protected
against damage, deterioration, or loss.  Quality assurance  records should  be
centrally maintained by the individual or organization assigned  the
responsibility for the quality assurance program.  Alternate  designees  may be
acceptable.  However, in all  programs, requirements and responsibilities  for
record transmittal, distribution, retention, maintenance, and disposition
should be established and documented.
     The types of records needed for verification of  the radiation protection
program include radiation exposure  records, bioessay  date,  radiation and con-
tamination survey reports, calibration records, and training  records.  A more
complete listing of required  radiation protection records is  located in ANSI
N13.6-1972.
14.4.16  Audits
     Audits by personnel responsible for quality assurance should be scheduled
periocicsMy -..spending on the importance ?~ the activity being audited) to
verify compliance with all aspects  of the quality assurance program and to
determine the effectiveness of the  program.  Trained  auditors who are  not
                                     14.15

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DARCOM-P 385-1
directly responsible for the areas being audited should follow written pro-
cedures or checklists.  Audit results should be documented and reviewed by
responsible management, and any necessary follow-up action should be taken.
                                  REFERENCES
American National, Standards  Institute (ANSI).  1972.  Practice for Occupational
  Radiation Exposure Records Systems.  ANSI N13.6-1956, Rev. :972, New YorK.
                                      14.16

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                                                              DAKCOK-P 385-1
            CHAPTER 15.   APPRAISAL  OF  RADIATION  PROTECTION  PROGRAMS
15.1   CONDUCTING A TECHNICAL  APPRAISAL   	    15.5
      15.1.1   The Appraisal  Team   ........    15.6
      15.1.2   The Appraisal  Process      ...     .     .     .     .     .    15.6
      15.1.3   Report  of Appraisal  Findings    ......    15.7
15.2   PROGRAM AREAS THAT SHOULD BE APPRAISED  	    15.8
      15.2.1   The Radiation  Protection  Organization     ....    15.8
      15.2.2   The Selection  and Training of Personnel   ....    15.10
              A.   Selection  of the Radiation  Protection Staff     .     .    15.10
              B.   Routine Training Programs  .     .     .     .     .     .15.11
              C.   Emergency  Preparedness Training .     .     .     .     .15.13
      15.2.3   The Radiation  Survey Program   	    15.13
              A.   Responsibilities and  Scope  of the  Program .     .     .    15.14
              B.   Instrumentation  Suitability and Use  .     .     .     ." 15.14
              C.   Records	15.15
      15.2.4   The Program for Internal-Exposure Control     .     .     .    15.16
              A.   Exposure Limitation Methods     .     .     .     .     .15.16
              B.   Dosimetry  Program     	    15.18
              C.   Exposure Review   .     .     .     .     .     .     .     .15.19
              D.   Quality Assurance Program for Internal  Dosimetry    .    15.19
      15.2.5   The Program for External-Exposure Control     .     .     .    15.20
              A.   Exposure Limitation Methods     .....    15.21
              E.   Dosimetry  Program     	    15.21
              C.   Exposure Review	15.22
              D.   Quality Assurance Program for External  Dosimetry    .    15.22
                                     15.1

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DARCOM-P Jb5-
      15.2.6  The ALARA Program    	   15.22
      15.2.7  Facilities and Equipment  	   15.23
              A.  Facilities	15.23
              B.  Protective Equipment  	   15.24
      15.2.8  Management of Radioactive Waste     	   15.25
      15.2.9  Records and Audits   	   15.25
15.3  CHECKLIST OF QUESTIONS FOR APPRAISING A RADIATION PROTECTION
      PROGRAM	15.25
      15.3.1  The Radiation Protection Organization    ....   15.26
      15.3.2  The Selection and Training of Personnel  ....   15.27
      15.3.3  The Radiation Survey Program   .    .    .    .     .     .15.28
      15.3.4  The Program for Internal-Exposure Control     .     .     .   15.29
      15.3.5  The Program for External-Exposure Control     .     .     .   15.31
      15.3.6  The ALARA Program    	   15.32
      15.3.7  Facilities and Equipment  	   15.33
      15.3.8  Management of Radioactive Waste     	   15.33
      15.3.9  Records and Audits   	   15.34
15.4.  NETWORK TECHNIQUES FOR PLANNING APPRAISALS 	   15.35
      15.4.1  The Function of Networks and Logic Trees .    .     .     .15.35
      15.4.2  Using Logic Trees to Plan a Radiation
              Protection Appraisal 	   15.38
REFERENCES	;	15.43
                                     15.2

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                                                              DARCOM-P 385-1
                                    FIGURES


15.1  Critical Paths for an Adequate Survey  	   15.36

15.2  MOF.T Logic Tree	15.37

15.3  Logic Tree for a Process To Be Operationally Ready    .     .      .   15.37

15.4  Radiation Protection Program, rirst Tier    	   15.38

15.5  Radiation Protection Program, Second Tier—Routine
      Operations    ...........   15.39

15.6  Radiation Protection Program, Second Tier—Emergency
      Operations	15.39

15.7  Upper Tiers for Logic Tree Depicting Radiation
      Protection Program ..........   15.40

15.8  Logic Tree Tiers Depicting Internal-Exposure Control
      Program	15.41
                                     15.3

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                                                              DARCOK-P 385-1
            CHAPTER 15.   APPRAISAL  OF RADIATION PROTECTION  PROGRAKS
     A variety of methods can be used to ensure  that e  radiation  protection
program is providing a reasonable degree of safety.   One  method,  a  quality
assurance program, was discussed earlier.   This  chapter focuses on  the  use of
a comprehensive appraisal.
     An appraisal' is a means of comprehensively  evaluating the  overall  ade-
quacy and effectiveness of the radiation protection  program.   Unlike a  com-
pliance inspection, which is an evaluation of a  program by discrete subject
areas, an appraisal is an integrated look at the total  program.   That is, it
looks at the total program needs, not just at regulatory  compliance.  It is
focused on identifying and correcting the underlying causes of deficiencies
rather than on identifying failures to follow specific  procedures or
regulatory requirements.
     The routine appraisal of a radiation protection program entails verifying
that the program is effective in protecting personnel,  property,  and the
environment.  This goal can be accomplished through  a thorough, technical
health physics appraisal  by experts, and follow-up by management  to ensure
that any problems found during the appraisal have been  corrected  and that
staff members are being protected.
     This chapter includes a brief overview of the steps  for conducting a tech-
nical appraisal; a discussion of the areas -of a  radiation protection program
that should be included in an appraisal; a checklist of questions for use dur-
ing appraisals; and an introduction to network techniques that can be used to
help plan an appraisal.


                Section 15.1  CONDUCTING A TECHNICAL APPRAISAL

     A thorough technical appraisal usually begins with the selection of  a
team of individuals who are familiar with the requirements of a health  physics
program and with applicable standards and regulations, and who have  the  ability
                                     15.5

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DARCOM-P 385-1
to conduct appraisals.   The team members  review  site documents and  then  con-
duct an onsite appraisal  that includes discussions with personnel,  observation
of work practices,  and  reviews of procedures  and reports.  Their  findings  are
then reported in writing.
15.1.1  The Appraisal  Team
     In order to characterize the radiation protection program and  identify
any deficiencies, it may be necessary to  expand  the onsite staff  with  addi-
tional personnel who are experts in the field of health physics or  to  bring in
a team of outside consultants who have experience in broad-based  health  physics
programs and who have conducted appraisals.  This expanded-team approach,  using
outside expertise,  provides objective viewpoints and reduces  the  time  required
for the technical appraisal.  In addition, a  team approach allows team members
with varying backgrounds to interact as they  investigate  deficiencies  and  rec-
ommend solutions.  Their interactions and discussions  can help  identify  prob-
lem areas and clarify the causes of symptomatic  deficiencies.
     The members of the appraisal team should be selected based on  the type
and size of the radiation protection program to  be evaluated.   Each team
member should have  both a broad and thorough  knowledge of health  physics,  and
an area of expertise that complements those of the rest of the  team.   The
appraisers should be familiar with current standards,  regulatory  guides, and
regulations, and should have shown through prior appraisal experience  that
they have an aptitude for conducting appraisals.
     A leader of the appraisal team should be selected.   This individual
disseminates documents, briefs the installation  commander, assigns areas of
responsibility to other team members, and functions  as the team coordinater to
ensure that all areas are covered.
15.1.2  The Appraisal  Process
     The appraisal  process begins with a  thorough review  of  site  documents,
which are distributed to the appropriate  team members  by  the appraisal team
leader.  These documents should include:    1) the operating  license, 2)  the
environmental impact statement or environmental  analysis, 3) program objec-
tives, 4) related missions, 5) organizational charts,  6)  job descriptions,
                                     15.6

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                                                             DARCOX-P 385-1
7) performance objectives, 8) training records,  9)  work  utilization,  schedul-
ing, and buoget documents, 10) radiation safety  manuals,  11)  health physics
procedures, 12) chemistry procedures, 13) respiratory protection programs,
U) applicaole regulations, 15) the emergency plan, 16)  procedures  for imple-
menting the emergency plan, 17) dosimetry records,  18) survey records,
19) minutes of meetings of the radiation protection committee,  20)  reports  on
previous inspections and appraisals, and 21)  any other documents needed to
complete the appraisal.  During the review and preparation period,  each team
member should:
 1.  review the documents received from the team leader to identify tasks  that
     ere crucial  to detecting and assessing radiation levels, notifying
     appropriate staff and officials, and implementing protective action
 2.  identify the individuals responsible for crucial tasks
 3.  identify the minimum equipment, procedures, and instruments required  for '
     the performance of those tasks
 4.  identify any deficiencies in standard operating procedures (SOPs)
 5.  identify any deficiencies in emergency plans and procedures.
     The time planned for the appraisal should be long enough to allow the
team to talk with installation personnel and radiation workers, review and
observe work practices, and review onsite radiation protection procedures  and
records relating to exposures, incidents, etc.  The appraisal team should  also
meet with the installation commander and^any other managers between the radia-
tion protection staff and the commander, to ensure that the radiation protec-
tion staff has sufficient support to carry out the routine ALARA program
(keeping exposures es low as is reasonably achievable) and to handle any
abnormal occurrences.
15.1.3  Report of Appraisal Findings
     At the completion of the appraisal, a report should be written specifying
whether eac!~ major component of the radiation protection program was  found  to
be adequate.  The total program should also be rated as acceptable, adequate
for present operations but having significant weaknesses,  or not acceptable.
                                     15.7

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DARCOK-P 385-1

 Deficiencies  or weaknesses are considered significant when they have a direct
 effect  on  the  level of protection provided or when they play a critical part
 in whether a  portion of the program is judged acceptable.   For example, fail-
 ure  to  calibrate  instruments or provide adequate dosimetry would be a signifi-
 cant deficiency that would make all or part of the program unacceptable
 depending  on  the  necessity of the devices to the overall safety of the pro-
 gram.   Isolated weaknesses and minor problems should not be judged as
 representing  a significant finding.  However, if a number of deficiencies are
 found within  a particular phase of the program, then an assessment that
 significant problems exist may be warranted for that phase.  If t deficiency
 or weakness requires immediate attention, the problem should be discussed with
 the  Radiation  Protection Officer (RPO) and the cognizant manager, and an
 immediate  solution  should be agreed on.
              Section  15.2  PROGRAM AREAS THAT SHOULD BE APPRAISED

      The  elements that make up an effective radiation protection program are
 the  radiation protection organization, the selection and training of personnel,
 survey  programs, programs for the control of internal and external exposure,
 the  ALARA program, facilities and equipment, waste management, and records and
 audits.   Some of the  aspects of each area that should be covered both in manage-
 ment reviews  and in technical appraisals by health physics experts (members of
 the  appraisal  team) are discussed below.
 15.2.1  The Radiation Protection Organization
      The  appraisal of a health physics program begins with an evaluation of
 the  radiation protection organization.  Both onsite and offsite support for
 the  radiation protection program should be reviewed.  For example, if the
 Ballistics Research Laboratory has an agreement with ARRADCOM, Dover, New.
 Jersey, or with the Material Test Directorate, Aberdeen, Maryland, to provide
 health  physics support either routinely or during emergencies, then the
 appraisers should ensure that the supporting organization is aware of the
                                      15.8

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                                                             DARCO.M-P 3E.5-1
magnitude of support needed by the requesting  organization.  The  RPO's man-
eaement should ensure thet there are written agreements delineating  responsi-
bilities.  Additionally, if offpost use of  radionuclides  is authorized,  the
appraisers should review the procedures and licenses  involved  to  en:ure  thet
they ere adequate.
     To ensure awareness of responsibilities,  an  organizational chart  depict-
ing the onsite and  offsite radiation protection organization,  together with
tne total command,  should be available to everyone.   This chart should clearly
snow that the RPO has a direct reporting chain to the base commander.  The
purpose of this direct access is to ensure  the authority  to stop  work  in the
event of potential  or actual hazardous situations.   A written  statement  of  the
duties, authorities, and responsibilities of the  RPO and  the  radiation protec-
tion staff should also be available.  If contractors and  private  organizations
(e.g., fire departments or hospital emergency  staff) provide  technical assis-
tance to and augmentation of the emergency  organization,  they  should be
specified and their roles clearly defined.
     Within the radiation protection organization itself, authorities  and
responsibilities should be clearly assigned.   Job descriptions are frequently
useful in delineating the scope of responsibilities  and  ensuring  a thorough
transition during staff turnovers.  The appraisers should also ensure  that the
radiation protection staff feel they have the  authority  to implement the
radiation protection program.  The management  review would include e check to
ensure that the RPO and the staff have job  descriptions,  are  aware of their
responsibilities, and are fulfilling those- responsibilities.
     The responsibility for preparing emergency  plans and procedures is  fre-
quently assigned to en individual, in addition to his or her  primary duties,
without any allocation of the authority, manpower, time,  or money needed to
accomplish the task.  Because the emergency planning program involves a  number
of persons and organizations, the extent of emergency planning necessary at
each site should be carefully evaluated, and the  organizations participating
in the planning should be aware of who in the  radiation  protection organiza-
tion is responsible for coordinating the program.
                                     15.9

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 D/vROOM-P  385-1
     The appraisers should ensure that the staff of  managers,  supervisors,  and
radiation workers is adequate for the  amount  of  radiation  work performed at
the site, both for operations during the day  and for operations  after normal
working hours.  There should be enough radiation protection  technicians  to
perform assigned responsibilities for  routine operations,  and  at installations
with a large radiation work force, the staff  of  radiation  protection techni-
cians should include specialists in such areas as dosimetry, respiratory pro-
tection, and ALARA review.  The technical  support personnel  should be relieved
                  \
from clerical duties as much as possible by administrative support personnel,
especially during emergencies.  There- should  be  emergency  plans for supple-
menting the radiation protection staff within 18 hours of  a  major accident.
This procedure will reduce the potential for  mistakes caused by fatigue.
15.2.2  The Selection and Training of"Personnel
     The quality of the radiation protection  program depends or, the qualifica-
tions of the RPO and on the support the RPO receives from management and the
staff.  During an appraisal, therefore, the appraisers should review the
criteria used to select the site's RPO and radiation protection staff, verify
that the RPO and the staff meet these criteria,  and assess the programs used
to train personnel.
     The routine management review should include verification that there  are
job descriptions for the RPO and the radiation protection staff.  These de-
scriptions should be discussed with the individuals to ensure that  they are
up-to-date and accurately reflect the current work assignments.   In conjunc-
tion with the work assignment, emergency and routine training should be
reviewed.  This review would include verification that:
 1.  training classes are scheduled
 2.  training is provided as specified
 3.  radiation workers receive annual  training
 4.  training records are up-to-date.
     A.  Selection of the Radiation Protection  Staff.  The  criteria used in
selecting a site's RPO should be based on the type of work  concocted at  the
installation and the size and type of radiation  program involved.   However,  in
all instances, the qualification criteria should  include  consideration  of  the
                                     15.10

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                                                             DARCOM-P 385-1
individual's formal education,  continuing  education, work experience,  previous
management experience, and technics"'  understanding of health physics.  The
individual selected should have demonstrated  experience  in the area  that he  or
she is to manage.
     The RPO should be responsible for developing selection criteria  for each
position in the radiation protection  organization and for selecting  the
technicians who help run the program.   The appraisers should ascertain whether
tue selection criteria are related to the  individual jobs and whether they
include an assessment of formal education  and experience.  These  criteria
should be used for both hiring  and promotions,  and the  staff should  be aware
of the promotion requirements.
     B.  Routine Training Programs.   The members of  the  appraisal  team who
are responsible for appraising  a site's training program should  have consider-
able experience in radiation protection training.  This  experience is neces-
sary in determining whether the training provided is adequate  in  content,
nature, and length.  The training must be  assessed against  10  CFR 19, against
the training requirements for and the complexity of  a  program,  and against  the
authority for the program.  Consideration  must be given to  the  type  of work
authorized for and conducted on the site.
     The training program should be assessed  in two  parts:   training for radia-
tion workers and other staff members, such as medical  personnel,  public  informa-
tion officers, and security support staff, and training for the  radiation
protection staff.  Training for both  groups should  include  the  following:
 1.  2 defined scope and written content for  the  program
 2.  instructors qualified in the subjects they are  teaching
 3.  instruction schedules and  lesson plans
 4.  objectives for trainee performance
 5.  demonstration of standards attained by trainees
 £.  frequency of required attendance
 7.  documentation of ettencance (including test  results, dates, subjects,
     etc.).
                                     15.11

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DAJU •>!-•? ".85-1
 corma]  on-the-job retraining should be provided periodically for all
 individuals.
      The following topics  should be included in each training program:
  1.   tne specific duties and responsibilities of those being trained
  2.   the site's  reporting  or communications chain
  3.   site-specific or job-specific hazards
  4.   industrial  and  radiation  safety
  5.   special  procedures
  6.   special  protection  (e.g.,  the use of respirators and protective
      clothing)
  7.   the ALARA philosophy.
 Training should  include  instruction in the capabilities and limitations of any
 instruments to be used.  Special procedures and the reasons the procedures are
 needed  should be written down  and explained to everyone involved.
      An adequate training  program should not consist solely of classroom  in-
 struction, demonstrations  of equipment to the group, and the use of maps  or
 building drawings to point out the location of equipment, work stations,  or
 emergency  response duty  stations.  Rather, training should include  hands-on
 use of  equipment and tours of  areas that the trainees may need to enter in the
 course  of  their  work.
      The individuals evaluating the training program should attend  the train-
 ing classes to verify the  level  of instruction.  Their evaluation should  also
 include a  thorough review  of class records for the  previous 2 years,  discus-
 sions with randomly  selected individuals to verify  that they received and
 understood the training  shown  in their records, evaluation of the training
 aids  used., and discussions with the instructors, the supervisors of radiation
 workers, the  radiation protection staff, and the RPO.   In evaluating  training
 for the radiation protection staff, the appaisers should also ascertain
 whether the operators of counting and analysis systems  are qualified  to
 operate them  and are using them properly.  The appraisers should verify that
                                      15.12

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                                                              DARCOM-P  385-1
when  new  instrumentation is put into use, the staff is retrained in the state
of  tne art  for that instrumentation and the range of its capabilities.
      C.   Emergency Preparedness Training.  Training the staff, especially the
radiation protection staff, for emergencies is extremely important because
emergency situations precipitate changes in reporting chains, the scope and
nature of duties, and the perceptions of individuals.  Individuals under
stress may  revert to established behavior patterns; training can help estab-
lish  patterns appropriate to emergencies and eliminate the randomness of
purpose that is characteristic of such situations.  The appraisal of the
emergency preparedness training program should involve ascertaining whether
individuals will respond appropriately when under stress.
      The emergency preparedness training program should contain provisions for
training the members of support organizations (e.g., the fire department and
ambulance service).   The purpose of the training should be to ensure mutual
understanding of roles, procedures, and interfaces.  Although the command
cannot always require offsite groups to participate in training sessions, the
appraisers  should assess the capabilities of these groups to support the RPO
and the radiation protection staff in emergencies.
15.2.3  The Radiation Survey Program
      The purpose of the radiation survey program is to evaluate actual or
potential radiation hazards at facilities where radiation sources are used.
The scope of survey activities should be clearly stated for all installations.
      The primary emphasis of management's review of the survey program should
be verification tnat survey procedures exist in written form and thet the pro-
cedures are followed when surveys are conducted.  Records should contain the
latest surveys  and include all required information.  The instrument storage
facility, the general  condition of the instruments, the condition of the emer-
gency kit, the  records showing dates for instrument calibration, whether the
dates are being met,  and whether outdated instruments are being used should
also be reviewed.
                                     15.13

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DARCOh P 38.;.-1
     A.  Responsibilities and Scope of the Drogram.   The RPO should be respon-
sible for the design, development, and maintenance of the survey program and
should ensure that there are procedures for performing routine and periodic
surveys of airborne and surface contamination.  The extent of the surveys
should be consistent with the hazards and work conditions at the installation.
If any special or unusually complex surveys are performed by an offsite team
or consultant, the RPO should ensure that the agreement for this work is well
defined and should specify the individual or individuals responsible for mon-
itoring the work.
     The appraisers should ensure that the RPO has adequately defined the
scope of the survey program to include all potential radiological hazards at
the installation.  They should review the scope and the frequency of the
survey routines to ensure that they are adequate for the needs of the program
and consistent with regulatory requirements.  The appraisers should also
determine whether the radiation protection staff and/or the RPO review the
routine and periodic survey data and assess the need for possible additional
actions.
     B.  Instrumentation Suitability and Use.  The appraisers should deter-
mine whether the instrumentation used in the survey program meets the minimum
standards required by regulations and the site's license.  The instrument tech-
nician or RPO should be required to demonstrate that the quantity, type, range,
and sensitivity of portable instruments are sufficient for the scope of  rou-
tine and nonroutine health physics activities.  Instruments should provide the
type of measurements required for the program.
     The appraisers should evaluate the calibration program using reference
documents such as Standard N323-1978 of the American National Standards
Institute (ANSI).  This standard is a general document on instrument testing
and calibration that contains extensive technical information.   It includes
functional testing criteria and calibration methods, and specifies sources,
calibration facilities, calibration frequencies, and required records.   All
calibration sources she-Id be traceable to the National Bureau of Standards
(NBS).
                                     15.14

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                                                             UARCON-F 385-1
     The RPO should be responsible for ensuring  that  a  thorough  evaluation  of
the best location is maoe before a fixed  or  semifixed instrument is  set  up.
These instruments should be nositioned to allow  for  ease  in  operational
checks, calibrations, and maintenance.
     Monitoring for airborne radioactive  materials  should involve the  use of
breathing-zone samplers, area air samplers,  portable  and  semiportable  air sam-
plers, and grab air samples.  The appraisers should  observe  several  air  samples
being taken and should review air-sampling records  to ensure that the  proper
procedures are being followed and that the air samples  taken are representa-
tive of the air being breathed by workers.  If the  persons taking the  samples
fail to consider air currents and the dilution and  turbulence caused by  work
activities, the samples taken may not represent  the  air being breathed.
     The appraisers should determine whether emergency kits  and  survey instru-
ments have been placed at appropriate locations.  If so,  the instruments
should be evaluated to determine their suitability  for each  location.
     The RPO should be responsible for specifying the methods and equipment to
be used for routine surveys of offsite locations and for emergency offsite
radiological surveys.  For all onsite and offsite surveys, each  member of the
survey team should be required to record  the following information:
 1.  date and time of each survey
 2.  location of each survey
 3.  name(s) of the individual(s) who performed  the survey
 4.  the instrument used, identified by type and serial number
 5.  the mode in which the instrument was used (i.e., window open or closed)
 6   the duration of the meter or instrument reading
 7.  air sampler flow rates
 8.  background radiation levels at the time of eir sample counting
 S.  sample count time
10.  work condition at the time of sampling.
     C.   Records.  The appraisers should  ensure that the RPO verifies the
documentation of all surveys.  Survey reports should be clearly written and
traceable as to instrument, date, time, location, and project.  The records of
SOPs should correctly reflect the job and work conditions.  The appraisers
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should thoroughly review the records  to  determine whether  the RPO ensures  that
survey results are distributed to staff  members  and  supervisors  as  necessary.
15.2.4  The Program for Internal-Exposure  Control
     Management review of the program for  internal-exposure  control  recuires
both a walk-through and a review of records.   The purpose  of the walk-through
is to ensure proper posting; proper cleanup of contaminated  areas;  proper  stor-
age of respirators (individually wrapped and stored  in  a  closed  container);
proper wearing of respirators; storage of  radiaoctive  liquids in nonbreakable
containers or in locked storage containers; proper wearing of lapel  air sam-
plers; proper positioning of breathing-zone samplers;  and  proper storage of
air-monitoring equipment.  In addition to  the walk-through reviews,  management
should ensure that internal-dosimetry records are maintained for everyone  who
has received an internal dose, is suspected of having  received  an  internal
dose, or has entered an area containing  airborne radioactivity,  whether with
or without a respirator.  The manager should also  review  individual  dose
records to ensure that they are up-to-date; calibration records  for bioassay,
air-monitoring, and air-sampling equipment to ensure that they  are  up-to-date;
the RPO's trend analysis for indication  of increased activity;  all  operations
to ensure that there are procedures for  each; and  the  RPO's  records of surveys
versus maximum permissible concentration-hours (MPC-hrs)  to  ensure  that the
RPO has a method for interpreting whole-body-counting  data that relate to  the
working environment.
     Individuals working with radioactive  materials  may work with  unencapsu-
lated sources in physical forms or in chemical solutions.   When these materials
are unintentionally released from their  containers,  they can be inhaled,
ingested, or absorbed through the skin.   Therefore,  radiation  protection pro-
grams for such workers should include 1) methods to  limit internal  exposures,
2) an internal-dosimetry program, 3)  reviews of exposures, their causes, and
the corrective actions taken, and 4)  a quality assurance program.   All of
these aspects of the program for internal-exposure  control should  be reviewed
during the technical appraisal.
     A.  Exposure Limitation Methods.  Two important methods of limiting in-
ternal exposures are administrative and  engineered  safeguards.
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                                                             DARCOM-P  385-1
      (1)  Administrative Safeguards.  The administrative approach to control-
 ling  internal exposures usually consists of site-assigned dose  limits and
 written procedures.  The limits should be considered in the establishment of
 procedural and physical controls.   The written procedures should be well
 disseminated and read by all radiation workers and support personnel who  may
 have  reason to go into an area of potential airborne radioactivity.  The
 appraisers should ensure that the procedures define clearly when protective
 clothing and equipment are needed and include a means of ensuring that only
 qualified personnel use respirators.
      The procedures should define the requirements for posting  controlled-
 access areas and areas where airborne or other contamination is known to
 exist.  The appraisers should ensure that suitable measures are taken to
 minimize leakage, control local releases, and clean up contaminated areas, and
 that  there are adequate plans for expanding the respiratory protection program
 in the event of an accident.
      (2)  Engineered Safeguards.  Engineered safeguards against internal
 exposure are provided by containment and ventilation systems, contamination
 control, alarm systems, and respirators.  The reviewers appraising the control
 of internal  exposure should begin by reviewing the first three  of these items
 to ensure that every effort has been made to minimize the number and sire of
 areas containing airborne radioactivity.
     Respirators are the primary physical device for minimizing internal  expo-
 sures.  The  use of respirators in either an NRC-approved or a nonapproved pro-
 gram to reduce the potential for inhalation of radioactive material constitutes
 a respiratory protection program.   The commitment to a quality respiratory
 protection program should begin with a written policy statement on respirator
 usage issued by a high management level  (e.g., by the installation commander).
This  policy  statement should discuss the objectives of the program and assign
the responsibility for its operation to the RPO.
     The issuance,  maintenance, and repair of respirators, and the training of
personnel  for their use, should meet the guidelines found in such documents as
the Nuclear  Regulatory Commission's NUREG-0041 (NRC 1976).  Before beginning
the appraisal  of the respiratory protection program, the appraisers should
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DARCOM-P 385-1
 ensure  that they are intimately familiar with these support documents.   The
 degree  to which NUREG-0041  is applied at a site will  depend on the peculiari-
 ties of the individual program.  However, the appraisers should ensure  that
 every program  includes at least the following items:   medical  examination of
 each respirator user by  a qualified physician, including pulmonary measure-
 ments;  training in proper respirator use; use of only those respirators
 approved by the National  Institute of Occupational Safety and Health (NIOSH);
 an  inspection  program to ensure that breathing air meets the requirements of
 ANSI Z88.2  (1980)'and the Occupational Safety and1 Health Administration
 (OSHA); and a  program for cleaning and maintaining respirators for both
 hygienic and contamination  control purposes.
     The individuals responsible for training radiation workers in the proper
 use of  respirators should have received their training directly from a certi-
 fied respirator manufacturer.  Their training should have included proper
 fitting of masks and repair procedures.
     B.  Dosimetry Program.  An internal-dosimetry program consists of mea-
 surements of the concentration of airborne radioactive materials  in the work-
 place;  bioassay measurements, for estimating the quantity of radioactive
 materials deposited in various body organs; measurements for determining
 ionizing-radiation doses to body organs; and techniques for assessing these
 measurements.
     The appraisers should  determine whether the bioassay techniques and count-
 ing facilities used at an Installation are sufficient to permit a  reasonable
 assessment of  the internal  burdens of the radionuclides used at that installa-
 tion,   The bioassay techniques should include the use of models or calibra-
 tions to ensure the accuracy and reprodudbiHty of measured findings.  The
 operating manual at each site should state, for each technique used, the type
 of  radiation detectable  by  the technique, the sensitivity and  accuracy of  the
 system, the calibration  sources used and the activity and intensity of each,
 and whether the system is sensitive enough to detect a  concentration equiva-
 lent to 5% of  the maximum permissible body burden  (MPBB) for the  most  restric-
 tive radionuclide in a mixture of radionuclides.  This  determination must  be
 within  a 95% confidence  level.
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                                                             DAKCO.M-P 385-1
     Appraising an internal-dosimetry program is  a  detailed  and  complex  pro-
cess that requires the knowledge of an expert with  many years of practice!
experience.  The appraiser needs to use reference documents  such as  ANSI
N343-1978 and Publication 2 (1959) of the  International Commission  on
Radiological Protection (ICRP).
     C.  Exposure Review.  The appraisers  should  determine whether  radiation
exposure and/or dose limits for routine operations  and  nonroutine events  are
maintained ALARA and whether survey and internal-exposure data  on individuals
are routinely compared with each other and with  the limits.   When the  limits
are exceeded or closely approached, the appraisers  should determine  whether
the RPO takes corrective action and how effective that  action is.  To  support
the RPO, managers, supervisors,  and foremen of operations'and support  groups
should strive to keep both individual and  group  exposures,  and  the  number of
workers exposed, at a minimum.  The existence of SOPs  that  require  the signa-
ture of the RPO or a designee, the radiation worker, and  the worker's  super-
visor is a good indicator of such an effort.
     D.  Quality Assurance Program for Internal  Dosimetry.   The primary pur-
pose of the quality assurance program is to ensure  that the  data gathered in
the internal-dosimetry program represent the best efforts  possible  in  dose
assessment.  To this end, the RPO should establish  calibration frequencies
for, and the quality assurance staff should review, each  dosimetry  system and
dose assessment technique,  The appraisers should ascertain  whether the RPO
periodically evaluates the quality assurance and calibration reports to
determine whether the established calibration frequency is  adequate for each
system used.
     The appraisers should ensure that whole-body counting  equipment is cali-
brated using sources traceable to the NBS.  These sources should cover the
entire spectrum of radionuclides currently in use at the installation and
should vary in strength from the lower limit of  detection of the counting
system to realistic accident levels.  The  appraisers should also determine
whether the whole-body counting system is  calibrated at least annually.  The
routine calibration program should include an interim calibration with toler-
ance limits that, if exceeded, require recalibration of the entire system.
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DARCOM-P 385-1

Guidance on the calibration of.whole-body counting systems can be found in
documents such as ANSI N343-1978.
     The appraisers should determine whether the installation has a procedure
for estimating MPC-hr exposures from whole-body-counting data.  Because 10 CFR
20.103 expresses standards for internal emitters in terms of time-integrated
concentrations (MPC-hrs) and intakes rather than permissible body burdens or
doses, it is  important that the RPO 1) maintain a comprehensive breathing-zone
air-sampling  program, and 2) be able to compare whole-body or organ burden
data with the data generated by the air-sampling program.  To accomplish this,
the RPO must  have a method for interpreting whole-body-counting data in terms
of the MPC-hrs of exposure needed to produce the measured burden.  One of the
main reasons  for relating the data base on whole-body counting to the data
base on air sampling is to determine the effectiveness of the respiratory
protection program.
15.2.5  The Proaram for External-Exoosure Control
     Management  review of the program for controlling external exposure, like
 that of the  internal-exposure program, involves observation of work practices
 and review of  records.  Management can perform an informal, walk-through
 review by being  aware of whether radiation areas are posted properly, dosim-
 eters are worn properly where they are required, radioactive waste is stored
 properly, and waste containers are labeled.  Managers who are not familiar
 with proper  procedures in these areas can consult AR 40-14, AR 385-11, and the
 RPO.  In addition to the walk-through reviews, management should ensure that
 dosimetry records are maintained for everyone issued a dosimeter; that super-
 visors use dosimeter results when planning jobs and staff assignments; that
 all personnel  are given the results of their annual dosimeter reading; that
 the RPO knows  the procedure for reporting an overexposure; and that a suffi-
 cient number of  dosimeters are available for routine and emergency use as well
 as for visitors.
     The technical appraisal of the external-exposure control program should
 include review of 1) the methods used to limit exposures, 2)  the dosimetry
 program, 3)  the  reasons for exposures and any corrective actions taken, and
 4) the quality assurance program.
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                                                              DARCOM-P  365-1
     A.  Exposure Limitation Ketnods.  Both aoministretive and engineered
sefeguaros should De reviewed as part of the appraise!.   The aaministrative
means of control" ing external exposures usually consist  of site-assigned dose
limits and written procedures for minimizing exposures.   The appraisers should
review the dose  limits to Determine their usefulness end  the ability of the
staff t: meet them.  They should also talk with randomly  selected radiation
workers to verify their awareness of the administrative  guidelines.  These
guidelines should clearly reflect existing regulations  and recognize the ALARA
concept.
     The use of  physical barriers for exposure control  should be reviewed by
the RPO on a regular basis and the results should be documented.  The
appraisers should evaluate the use of barriers and talk  with radiation workers
to determine their effectiveness.  If remote-operating  and remote-handling
devices are available, the appraisers should ensure that  the individuals
authorized to use them have received special training and that the devices are
well maintained.  In areas with access alarms, periodic  tests of the alarms
should be performed to ensure their operation, and placards showing the
potential hazards of the areas should be clearly displayed.
     B.  Dos.inetry Program.  Before beginning this phase of the appraisal, the
appraisers should assure themselves that they are familiar with the current
standards in the area of external dosimetry, including ANSI N13.11-1980.
     The appraisers should review ell sources licensed ft>r use at  the  instal-
lation tc ensure that the dosimetry program is suitable for the types  and
levels of radiation exposure anticipated during routine end r,onroutine work.
They should also evaluate the personnel responsible for the dosimetry  function
tc net.errr.ine whether they neve adequate knowledge to perform  routine duties
n ;ii to recognize unusual events that may require special  interpretations or
evaluations.  Appraisals frequently reveal that the readings  from  film or
th?rmoluminescence dosimeters ere not compared with the readings from
secondary dosimeters (e.g., pocket ionizetion chambers).   When comparisons ere
m=ae,  an acceptance criterion, or level at which follow-up action  is required,
should be specified.
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DARCOM-F 3S<- i
     An installation that sends dosimeters .offsite  for  calibration  ana/or pro-
cessing should have a quality assurance program.  This  program  should include
the use of spiked dosimeters and blanks.   Secondary dosimeters  should be care-
fully screened before they are put into service.  The acceptance  of offsite
work without an independent quality control  check represents  a  failure by the
RPO to take responsibility for the accuracy  of  the  dosimetry  program.
     The appraisers should ensure that the equipment and  facilities available
are adequate for nonroutine dosimetry and exposure  control.   Enough dosimeters
of acceptable quality and sensitivity should be available for short-term use
by personnel or visitors to areas requiring  dosimeters.  Exposure records
should be kept current and should be sent to workers and  their supervisors
frequently and promptly enough to ensure their  usefulness.
     C.  Exposure Review.  The appraisers should determine whetner the expo-
sure data generated by dosimeters and instruments  are  routinely reviewed by
management and whether any discrepancies between the primary and secondary
dosimeter readings that exceed the acceptance criterion are followed up by an
investigation of the exposure conditions.
     The RPO should maintain a plot of exposures that shows trends and  indi-
cates whether doses are being kept ALARA.  These plots can be cross-referenced
by job, location, profession, and total work force.  The RPO should  also have
records of each review of the trend plot and the results of that review.
     D.  Quality Assurance Program for External Dosimetry.  The quality  assur-
ance organization should play an active role in the program for controlling
external exposures.  The appraisers should assess  the quality assurance  func-
tions performed by the RPO and determine whether the quality assurance  repre-
sentative assists in reviewing procedures and ensures  that there is  suitable
feedback from management.  For onsite calibration  of instruments,  devices,' and
processes, the quality assurance representative should assist  in establishing
acceptance criteria.  The appraisers of the quality assurance  program should
pay careful attention to the records maintained by the quality assurance office.
15.1.6  The ALARA Program
     The appraisers  should verify that management  has  a  written  policy showing
commitment to ALARA  and administrative procedures  to implement the policy.

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                                                             DARCOM-P 385-1
The RPO should be responsiole for overseeing  the  ALARA program as described in
NRC Regulatory Guide 8.10 (1975).  However, an  individual  in management  should
be responsible for working with the RPO to ensure that mechanisms for keeping
exposures ALARA are instituted at the site.
     The appraisers should determine whether  an adequate  system  has  been estab-
lished to avoid unnecessary or inadvertent personnel  exposures.   Shielding
should be used when equipment is being serviced;  measures  should  be  taken to
provide distance from sources, when possible; and easy access to  equipment
should be provided.  The appraisers should determine  whether remote-handling
tools or remote readouts are used when necessary.  They should also  thoroughly
review the entire radiation protection program  to determine the  effectiveness
of the ALARA program in reducing exposures.
     The appraisers should interview radiation  workers to determine  their con-
cept of ALARA, whether adequate training, preparation, and planning  are
incorporated into work activities, whether the  radiation  protection  staff
become involved early in the planning of work,  and whether a debriefing  is
held when a job is completed to determine more  effective  means of reducing
exposures.
     Management review of this area should be limited to  ensuring that there
is an ALARA program review committee, that exposure information  is  used  for
job planning, and that personnel are familiar with the ALARA  principle.
15,2.7  Facilities and Equipment
     Management review of facilities and equipment should include a  walk-
through and visual inspections of equipment.  The walk-through  should center
around the availability of sufficient space for calibrations,  sample analysis,
and the use of laboratory counters.  Management should also  inspect the condi-
tion of support equipment (e.g., protective clothing).   The  technical appraisal
should cover the topics discussed below.
     A.  Facilities.  At each installation, the appraisers should evaluate
whether there are sufficient locations and space  for the.following:   counting
room,  calibration of instruments, personnel  decontamination,  access control,
offices, equipment decontamination, instrument  storage,  external dosimetry,
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internal  dosimetry, the fitting,  testing, and cleaning of respirators, train-
ing, contaminated-equipment storage,  and  laundry.  When a new facility is
being designed, the RPO should be involved  in an ALARA review of that
structure.
     If the installation uses large enough  quantities of radioactive material
so that the potential  for offsite releases  is a concern, the appraisers  should
ensure that the RPO has made provisions  for offsite decontamination of per-
sonnel -and has determined whether local  hospitals have sufficient  space  and
equipment to handle emergencies involving contaminated victims.
     B.  Protective Equipment.  Respirators,  protective clothing,  temporary
shielding, and containment materials  should all be  reviewed  as  part of the
equipment appraisal.
     (1)  Respirators.  The supply of respirators should be  adequate  for han-
dling routine and abnormal operations.   The installation should have  an  agree-
ment with a commercial company or another command for  the  rapid procurement  of
extra respirators and for the expansion  of  decontamination  and  repair services
in the event of an emergency.
     (2)  Protective Clothing.  Protective  clothing should  be  stored  in  a
number of locations so that all of it 1s not lost  in  the event  of a  fire or
accident.  The supply should be adequate for handling  routine  and nonroutine
operations.  For accident situations, special  clothing such as  disposable
paper and plastic suits should be available.   Contamination limits for reus-
able clothing should be established.   When  the level  of contamination on the
clothing exceeds the limit, the clothing should be  disposed of.
     (3)  Temporary Shielding.  The appraisers should ensure that an adequate
supply of temporary lead shielding, such as bricks, blankets,  lead shot, and
lead sheets, is available.  The radiation protection staff should be trained
in the proper use of these supplies and instructed to carefully survey the
temporary shielding before removing it to avoid spreading contamination.
      (4)  Containment Materials.  The supply of containment materials (e.g.,
heavy-gauge plastic sheeting, plastic windows, and nonskid floor  covering)
should be adequate for handling routine and nonroutine operations.  The  RPO
should carefully analyze these materials for compatibility with the work

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                                                             UARCO.'.-P 385-1
environment they will  be used in.   The  site  should  have  detailed  procedures  on
trie use of these materials, and the radiation  protection  staff  should  oe
trained in their use.
15.2.8.  Management of Radioactive  Waste
     Appraisal of the waste management  program should  include  a review of
records and a wall'-through inspection of  all  areas  where  waste  is either
generated or stored.   All waste should  be stored in authorized, appropriately
labeled containers free of exterior contamination,  rust,  and  corrosion.   The
appraisers should ensure that radioactive waste is  collected  separately from
nonradioactive waste  and that it is promptly removed from the  generator's loca-
tion and stored in properly posted  areas  apart from work locations.   Control
procedures should be  used to minimize personnel exposures.
     The appraisers should inspect  waste  that is ready for transport to
determine whether it  has been packaged  and labeled  accoording  to Department
of Transportation (DOT) and Department  of the Army  (DA)  regulations.  The
appraisers should verify the availability of suitable packaging material, as
well as packaging procedures.  Trucks holding waste for transport should be
inspected to determine that they are in compliance  with DOT and DA regulations.
     Waste records should be reviewed to  ensure that an inventory of all  waste
generated and disposed of is maintained.   The total quantity of radioactive
material disposed of  into the sanitary  sewage system, the air, and nearby
streams as a result of all activities at  an installation must not exceed the
quantity for a single licensee given in 10 CFR 20.   Records for the transport
of waste should be reviewed to determine  whether they meet DOT and DA
regulations.
15.2.9  Records and Audits
     The records management system  should be reviewed to determine whether
records of each component of the radiation protection program are maintained.
In addition, the appraisers should  review the procedures for records disposi-
tion, traceability, retrievability, and physical protection.   Audit records
should be specifically reviewed to  determine whether the program  is periodi-
cally audited by individuals with the -appropriate technical expertise and to
verify that all audit findings have been  corrected promptly.

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DARCOM-P 385-1
       Section 15.3  CHECKLIST OF QUESTIONS FOR APPRAISING A RADIATION
                     PROTECTION PROGRAM

     A checklist of questions should be developed for use during appraisals.
The example checklist presented in this section is not comprehensive, but is
intended to provide an overview of the areas of interest in an appraisal,
based on the discussion in the preceding section.
15.3.1  The Radiation Protection Organization
  1.  Is there an organizational chart depicting the installation's interrela-
     tionship with the radiation protection organization?
  2.  Does the base commander have a working relationship with the RPO?
  3.  Does the organizational chart show that'the RPO has a direct reporting
     chain to the base commander?
  4.  Does the RPO's manager exhibit a clear understanding of the goals of the
     radiation protection organization?
  5.  Is there evidence of strong management qommitment to radiation protection
     (e.g., written policies or administrative procedures)?
  6.  Is there a clear assignment of authority and responsibility within the
     radiation protection organization?
  7. 'Does the radiation protection staff have adequate authority to ensure
     that the radiation protection program is implemented?
  8.  If classified work is being done, does the RPO have adequate clearance
     and unfettered access to ensure that the work is being conducted safely?
  9.  Is there sufficient staffing within the radiation protection organization
     to provide adequate coverage of all work with radiation?
10.  Is the RPO included in the design phase of operations  involving  radioac-
     tive material?
11.  Is the RPO or a designee required to authorize all  SOPs?
12.  Does there appear to be open communication between  the RPO  and  both
     radiation workers and other staff members?
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                                                             DARCO.M-P 385-1
13.   Is there adequate  administrative  support  to  relieve technical  personnel
     frorr: clerical  duties?
14.   Has a management  level  individual  (e.g.,  the  RPO or a higher-level  per-
     son) been designated  the  responsibility for  emergency preparedness?
15.   Are written emergency  plans  and procedures available that  are  commensu-
     rate with the  degree  of hazard?
16.   Are there established  procedures  for  obtaining  offsite support?
15.3.2  The Selection  and  Training  of  Personnel
 1.   Is there a radiation  safety  training  program for staff members
     commensurate with  their responsibilities?
 2.   Is there a training program  for the radiation protection staff?
 3.   Does training  for  the  general  staff members  and the radiation  protection
     staff include  the  following?
      a.  a defined scope  and  written  content  for the program
      b.  instructors qualified in  the subjects they are teaching
      c.  instruction  schedules and lesson plans
      d.  objectives for trainee  performance
      e.  demonstration of  standards attained  by  trainees
      f.  frequency of  required attendance
      g.  documentation of  attendance  (including  test results,  dates, and
          subjects).
 4.   Is the scope of the training adequate.in  content,  nature,  and  length?
 5.   Do training programs  include hands-on use of equipment end tours of areas
     that the trainee may  need to enter in the course of work?
 6.   Are the operators  of  the  various  counting and analytical systems properly
     and adequately trained  in the  use of  the  systems?
 7.   Is formal  on-the-job  training  available at appropriate  intervals for all
     individuals?
 8.   Are requalification and retraining in the state of the art of  instrumenta-
     tion available for personnel?
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DARCOM-P 385-1
 9.  Is there a documented training program covering emergency preparedness?
10.  Are operators of computer-based analysis systems capable of manual calcu-
     lations  in the event of a power loss?
11.  Are training records maintained for a minimum of 5 years?
12.  Are the  training records complete enough so that the quality, duration,
     location, and content of training can be ascertained?
15.3.3  The Radiation Survey Program
 1.  Is there a clear definition of and basis for the survey program?
 2.  Are procedures for performing routine and periodic surveys well defined?
 3.  Does each survey record contain as a minimum the following?
     a.   survey purpose
     b.   survey frequency and locafion
     c.   survey technique
     d.   instrument selection, calibration, and use
     e.   data and records disposition
     f.   status of follow-up actions.
 4.  Do procedures or policy statements delegate to the radiation  protection
     organization the responsibility for reviewing all SOPs?
 5.  Are the  data from routine and periodic surveys reviewed by the  RPO for
     technical content and possible additional action?
 6.  Are all  surveys well documented?
 7.  Do SOPs  correctly reflect job and work conditions?
 8.  Is there timely and adequate feedback of analytical  results to  staff
     personnel?
 9.  Is the recordkeeping system commensurate with the guidelines  outlined  in
     Chapter  13 of this manual and those in ANSI N13.6-1966?
10.  Are radiation areas properly posted in accordance with  10  CFR 20.203?
11.  Are portable instruments of sufficient number, type,  range, and sensi-
     tivity available for routine and nonroutine activities?
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                                                             DARCOM-P 385-1
12.   Do  instruments  have  a  calibration  sticker that specifies the date the
     instrument  should  be recalibrated,  the name or initials of the  person
     that performed  the calibration,  the  actual calibration date, the source
     used for calibration,  and  the  location of the calibration facilities?
13.   Are air-samoling  instruments sufficient  in number, sampling range,  and
     type for the  scope of  routine  and  nonroutine activities?
14.   Are there procedures that  specify  the calibration frequency for all
     instruments?
15.   Are calibration sources  traceable  to NBS?
16.   Are inoperative instruments  properly marked, stored, and repaired?
17.   Are instruments dedicated  to sample  analysis properly maintained?
18.   Are instrument  dials and scales  clearly  legible?
19.   Are survey  results plotted and reviewed  for possible trends?
15.3.4  The Program  for Internal-Exposure Control
 1.   Is  there a  bioassay  program commensurate with the level  of  hazard  at the
     installation?
 2.   Are baseline  whole-body  counts or  urinanlyses performed  on  personnel
     before they begin  work with radioactive  material?
 3.   Are the bioassay  techniques  used at the  site based on the  radionuclides
     used there?
 4.   Are the sensitivities  of the bioassay procedures  adequate  for assessing
     maximum permissible  body burdens and maximum permissible concentrations?
 5.   Is  there a  written procedure for correlating air-sampling  results  and
     bioassay results?
 6.   Are internal-dose  limits for routine operations  and  nonroutine events
     maintained  ALARA?
 7.   Are incidents of  personnel  contamination documented,  and are  the causes
     investigated?
                                     15.29

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     -P 385-1
 8.   Are  adequate records  maintained  on  all  individuals who have  received an
     internal  deposition of radioactive  material?
 9.   Are  uptake limits  considered  in  the establishment of administrative and
     engineered safeguards?
10.   Are  there procedures  that aid in determining  the need for  protective
     clothing  and equipment?
11.   Are  there well-defined procedures for  posting controlled-access  areas  and
     areas where 'airborne  or other contamination  is known to  exist?
12.   Are  proper measures taken to  minimize  leakage, control local  releases,
     and  clean up contaminated areas?
13.   Are  there adequate procedures for preventing  or controlling  cross-
     contamination of samples?
14.   Are  air flows from areas of low  to areas  of  high airborne  radioactivity?
15.   Has  management issued a written  policy statement on  the  use  of
     respirators?
16.   Are  there methods  of ensuring that only qualified  personnel  use
     respirators?
17.   Does the  person responsible for  the respiratory protection program have
     the  ability, training, and experience  to do  the following?
     a.    evaluate total hazard
     b.    recommend engineering controls
     c.    specify appropriate respiratory protection factors  and equipment
     d.    forbid use of equipment when conditions warrant.
18.   Are  sufficient records maintained to evaluate the  effectiveness  of the
     respiratory protection program?
19.   Do  the issuance, maintenance, and repair of  respirators, and the training
     of  personnel for their use, meet the guidelines found in such documents
     as  NUREG-0041?
20.   Do  all personnel who wear respirators  have documentation of a complete
     bronchio-pulmonary examination?
                                     15.30

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                                                             DARCO.v-P 385-1
21.  Are all respirators used at the  installation of the type approved by
     NIOSH?
22.  Are there provisions to ensure the  proper  fit of  respirators?
23.  Are medical personnel given enough  guidance to adequately  evaluate  the
     ability of wearers to use the equipment?
24.  Are respirators fitted, inspected,  tested, and repaired, and  are  tne
     wearers trained, in accordance with NUREG-0041 or its  equivalent?
15.3.5  The Prooram for External-Exposure Control
 1.  Is the dosimetry program suitable for the  types  and  levels  of radiation
     exposure anticipated during routine and  nonroutine operations?
 2.  Are there suitable devices, exposure models,  and data  bases for measuring
     or calculating extremity exposures?
 3.  Can skin exposures be determined by modeling  or  measurement?
 4.  Are there suitable techniques, devices,  or instruments for  measuring
     neutron exposures?
 5.  Are dosimeters of acceptable quality and sensitivity available for
     short-term use by personnel or visitors  to areas requiring  dosimeters?
 6.  Are dosimeters being worn in the proper  position on  the body and/or
     extremities?
 7.  Are exposure records on ell personnel wearing dosimeters kept
     up-to-date?
 8   Are exposure data reviewed routinely by  management,  end ere the reviews
     documented?
 9.  Are discrepancies between the readings of primary and secondary
     dosimeters reviewed by management (RPO or higher levels)?
10.  Are exposure results and exposure histories evaluated against the ALARA
     requirements of AR 40-14 and 10 CFR 20 as  part of a  routine review?
11.  Do administrative procedures clearly establish action levels and required
     actions in the event of an exposure that exceeds administrative limits?
                                     15.31

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DARCOM-r 385-1

 12.   Do  procedures  clearly  reference and  reflect existing regulations and
      recognize  and  incorporate  the ALARA  concept?
 13.   Are there  written  procedures for  the posting of various hazardous or
      potentially  hazardous  areas  in accordance with 10 CFR 20?
 14.   Is  the  RPO thoroughly  familiar with  the location of all radioactive mate-
      rial  used  at the  installation?
 15.   Does  the dosimetry program include the use of dosimeters spiked with
                  t
      known types  and quantities of radiation, to provide a quality  assurance
      check during processing?
 16.   Are dosimeters stored  in a controlled location to reduce adverse environ-
      mental  effects?
 17.   Are control  dosimeters included in all shipments to the dosimeter
      processor?
 18.   Is  the  RPO responsible for the control, issuance, and evaluation of all
      dosimeters?
 19.   Are there  routine  quality  assurance  reviews of the  dosimetry program?
 20.   Is  quality assurance extended to  the review of procedures?
 15.3.6   The  ALARA Program
  1.   Is  there a written management policy showing  commitment to  ALARA?
  2.   Are there  written  administrative  procedures to implement  the ALARA
      policy?
  3.   Do  facility  and equipment  design  features  incorporate  ALARA concerns?
  4.   Is  work adequately prepared and planned for?
  5.   Is  the  radiation  protection staff involved  in the  planning  of work?
  6.   Are formal or  informal postoperational briefings  held?
  7.   Are engineered safeguards  used to keep exposures  ALARA?
  8.   Is  surface contamination controlled  adequately?
  9.   Are remote readouts available?
 10.   Are unnecessary exposures  during  routine  surveys  minimized?

                                      15.32

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                                                             DARCOM-P 385-i
15.3.7  Facilities  and  Equipment
 1.  Are there sufficient locations  and  space for the following:   sample
     counting, calibrations,  personnel and equipment decontamination, access
     control, offices,  instrument  storage, external and  internal dosimetry,
     fitting, testing,  and cleaning  of respirators, training, contaminated-
     equipment storage,  and laundry?
 2.  If a new facility  has been  designed, was an ALARA review of the  structure
     performed?
 3.  Are adequate supplies of protective clothing,  respirators, temporary
     shielding, and containment  materials available, and  is the radiation
     protection staff trained in their use?
 4.  Are all  radiation  areas  posted  and  isolated from controlled areas?
 5.  Is access to radiation areas  controlled?
 6.  Are sinks, drain lines,  and water supplied to  a radiation  area  isolated
     from the sanitary  sewer?
 7.  Are radiation  areas ventilated  to prevent the  flow  of  air  into  uncon-
     trolled  areas?
 8.  Is emergency equipment available  (e.g., fire extinguishers, safety"
     showers, telephones)?
 9.  If a potential  for  offsite  releases exists, have provisions been made  for
     offsite  decontamination  of  personnel, and do local  hospitals  have  suffi-
     cient space to handle emergencies involving contaminated  patients?
15.3.8  Management  of Radioactive  Waste
 1.  Is the use of  radioactive material  planned so  that  a minimum  of radioac-
     tive waste is  generated?
 2.  Is radioactive waste separated  from nonradioactive  waste?
 3.  Is waste segregrated by  physical form, half-life, and  type of nuclide?
 4.  Are containers  used for  temporary storage properly  labeled,  strong,
     leaktight, end free of exterior contamination, rust, and  corrosion?
                                     15.33

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JAXCOM-F  385-1
 5.  Is radioactive waste stored away from the work area?
 6.  Are appropriate control procedures used to minimize personnel  exposure?
 7.  Are all waste generation and storage areas monitored to ensure contamina-
     tion control?
 8.  Is waste for transport packaged and labeled according to DOT and DA
     regulations?
 9.  Is the total quantity of radioactive material disposed of into the
     sanitary sewage systems, the air, and nearby streams as a result of all
     activities at the installation less than the quantity for a single
     licensee given in 10 CFR 20?
15.3.9  Records and Audits
 1.  Are records maintained for each component of the radiation protection
     program?
 2.  Does the records management system include the icentification of specific
     records, the disposition of records (review, storage, retention period),
     traceability to the originator, retrievability for audits or investiga-
     tions, provisions for periodic audits, and physical protection for legal
     records?
 3.  Are there complete and up-to-date personnel  files for all radiation
     workers?
 4.  Is DD Form 1141 (or the automated dosimetry  records) filed ir. each
     individual's personnel file?
 5.  Are records maintained in accordance with the guidance  in Chapter  13 of
     this manual?
 6.  Are records maintained in accordance with the guidance  in 10 CFR  19 and
     10 CFR 20?
 7.  Is the radiation protection program audited  periodically?
 8.  Does the quality assurance staff conduct performance audits?
 9.  Are previous audit reports reviewed before new audits  are conducted?
                                     15.34

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                                                             DARCOK-P 385-1
10.  Are audit findings corrected within a  reasonable  time?
11.  Are technical audits performed by individuals  with  extensive experience
     in the areas in question?
           Section 15.4  NETWORK TECHNIQUES FOR PLANNING APPRAISALS

     The manager or command group that is looking for an appraisal  technique
to aid in planning a radiation protection appraisal  is faced with a bewilder-
ing family of network methods.  A network is an organized way of thinking
about complex problems by using common sense to determine a  sequence of
logical steps.  Managers (such as RPOs) today face a great increase in the
complexity of their work; because they are often dealing with the future
(limiting future exposure, planning future facilities), they also face uncer-
tainty.  Network techniques were designed specifically to deal with the
factors of complexity and uncertainty.
15.4.1  The Function of Networks andLogic Trees
     The first network method for controlling projects, PERT (Program Evalua-
tion and Review Technique), was developed for use on the Polaris Submarine
program by the U.S. Navy in 1958.  The second, more successful method was
developed by the DuPont Company and is called the Critical Path Method (CPM).
A critical path is defined as a sequence of elements of e program that are
dependent upon one another.  For example, the radiological survey of a waste
container is dependent upon the training of the staff members and the proper
rev-onse of the survey instruments.  In turn, the proper response of a survey
instrument is dependent upon its calibration, physical condition (whether it
is damaged), and power source (strength of batteries).  Therefore, an adequate
survey is dependent on several critical paths.  The critical paths for this
particular example are shown in Figure 15.1 by the use of arrows.  Although
the PERT and CPM networks cannot be used directly in planning appraisals, they
demonstrate the value of logically displaying the relationships airing trie
basic elements of a program, and thus they led to the development of more
useful  methods such as MORT (Management Oversight and Risk Tree Analytical
Logic Methodology).
                                     15.35

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i)ARCOM-F .,85-1
              FIGURE  15.1.   Critical Paths for an Adequate Survey

      The MORT system is a  logic tree network that was designed for use in
 investigating the causes of accidents, or undesirable conditions.  However,
 this  system and others like it can be modified to graphically depict a single
 desirable condition, the starting point for the tree, and systematically pro-
 ceed  through  lower levels  or  tiers until all important factors that produce
 the desirable condition have  been identified.  This concept is shown in Fig-
 ure 15.2, where a tier may be dependent upon several critical paths, as shown
 by an  "AND" gate, or may be dependent upon only one critical path, as shown by
 an "OR" gate.  An example  of  a logic tree structure is shown in  Figure 15.3,
 where  the desirable condition is for a process to be operationally ready.
      An appraisal program  developed using logic trees would be broken down
 into  many branches, each specific to a single desirable condition or set of
 related conditions.  Each  branch would have some point of interface with at
 least  one other branch or  tree.  The  -nterfaces between branches or trees are
 important in  the evaluation process:  data collected from the appraisal of one
 area must be  transferred to another area and considered in the evaluation of
 bct^.  Through th;s process,  the impact of a particular finding  can be assessed
 in a  systematic way, with  a minimum expenditure of time and effort.  The
 examples of logic trees presented in this section are a combination of several
                                     15.36

-------
                                                              DARCOM-P  385-1
                         BASIC   BASIC   BASIC  
-------
DARCOM-P 385-1
15.4.2  Using Logic Trees  to  Plan a Radiation Protection Appraisal
     The first step in developing an appraisal program is to establish  the
objectives of the radiation protection program (e.g., to keep exposures  as  low
as is reasonably achievable (AlARA) and to minimize the potential  for acci-
dental exposures).  On the diagram of the logic tree, this objective is  placed
in a box that becomes the  goal of the total logic tree (see Figure 15.4).
Accomplishing such an objective  requires both a routine operation  and an
emergency operation.  This dual  requirement is shown by the "AND"  gate  beneath
the top box  in Figure 15.4.   Figure 15.5 shows the further subdivision  of  an
effective routine program  into its major components.  A deficiency in any  of
these components could cause  the entire routine program to be inadequate.
Therefore, an "AND" gate is used to show the  relationship of the  routine
program to its major components.  The emergency operation, however,  can be
satisfied by either a modified routine operation or a special emergency
program.  Therefore, the emergency operation  diagrammed in Figure  15.6  has  an
"OR" gate to show its relationship to its components.  The combination  of
Figures 15.4, 15.5, and  15.6  yields the logic tree structure for  the first two
tiers of the radiation protection program, as shown in Figure  15.7.
     In a complete appraisal  program developed using  logic trees,  each  of  the
components of a routine  program  would be further subdivided from  two to eight
times.  The  subdivisions of the  component "Internal-Exposure Controls"  are
shown in Figure 15.8.  The degree of subdivision into lower tiers  depends  on
the complexity of the radiation  protection program.   A recent  appraisal of the
radiation protection programs a- operating power reactors  involved the  use of
                                KEEP EXPOSURES ALARA;
                                MINIMIZE POTENTIAL FOR
                                ACCIDENTAL EXPOSURES
1
ROUTINE
OPERATIONS

1
EMERGENCY
OPERATIONS
            FIGURE  15.4.   Radiation  Protection  Program,  First Tier
                                      15.38

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                                                           DARCOX-P 3E5-1



RADIATION
PROTECTION
ORGANIZATION




PERSONNEL
SELECTION
AND TRAINING





EXPOSURE
CONTROLS







ALAR A
PROGRAM




FACILITIES AND
EQUIPMENT




RADIOACTIVE-
WASTE
MANAGEMENT
1
RECORDS AND 1
AUDITS 1
1

1
SURVEILLANCE
PROGRAM







INTERNAL
EXPOSURE
CONTROLS



EXTERNAL
EXPOSURE
CONTROLS
 FIGURE 15.5.   Radiation  Protection Program, Second Tier—Routine  Operations
FIGURE 15.6.   Radiation Protection Program,  Second Tier—Emergency  Operations
                                   15.39

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DARCOM-P 385-1

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DARCOM-P 385-1
18 trees,  two of which interfaced with  each  of  the  remaining  16.   The  inter-
faces are  usually designated by transfer functions  (triangles with arrows  and
a letter or number) thai indicate when  data  from one  area  should  be used  in
the evaluation of another.
     The analytical trees should be designed so that  they  graphically  depict
the total  radiation protection program.   To  help the  appraisers  prop'-ly
evaluate each area included in the trees, a  checklist of questions, such  as
those in the previous section, is designed to accompany each  element in every
tier.  The questions define the scope of the appraisal and ensure considera-
tion of the essential elements of a radiation protection program.   They are
not intended to be an all-inclusive listing  of  the  significant  items for
appraisal, but should provide the appraisers with the foundation  upon  which  to
evaluate the program.  The appraisers should find that the answers to  some
questions  lead them to a series of other questions  that are not  written  in the
appraisal  guide.
     The complexity of the appraisal process requires that the  appraisers be
familiar with a large number of regulations, regulatory guides,  and industry
standards.  These documents will be useful in judging the  adequacy of all or
part of a  specific area (e.g., dosimetry).  In  addition, the criteria used for
designing  the logic trees and for evaluating the program should  be taken  from
DA and NRC rules and regulations, ANSI  standards, National Council on Radia-
tion Protection and Measurements (NCRP) guides, and recommendations of the
ICRP and the International  Commission on Radiation  Units and Measurements
(ICRU).  However, the use of the logic tree system does  not eliminate the need
for professional judgment where standards and regulations  do not provide
sufficient detail; rather, its purpose is to help the appraisers clarify where
their judgment is needed.
                                     15.42

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                                                                   .M-P 3£5-i

                                  REFERENCES


American National Standards Institute (ANSI).   1966.   Practice for Occupa-
  tional Radiation Exposure Records Systems.  ANSI  N13.6,  New York.

American National Standards Institute (ANSI).   1978.   American National
  Standard for Internal Dosimetry for Mixed Fission and Active.ion Prod'ucts.
  ANSI N343, New York.

American National Standards Institute (ANSI).   1978.   Radiation Protection
  Instrumentation Test and Calibration.   ANSI  N323, New York.

American National Standards institute (ANSI).   1980.   American National Stan-
  dard Criteria for Testino :ersonnel Dosimeter Performance.  ANSI K15.il, New
  York.

American National Standards Institute (ANSI).   1980.   Practices for Respira-
  tory Protection.  ANSI 288.2, New York.

International Commission on Radiological Protection (ICRP).  1959.  Report of
  Committee II on Permissible Dose for Internal Radiation.   ICRP 2, Pergamon
  Press, Oxford.

U.S. Code of Federal Regulatir-ns.  1982.  Title 10, Part 19, "Notices,  Instruc-
  tions and Reports to Workers; Inspections."   U.S. Government Printing
  Office, Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 10, Part 20, "Standards for
  Protection Aaainst Radiation."  U.S. Government  Printing  Office, Washinoton,
  D.C.

U.S. Department of the Army, Headquarters.  Safety - Ionizing  Radiation Protec-
  tion (Licensing, Control, Transportation, Disposal, and Radietion SafetyT
  AR 385-11, Washington, D.C.

U.S. Department of the Army and Defense Logistics  Agency.   Medical Services  -
  Control and Recording Procedures for Exposure to Ionizing  Radiation and
  Radioact'ive Materla!s.  AR 40-14, DLAR 1000.28,  Washington,  D.C.

U.S. Nuclear Regulatory Commission (NRC).   1975.   "Operating Philosophy for
  Maintaining Occupational Radiation Exposure As Low As Reasonably Achiev-
  able."  Regulatory Guide 8.10, Washington, D.C.

U.S. Nuclear Regulatory Commission (NRC).   1976.   Manual of Respiratory Pro-
  tection Against Radioactive Materials.  NUREG-0041, Washington,  D.C.
                                     15.43

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                                                              DARCOM-P 385-1
                         CHAPTER 16.   REFERENCE DATA
16.1  LIST OF ELEMENTS   	



16.2  GREEK ALPHABET     	



16.3  ACRONYMS 	



16.4  ABBREVIATIONS AND SYMBOLS    ....



16.5  SELECTED CONVERSIONS    	



16.6  FREQUENTLY  USED CONSTANTS    ....



16.7  ADDRESSES FOR ORDERING REFERENCE DOCUMENTS  .



16.8  GLOSSARY 	
16.3



16.4



16.5



16.7



16.11



16.14



16.15



16.16
                                     16.1

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DARCOM-P  385-1
                          Section 16.1  LIST  OF ELEMENTS
    Atomic
    Nurri'er
Svmool
1
2
3
4
C
C
7
e
9
10
n
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
29
30
31
22
33
34
25
36
37
38
29
40
41
42
43
44
45
46
47
48
49
50
51
52
H
He
Li
Be
E
r
N
0
F
Ne
Na
Mo
Al
Si
P
S
Cl
Ar
K
Ca
Sc
Ti
V
Cr
Mn
Fe
Co
Ni
Cu
Zn
Ga
Ge
As
Se
Br
Kr
Rb
Sr
V
Zr
Nb
Mo
Tc
Ru
Rh
Pd
Ac
Cd
In
Sn
Sb
Te
hydrcjen
he! ium
1 i thium
beryl 1 ium
boron
carbon
nitrogen
oxygen
f luor ine
neon
sooium
magnes ium
aluminum
s il icon
phosphorus
sulfur
chlorine
argon
potassium
cal cium
scanoium
titanium
vanadium
chromium
manganese
iron
cobalt
nickel
copper
2 inc
gal 1 ium
germanium
arsenic
selenium
bromine
krypton
rub idium
strontium
yttrium
2 irconium
niob ium
molyoaenum
technetium
ruthenium
rhodium
pal ladium
silver
cadmium
indium
tin
.antimony
tel lur ium
Atomic
Njnoer
53
54
55
56
57
58
59
60
6]
62
63
64
65
66
67
68
69
70
71
72
73
74
75
76
77
78
79
80
81
62
83
84
85
86
67
88
89
90
91
92
93
94
95
96
97
98
99
100
101
102
103

Sy.TOol
1
Xe
Cs
Ba
La
Ce
Pr
Nd
Pm
Sm
Eu
Gd
Tb
Dy
HO
Er
Tm
Yb
Lu
Hf
Ta
W
Re
Os
Ir
Pt
Au
Hg
Tl
Pb
Bi
Po
At
Rn
Fr
Ra
Ac
Th
Pa
U
Up
Pu
Am
Cm
Bk
Cf
Es
Fm
K-
No
Lw

N'ane
iodine
xenon
ces ium
bar ium
lanthanum
cerium
praseodymium
neocyrr.ium
prometh ium
sama- ium
europium
gaool in ium
terbium
dysprosium
holmium
erbium
th u 1 i urn
ytterbium
lutetium
hafnium
tantalum
tungsten
rhenium
osmium
ir idium
platinum
gold
mercury
thai 1 ium
lead
bismuth
polonium
astatine
raoon
f r cnci urn
r 8 ; i urn
act in ium
tno-ium
protactinium
uranium
neptunium
plutonium
amer icium
curium
berk el ium
cal ifornium
einsteinium
' ermium
rer.oelev iur
nooel iun.
1 awrencium
                                      16.3

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DARCOM-P 3£5-1
                         Section 16.2  GREEK ALPHABET
Name
Alpha
Be -a
Gamma
Delta
Epsilon
Zeta
Eta
Theta
Iota
Kappa
Lambda
Mu
Nu
Xi
Omicron
Pi
Rho
Sigma
Tau
Upsilon
Phi
Chi
Psi
Omega
Upper Case
A
B
r
A
E
2
H
0
I
K
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M
N
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n
P
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T
Y
$
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c
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Y
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n
6
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X
u
V
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                                     16.4

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                                                              DAJRCOM-P 385-1

                            Section 16.3  ACRONYMS
AEC               U.S. Atomic Energy Commission
ALARA             es  low es  is reasonably achievable
ALI               annual limit of intake
AMC               Army Materiel Command
ANSI              American National Standards Institute

CAK               continuous air monitor
CFR               U.S. Code  of Federal Regulations
CP                Cutie Pie

DA                U.S. Department of the Army
DAC               derived air concentration
DCP               disaster control plan
DD                U.S. Department of Defense
DF                decontamination factor
DOF               dioctyl-phthalate
DOT               U.S. Department of Transportation
DSA               Defense Supply Agency
DU                depleted uranium

EAL               emergency  action level
ECC               emergency  control center
ED                emergency  director
EDTA              ethylene-diamine-tetra-acetic acid
EPA               U.S. Environmental Protection Agency
EPZ               emergency  planning zone

GM                Geiger-Mueller

HEPA              high-efficiency particulete air
HEW               U.S. Department of Health, Education, and Welfare
HKS               Health and Human Services
HQ                Headquarter
HVL               half-value layer

IAEA              International Atomic Energy Agency
IATA              International Air Transport Association
ICRP              International Commission on Radiological Protection
ICRU              International Commission on Radiation Units and Measurements
IMCO              Inter-Governmental Maritime Consultative Organization
IRCC              Ionizing Radiation Control Committee

LET               linear energy transfer
LSA               low specific activity
                                    16.5

-------
DARCQM-P 315-1


MPBB              maximum permissible body burden
MPC               maximum permissible concentration
MSHA              Mine Safety and Health Administration

NBS               U.S. National Bureau of Standards
NCRP              National Council on Radiation Protection and Measurements
NIOSH             National Institute of Occupational Safety and Health
NRC               U.S. Nuclear Regulatory Commission
NTA               nuclear track emulsion
NTIS              National Technical Information Service

OSHA              Occupational Safety and Health Administration

PF                protection factor

RAC               radiological assessment and control
RAM               remote area monitor
RBE               relative biological effectiveness
RPO               Radiation Protection Officer
RSR               radioactive shipment record

SEE               specific effective energy
SI                international system of measurement units
SOP               standing operating procedure
STP               standard temperature and pressure  (0°C, 760 mm Hg)

TI                transport index
TL                thermoluminescence
TLD               thermoluminescence dosimeter
TVL               tenth-value layer

WB                whole body
                                    16.6

-------
                                                              DARCOM-P 3E5-1

                    Section 16.4  ABBREVIATIONS AND SYMBOLS
ABBREVIATIONS

A              mass number
A              radionuclide activity or source activity
AF(T-S)        absorbed fraction of emitted energy, target from source

bis-KSB        p-bis-(0-methylstyryl) benzene
Bq             becquerel
Butyl-PBD      [2-(4'-tert-butylphenyl), 5-(4"-biphenylyl) - 1,3,4-oxadiazole]

C              Celsius
C              centigrade ?
c              centi- (10  )
cc             cubic centimeter
Ci             Curie
cnu            centimeter
cm             square centimeter
cpm1           counts per minute
cm             reciprocal centimeter or I/cm

D              radiation dose or absorbed radiation dose
D              radiation dose rate
d              day
dis            disintegration
dpm            disintegrations per minute
dps            disintegrations per second
dx             differential of x

E              radiation energy
e              base of natural logarithms (2.71828)
e.g.           exempli gratia (for example)
esu            electrostatic unit
eV             electron volt

F              Fahrenheit
f2             fraction of body burden in a given organ
ft             foot

g              gram
gal            gallon
Gy             gray

H              dose equivalent
Hj             committed dose equivalent to a target organ
Hy             dose-equivalent rate to a target organ
h              Planck's constant
hr             hour
Hz             hertz
                                    16,7

-------
DARCOM-P 385-1
I
!
1
i°d.
i.e.
In.
J
k
kg
keV
kV
L
Ib
In
K
m
m
flu
m3
m
max
mCi
MeV
min
ml
mm
mR
mrad
mrem
N
N
N
N
f\
n°
o.d.
oz
P
P
pCi
PPG
POPOP

photon flux
radiation intensity
original radiation intensity
inside diameter
id est (that is)
i nch
joul e
ki To- (103)
l:i "logram
ki'loelectron volt
k"i 1 ovol t
liter
pound
natural logarithm
mega- (106)
mass
meter ,
milli- (10"&)
square meter
cubic meter
maximum
mi 1 1 i curie
million electron volt
minute
mill il iter
mi 1 1 i meter
mi 1 1 iroentgen
mi 1 1 i rad
milli rem
neutron number
number of radioactive atoms present at a time t
product of modifying factors
number of radioactive atoms originally present
any number
outside diameter
ounce
specific ionization, or number of ion pairs produced by
radiation per path length
pico- (10~12)
picocurie
2,5-d:phenylc; ezole
I,4-bis-[2-(5-phenyloxazolyl )] -benzene
16.:

-------
                                                              DARCOM-P 3E5-1

Q              quality factor
q(t)           body burden at time t

R              roentgen
r              radius of a circle

S              source
S              surface area
s              distance
s              thickness
sec            second
SEE(T*S)       specific effective energy per disintegration, target from
                 source
Sv             sievert

T              kinetic energy
T              target
t              time
t.,p           radionuclide half-life

U              number of transformations in a source organ
u              mass unit

V              volt
v              velocity

wk             week

X              exposure
X              exposure rate
x              times (multiplication)

Y              radiation yield
yr             year

Z              atomic number


GREEK SYMBOLS

a              alpha particle
£              beta particle
Y              gamma ray
r              gamma-ray constant
E              effective absorbed energy per disintegration
E              energy imparted by ionizing radiation
0              angle
?.              effective decay constant
X              wavelength
>._             redionuclioe decay constant
 P
                                     16.9

-------
DARCOM-P 385-1

u               linear attenuation coefficient
u               micro- (ID"6)
p               n.ass energy absorption  coefficient
u|in             mass attenuation coefficient
uCi             microcurie
urn              micrometer
v               frequency
v               neutrino
IT               pi  (3.1416)
p               density
I               summation of
MATHEMATICAL  SYMBOLS

0              degree
*              percent
a              proportional to
               times  (multiplication)
                                    16.10

-------
                                                              DARCOM-P 385-1
                      Section 16.5  SELECTED CONVERSIONS
MULTIPLY

Length

  centimeters

  feet

  inches
  meters
  barns
  square centimeters
  square feet


  square inches


  square meters

Volume

  cubic centimeters



  cubic feet


  cubic inches



  cubic meters
BY
0.3937
3.28 x
30.48
0.3048
2.54
3.2B1
39.37
5280
io-2f

1.076 x
0.155
929
144
9.29 x
6.452
6.944 x
6.452 x
10.76
       ID
         '2
                                      10
6.102
3.531 x
2.642 x
10-3
2.832 x
7.481
28.32
16.39
5.7£,7 x
1.639 x
4.329 x
35.31
2.642 x
                                        -3
        10
      x 10
          -2
        10

        IO
          -2
        10
        10
          -3
                                                       TO OBTAIN
inches
feet
centimeters
meters
centimeters
feet
inches
feet
square
barns
square
square
square
square
square
square
square
square
square
                                centimeters

                                feet
                                inches
                                centimeters
                                inches
                                meters
                                centimeters
                                feet
                                meters
                                feet
                                      IO
cubic inches
cubic feet
U.S. gallons
liters
cubic meters
U.S. gallons
liters
cubic centimeters
cubic feet
liters
U.S. gallons
cubic feet
U.S. aallons
                                     16.11

-------
DARCOM-P 385-1

MULTIPLY

Volume (cont'd)

  gallons, U.S.



  liters




Mass

  grams
  kilograms
  ounces

  pounds

Energy

  British thermal units

  electron volts

  ergs
  gram-calories
  joules

  kilogram-calories
  megaelectron volts

 Radiation

  curies
  becquerels

  disintegrations/minute

  d'is integrations/second
EY
231
0.1337
3.785 x 10-3
3.785
3.53 x  10
61.02
0.2642
103
         -2
2.205 x 10
2.205
28.35
6.25 x 10
453.6

         -2
1.055 x 10
0.252
        -12
        -19
1.6 x 10
1.67x 10
10"7
6.24 x 10
6.24 x ID-
S'. 968 x 10
                                       11
                                        -3
9.48 x 10
3.968
         -4
1.6 x 10
3.7 x 10
        -6
        10
                                       12
2.22 x 10
3.7 x 10

Jp6
  -3
10 •*
1        n
2.7 x 10':,,
4.55 x 10 ,
4.55 x 10"
2.7 x 10"",:
2.7 x 10"5
1
iO OBTAIN
cubic inches
cubic feet
cubic centimeters
1iters
cubic feet
cubic inches
U.S. gallons
cubic centimeters
pounds
pounds
grams
pounds
grarcs
joules
kilogram-calories
ergs
joules
joules
electron volts
megaelectron volts
British thermal units
ergs
British thermal units
British thermal units
ergs
becquerels
disintegrations/minute
disintegrations/second
millicuries
microcuries
kilocuries
disintegrations/second
curies
millicuries
microcuries
mi Hi curies
microcuries
becquerels
                                     16.12

-------
MULTIPLY

Radiation (cont'd)

  gray

  microcuries

  mil 1iciuries

  rad


  rem

  roentgen
  sievert
Temperature

  degrees Celsius
  degrees Fahrenheit
         BY
         3.7 x 104
         2.22 x 10
         3.7 x 10?
         2.22 x 10
         10-2
         4
         1C2
         10.2
         10       -4
         2.58 x 10.
         1
                              2.082  x  103
                                      J2
- 32
                              i.6i  x  10;
                              7.03  x  10

                              5.44  x  10
         It
         1
1.8
0.5555
                  7
                               DARCOM-P 3E5-1

                        TO OBTAIN
rad
joules/kilogram
disintegrations/second
disintegrations/minute
disintegrations/second
disintegrations/minute
gray
joules/kilogram
ergs/gram
sievert
joules/kilogram
coulombs/kilogram
electrostatic units/
  cubic centimeter air
  (at STP)
ion pairs/cubic centi-
  meter air (at STP)
ion pairs/gram air
MeV/cubic centimeter
  air (at STP)(a)
MeV/gram airfe) ^
ergs/gram air^ '
rem
joules/kilogram
degrees Fahrenheit - 32
degrees Celsius
(a)  Assuming  that  the  averaoe  energy  expended per ion pair formed  is
    5.4  x  ID"13  ergs  (34  eV).
                                    16.13

-------
DARCOM-P 385-1










 Avogadro's  number



 Velocity of 1ight



 Electronic  charge



 Planck's constant
Mass of electron



Mass of proton



Mass of neutron
                    Section 16,6  rREQUENTLY USED CONSTANTS
N  = 6.0220 x 1023 moT1



c  = 2.997925 x 108 m/sec



e  = 0.16022 x 10"18 C



h  = 6.626 x 10"34 J-sec



   = 6.626 x 10"£7 erg-sec



   = 0.41355 x I0'}  eV-sec



me = 0.910953 x 10"30 kg



m  = 0.167265 x 10"26 kg



mn = 0.167495 x 10'26 kg
                                    16.14

-------
                                                              DARCOM-P  385-1

           Section 16.7  ADDRESSES FOR ORDERING REFERENCE DOCUMENTS
American National Standards Institute (ANSI)
     Sales Department
     American National Standards Institute
     1430 Broadway
     New York, NY  10018

Code of Federal Regulations (CFR)
     Superintendent of Documents
     U.S. Government Printing Office
     Washington, DC  20402

International Atomic Energy Agency  (IAEA)
     UNIPUB
     345 Park Avenue South
     New York, NY  10010

International Commission on Radiation Units and Measurements  (ICRU)
     1CRU Publications
     P.O. Box 30165
     Washington, DC  20014

International Commission on Radiological Protection  (ICRP)
     Pergamon Press
     Maxwell House
     Fairview Park
     Elmsford, NY  10523

National Council on Radiation Protection and Measurements  (NCRP)
     NCRP Publications
     P.O. Box. 30175
     Washington, DC  20014

National Technical Information Service  (NT1S)
     U. S. Department of Commerce
     5285 Port Royal Road
     Springfield, VA  22151

U. S. Department of Transportation (DOT)
     Superintendent of Documents
     U.S. Printing Office
     Washington, DC  20402

U.S. Nuclear Regulatory Commission  (NRC)
     Superintendent of Documents
     U.S. Printif:. Office
     Washington, DC  20402
                                     16.15

-------
DARCOH-P 385-1
                            Section 16.8  GLOSSARY
ABSORPTION:
ACCELERATOR (PARTICLE
ACCELERATOR):
ACTIVATION:
AC'IVITY:
ACUTE  EXPOSURE:

AGREEMENT STATE:
AIRBORNE CONTAMINATION:
AIR-WALL  IONIZATION CHAMBER:
ALARA:
ALPHA PARTICLE:
The process by which radiation  imparts somp or •
all of its energy to any material  through which
it passes.

A device for imparting large quantities of
Mnetic energy to electrically  charged particles
such as electrons, protons,  and helium ions.

The process of inducing radioactivity by
irradiation.

The number of nuclear transformations occurring
in a given quantity of material per unit time.
The unit of measure is the curie (Ci).

Radiation exposure of-short duration.

Any state in the United States  with which NRC
has made an effective agreement under Subsec-
tion 274(b) of the Atomic Energy Act of 1954, es
amended, relative to the licensing and control
•of radioactive material used or produced within
that state.

The term applied to radioactive contamination
loose in the air, filtered from the air, or
deposited from the air, as contrasted with
contamination spread by splashino, dripping,
etc.

An ionization chamber in which the materials  of
the wall and electrodes are so selected as  to
produce ionization essentially equivalent to
that in a free-air ionization chamber.  This  is
possible only over limited ranges of photon
energies.  Such a chamber is more appropriately
termed an "air-equivalent ionizetion chamber."

An acronym for "es low as is reasonably achiev-
able"; refers to an operating philosophy in
which occupational exposures are reduced as far
below specified limits as is reasonably
achievable.

A charged particle thez is emitted from the
nucleus of an atom and that has a mass and
charge equal in magnitude to those of  a helium
nucleus, i.e., two protons and two neutrons.
                                    16.16

-------
AMPLIFICATION:
ANALYZER, PULSE HEIGHT:
ANGULAR DEPENDENCE:
ANODE:
APPRAISAL:
ARTiFlCAL RADIOACTIVITY:
ATOM:
ATOMIC NUMBER:
ATTENUATION:
AUTHORIZED MATERIAL:
AVALANCHE:
BACKGROUND RAHATION:
                                DARCOM-P 385-1

As related^to radiation  detection instruments,
the process (gas,  electronic,  or both) by which
ionization effects are magnified to a degree
suitable for their measurement.

An electronic circuit that  sorts and records
pulses according  to their height.

The varying ability of an instrument to
accurately measure radiation,  depending on
its orientation with respect to the radiation
field.

A positive electrode; the electrode to which
negative ions are attracted.

A comprehensive evaluation  of the overall
adequacy rnd effectiveness  of a radiation
protection program.

Manmade radioactivity produced by particle
bombardment or electromagnetic irradiation, as
opposed to natural radioactivity.

The smallest unit of an  element that is capable
of entering into  a chemical reaction.

The number of protons in the nucleus of a
neutral atom of a nuclide.

The process by which a beam of radiation is
reduced in intensity or  energy when passing
through some material.

Radioactive materiel not requiring a specific
NRC license.  The receipt,  possession, use, or
transfer of radioactive  material requires spe-
cific authorization or permit by a specific
agency or service organization.

The multiplicative process  in which a single
charged particle  accelerated by  a strong elec-
tric field produces additional charged particles
through collision with neutral gas molecules.
This cumulative increase of ions is also known
as "Townsend ionization" or "Townsend avalanche."

Radiation arising frorr radioactive material
other than the one directly under considera-
tion.  Background radiation due  to cosmic rays
                                     16.17

-------
DARCOM-P 385-1
BEAM:
BECOUEREL:
 BETA  PARTICLE:
BIOASSAY:
BREMSSTRAHLUNG:
BYPRODUCT MATERIAL:
CALIBRATION:



CATHODE:


CELL  (BIOLOGICAL):


CHAIN REACTION):
                              and natural radioactivity is always present.
                              There may also be background radiation due to
                              the presence of radioactive substances in other
                              parts of a building, in the building material
                              itseV
        etc.
A unidirectional  or approximately unidirec-
tional  flow of electromagnetic  radiation or  of
particles.

The SI  unit of activity equal  to a nuclear
disintegration rate of 1 disintegration per
second.

A charged particle emitted from the nucleus  of
an atom, with a mass and charge equal  in mag-
nitude  to those of the electron.

An evaluation of the amount of radioactivity
taken into the body.

Secondary photon radiation produced by the
deceleration of charged particles passing
through matter.

Any material (except special nuclear material)
made radioactive by either exposure to
radiation, or the process of producing or using
special  nuclear material .

The determination of a measuring instrument's
variation from a standard, to ascertain
necessary correction factors.

A negative electrode; the electrode to which
positive ions are attracted.
The fundamental
in organisms.
                unit of structure and function
Any chemical or nuclear process in which some
products or energy released by the process are
instrumental in the continuation or magnifica-
tion of the process.
                                     16.18

-------
                                                             DARCOM-P 385-1
CHARACTERISTICS (DISCRETE)
RADIATION:
CHRONIC EXPOSURE:
COLLECTIVE DOSE EQUIVALENT;
COLLISION:
COMMODITY (RADIOACTIVE);
COMPOUND:
COMPTON EFFECT:
CONDENSER R-METER:
CONTAMINATION (RADIOACTIVE):
Radiation originating  from  an  atom after the
removal  of an  electron or the  excitation of the
nucleus.   The  wavelength of the  emitted radia-
tion is  specific,  depending only on the nuclide
and the  pf.'-ticular energy levels involved.

Radiation, exposure of  long  but not necessarily
continuous duration.

The sum of dose equivalents received by e given
population or  group of workers,  expressed in
units of person-rein.

An encounter between two subatomic particles
(including photons) that changes the initial
momentum and energy conditions.   The products
of the collision need  not be the same as the
initial  systems.

An item of government  property made up in whole
or in part of radioactive materials.  A national
stock number (NSN) (formerly called a federal
stock number (FSN)) or part number is assigned
to items that contain  radioactive material in
excess of 0.01 yd.

A distinct substance formed by the union of two
or more ingredients in definite proportions by
weight.

An attenuation process observed for x or gamma
radiation in which an  incident photon interacts
with an orbital electron  of an atom to produce a
recoil electron and a  scattered photon with an
energy less than that  of  the Incident photon,

An instrument consisting  of an air-walU ionize-
fion chamber together  with  auxiliary equipment
for charging and measuring  its voltage.  It is
used as an integrating instrument for measuring
the exposure of x or gamma  radiation in
roentgens (R).

The deposition of radioactive material in any
place where it is not  desired, and particularly
in any place where its presence might be harmful.
                                    16.19

-------
DARCOK-P 385-1
COUNT (RADIATION1
MEASUREMENTS!!
COUNTER:
CRITICAL:
CRITICAL ORGAN:
CROSS-CONTAMINATION:
CUMULATIVE DOSE (RADIATION)
CURIE:
DAUGHTER:

DECAY CONSTANT:
DECAY, RADIOACTIVE:
DECONTAMINATION:
The external  indication  of  a  device  designed  to
enumerate ionizing events.   It  may  refer to a
single detected event or to the total  number
registered in a given period  of time.   The  term
is often used erroneously to  designate a disinte-
gration, ionizing event, or voltage  pulse.

A gas-filled radiation detector (chamber or
tube) connected to an auxiliary electronic
circuit in such a way that  individual  pulses
from ionization events inside the chamber
register in an external  counting device.

Capable of sustaining (at a constant level)  a
chain reaction.  "Prompt critical"  means sustain-
ing a chain reaction without the aid of delayed
neutrons.

The organ of the body receiving a specified
radioisotope that results in the greatest
physiological damage to the body.  For exposure
to ionizing radiation from external  sources,  the
critical organs are the skin, blood-forming
organs, gonads, and eyes.

Contamination not from an original  source, but
acquired from another contaminated object.   The
term is used in laboratory, bioassay, and
counting-room work to refer to the spread of
contamination from contaminated samples to
relatively uncontaminated  samples, thus giving
erroneously high readings  to the latter.

The total dose resulting from  repeated  exposures
to radiation.

The special unit of activity (abbreviated Ci).
One curie equals exactly 3.7 x  10^ nuclear
disintegrations per second.

Synonym for decay product.

The fraction of the number of  atoms of  a radio-
active nuclide that decay  per  unit  time.

The disintegration of the  nucleus of  an unstable
nuclide by the spontaneous emission of  charged
particles and/or photons.

The reduction or removal of  radioactive contami-
nation from any given surface.
                                     16.20

-------
DELTA RAY:
DETECTOR, Ge(Li):
DETECTOR, INTEGRATING:
DETECTOR, RADIATION:
DETECTOR. SCINTILLATION;
DETECTOR, SOLID-STATE:
DETECTOR, TRACK (ETCH)
DISINTEGRATION, NUCLEAR:
                                DARCOX-P  385-1

Any secondary Ionizing  parf:cle  ejected  by
recoil when a primary ionizing  panicle  passes
througn matter.

A solid-state detector  in which  the crystal
used is germanium (Ge)  with a  minute quantity of
lithium (Li) impurity added to  stabilize  the
action.  (It is  sometimes referred to as  a
"jelly" detector.)

A detector that  measures  a total accumulated
radiation quantity  (such  as exposure or  dose)
rather than the  rate of accumulation of  the
radiation.  Devices that  accumulate and  hold
charges (e.g., electrometers)  and that indicate
measures proportional to  the total dose  are of
this type.  Examples of integrating detectors
are electrometers,  film badges,  pocket dosim-
eters, and neutron  activation  detectors.

Any device for converting radiant energy to a
form more suitable  for observation.  An  instru-
ment used to determine the presence, and some-
times the amount, of radiation.

A radiation detector whose response is a light
signal generated by the incident  radiation and a
scintillating medium.  The light  signal  is trans-
formed into an electronic signal  through an adja-
cent, optically coupled,  photo-sensitive device
such as a photomultiplier tube.

A generic name for a radiation  detector that
uses solid-state devices, such  as  the semi-
conductors germanium or silicon,  which respond
to  incident radiation with an electronically
measurable pulse.

A device  that records the paths  of heavy charged
particles in  a transparent solid.   The tracks
may be directly visible,  or they may  be enhanced
by  etching with an appropriate  reagent (such as
potassium hydroxide  for  etching cellulose
acetate).

A spontaneous nuclear transformation  (radio-
activity) characterized  by the  emission of
energy and/or mass frorr,  the nucleus.  t-Jnen
numbers of nuclei  are  involved,  the process  is
characterized by a definite half-life.
                                     16.21

-------
DARCOM-P 385-1

DOSE:
DOSE. ABSORBED:
DOSE, WHOLE-BODY:
DOSE EQUIVALENT:
DOSE METER, INTEGRATING;
DOSIMETER:
DOSIMETER. PERSONAL;
A general  term denoting the quantity of radia-
tion or energy absorbed.   For special  purposes,
the term must be appropriately qualified.   If
unqualified, it refers  to absorbed dose.

The amount of energy -imparted to matter in a
volume element by ionizing radiation,  divided by
the mass of irradiated' material  in that element.
Also called dose.  The  common unit of absorbed
dose is the rad, which  is equal  to 100 ergs of
absorbed energy per gram of material (or
0.01 J/kg).  The SI unit of absorbed dose is the
gray, which is equal to 100 rad or to 1 joule of
absorbed energy per kilogram of material.

The average uniform absorbed dose or dose
equivalent received by a person whose whole body
is exposed to ionizing radiation from an
external source.

The product of the absorbed dose, the quality
factor, and other modifying factors necessary to
evaluate the effects of irradiation received by
exposed persons.  This unit of measure takes
into account the particular characteristics of
the exposure. The common unit of dose equivalent
is the rem.  The SI unit is the sievert.  Absoroed
doses of different types of radiation are not
additive, but dose equivalents are, because they
express on a common scale the amount of damage
incurred.

An ionization chamber and measuring system
designed to determine the total radiation admin-
istered during an exposure.   In medical radiol-
ogy, the chamber is usually designed to be
placed on the patient's skin.  A device may be
included to terminate the exposure  when it  has
reached a particular value.

An instrument to detect and measure accumulated
radiation exposure.  In common usage, a pencil -
sized ionization chamber with a self-reading
electrometer, used  for personnel monitoring.

A dosimeter of small size  carried by a person to
determine  the exposure, absorbed dose, and/or
dose equivalent  received  during the carrying
time.  Also called  personal exposure meter.
                                    16.22

-------
                                                             DARCOM-P 385-1
DOSIMETER. POCKET:
DOSIMETER.
THERMOLUMINESCENCE:
DOSIMETRY, PHOTOGRAPHIC:
EFFICIENCY (OF COUNTERS)
ELASTIC COLLISION:
ELECTRODE:
ELECTRON:
ELECTRON VOLT:
ELEMENT:


EMULSION, NUCLEAR:
A dosimeter the  shape  and  size
pen with a clip,  to  be worn  in
fountain pen.
of a fountain
the pocket like
An integrating detector that  utilizes  a  phosphor
sensitive to ionizing  radiation.   The  phosphor
stores the energy of  the ionization within
itself and releases  it as  low-energy photons
(light) when heated.   The  total  amount of light
released is proportional to the  total  absorbed
dose.

The determination of  cumulative  radiation dose
using photographic film and density measurement.

A measure of the probability  that a count will
be recorded when radiation is incident on a
detector.  Uses of this term  vary considerably,
so it is well  to ascertain which factors (window
transmission,  sensitive volume,  energy depen-
dence, etc.) are included in  a given case.

A collision in which  there is no change either
in the internal energy of each participating
system or in the sum of their kinetic energies
of translation.

A conductor used to establish electrical contact
with a nonmetallic part of a  circuit.

A stable elementary particle  that has an electric
charge equal to ±1.60210 x 10"1<3 coulomb and a
rest mass equal to 98.1091 x  10-31 kg.

A unit of energy equivalent to the energy gained
by an electron in passing through a potential
difference of 1 volt.   Larger multiple units of
the electron volt are frequently used:  keV for
thousand or kilo-electron volts; MeV for million
or mega-electron volts.  1 eV =  1.6 x 10'^ erg.

A category of atoms all of which have the same
atomic number.

A photographic emulsion specially designed to
permit observation of the individual tracks of
ionizing particles.
                                    16.23

-------
DARCOM-P 385-1

ENERGY DEPENDENCE:
ENRICHED MATERIAL:
EXCITED STATE
(OF A NUCLEUS)
EXPOSURE:
EXPOSURE RATE:
EXTERNAL RADIATION:
FALLOUT:
FILTER (RADIOLOGY);
The characteristic response of a radiation
detector to a given range of radiation energies
or wavelengths, compared with the response of a
standard free-air chamber.

(1) Material in which the relative amount of one
or more isotopes of a constituent has been
increased.
                                          O'JC
.(2) Uranium in which the abundance of the  " U
isotope is increased above normal.

An unstable condition of the nucleus of an atom
after the entrance of a nuclear particle or
gamma-ray photon.

(1)  The incidence of radiation upon inanimate
or living matter by intent or accident.

(2)  For x or gamma radiation, the sum of the
electrical charges of all the ions of one sign
produced in air when all electrons liberated by
photons in a suitable small volume of air are
completely stopped in air, divided by the mass
of air in the volume.

The unit of exposure is the roentgen (R).

(1)  The exposure divided by the time over which
it was accumulated.

(2)  The increment of exposure during a suitably
small interval of time, divided by that interval
of time.

The usual unit of exposure rate is roentgens per
hour (R/hr).

Radiation from a source outside the body.

Radioactive debris from a nuclear detonation,
which is airborne or has been deposited on the
earth.  Special forms of fallout are "dry
fallout," "rainout," and "snowout."

Primary—A sheet of material, usually metal,
placed in a beam of radiation to absorb pre-
ferentially the less penetrating components.
                                    16.24

-------
                                                              DARCOK-P  385-1
FINGEF: DOSIMETER:
FISSILE:
FISSILE MATERIAL:
FISSION  (NUCLEAR):
FISSIONABLE:
FISSION PRODUCTS;

FLUENCE:
FLUORESCENCE:
PL HOROSCOPE:
GAS AMPLIFICATION:
GEIGER-MUELLER COUNTER:
Secondary—A sheet of material  of low atomic
number (relative to the primary filter)  placed
in the filtered beam of radiation to remove
characteristic radiation proauced by the primary
filter.

A dosimeter in the form of a ring to be  worn by
personnel to determine radiation doses to the
.hands.

A nuclide capable of undergoing fission  by
interaction with slow neutrons.

Plutonium-238, plutonium-239, plutonium-241,
uranium-233, uranium-235, or any material
containing any of the foregoing
[49 CFR 173.389(a) and 173.398(a)].

A nuclear transformation characterized by split-
ting of a nucleus into at least two other nuclei
and the release of a relatively large amount of
energy.

Pertaining to a nuclide that is capable  of
undergoing fission by any process.

Elements or compounds resulting from fission.

The number of particles passing through  a unit
cross-sectional area.

The emission of radiation of particular wave-
lengths by a substance as a result of the absorp-
tion of radiation of shorter wavelengths.  This
emission occurs essentially only during the
irradiation.

A fluorescent screen, suitably mounted with
respect to an x-ray tube for ease of observation
and protection, used for indirect visualization
(by x rays) of internal organs in the body or
internal structures in apparatus or in masses of
material.

As applied to gas-ionization instruments for
detecting-radiation, the ratio of the charge
collected to the charge produced by the initial
ionizing event.

A highly sensitive, gas-filled radiation-
measuring device.   It operates at voltages high
enough to produce avalanche ionization.
                                    16.25

-------
DARCOM-P 385-1


GEOMETRY. GOOD:





GEOMETRY. POOR:
GEOMETRY (RADIATION):
GLOW CURVE:
GLOW PEAK:
GRAY:
GROUND STATE:
HALF-LIFE. BIOLOGICAL:
HALF-LIFE, EFFECTIVE:
'In nuclear physics  measurements,  an  arrangement
of source and detecting  equipment that  introduces
little error when a finite  source size  and
finite detector aperture are  used.

In a nuclear experiment, an arrangement in  which
the angular aperture between  the  source and
detector is large,  introducing  into  the meas-
urement a comparatively  large uncertainty for
which a correction  may be necessary.

A nuclear physics term referring  to  the physical
relationship and symmetry of  the  parts  of a
radiation detection assembly.   Counting effi-
ciency is closely related to  geometry.

In thermoluminescence dosimetry,  a graph of the
released luminescence photon  fluence as a func-
tion of temperature or time of  heating.  The
area under the bell-shaped curve  plotted against
time is proportional to  the total absorbed  dose
or exposure.

In thermoluminescence dosimetry,  the time or
temperature during heating of a thermolumi-
nescence phosphor at which the  release  rate of
the luminescence photons is maximum.

The SI unit of absorbed  dose, equal  to  the
absorbed energy from ionizing radiation of
1 joule/kg, and equal to 100  rads.

The state of a nucleus,  atom, or molecule at its
lowest energy.  All other states  are "excited."

The time required for the body  to eliminate
one-half of an administered dosage of any
substance by processes of elimination.
Approximately the same for both stable  and
radioactive isotopes of a particular element.

The time required for a  radioactive element  in
an animal body to be diminished 50% as  a result
of the combined action of radioactive decay  and
biological elimination.

Effective half-life

  Biological half-life  x Radioactive half-life
  Biological half-life  + Radioactive half-life
                                    16.26

-------
HALF-LIFE, RADIOACTIVE:
HALF-VALUE LAYER
(HALF THICKNESS) (HVL):
                                DARCOK-P  385-1

Tne time required  for a  radioactive  substance  to
lose 50* of its  activity by  decay.   Each  radio-
nuclide has a unique  half-life.

The thickness of a specified substance that,
when introduced  into  the path of a  given  beam  of
radiation, reduces the exposure  rate by one-half.
HEALTH PHYSICS:
A science and profession devoted to protecting
man and the environment against unnecessary
radiation exposure.
HOLE (SOLID-STATE THEORY):
INDUCED RADIOACTIVITY:
INELASTIC COLLISION:
INFRARED RADIATION:
INGESTION
'(•'•' RADIOACTIVITY)

INHALATION •
TO." RADIOACTIVITY):
INTENSITY:
INTENSITY, RAD IATI ON:
A position in the valence bands of semiconductor
or insulating materials denoting the absence of
an electron.   Such a position carries a positive
charge that (like an electron) is able to
migrate within the band.

Radioactivity produced in a substance after
bombardment with neutrons or other particles.
The resulting activity is "natural radio-
activity" if formed by nuclear reactions
occurring in nature, and "artificial radio-
activity" if the reactions are caused by man.

A collision in which there are changes both in
the internal  energy of one or more of the col-
liding systems and in the sums of the kinetic
energies of translation before and after the
collision.

Invisible thermal radiation whose wavelength is
longer than the red segment of the visible
spectrum.

The entry of radioactivity into the body through
the mouth.

The entry of radioactivity into the body through
the breathing of airborne radioactive particulate
matter.

The amount of energy per unit time passing
through a unit area perpendicular to the line  of
propagation at the point in question.

A generic term for the magnitude of a radiation
quantity.
                                     16.27

-------
    )>'   385-1


INTENSITY. SOURCE:





INTERNAL RADIATION:




IN-VIVO COUNTING:
ION:
IONIZATION:
IONIZATION CHAMBER:
IONIZING RADIATION CONTROL
COMMITTEE:
IONIZIN6-RADIATION-
PRODUCING DEVICES:
ION PAIR:
ISOMERS:
A generic term for the magnitude of a source
emission rate.  The source intensity of a
radioisotope source is related to its dis-
integration rate in curies or bequerels.

Radiation from a source within the body (as a
result of the deposition of radionuclides in
body tissues).

Measurements of internal radiation made at the
surface (outside) of the body and based on the
fact that radioisotopes emit radiation that can
traverse the tissues and be measured outside the
organism.  In-vivo counting is synonymous with
whole-body counting.

An atomic particle or atom bearing an electric
charge, either negative or positive.

The process by which a neutral atom or molecule
acquires a positive or negative charge.

An instrument designed to measure a quantity of
ionizing radiation in terms of the charge of
electricity associated with ions produced within
a defined volume.

A group of qualified personnel officially
appointed by a commander to set local policy and
to guide the radiation protection program.

Electronic devices that are capable of making
ionizing radiation.  Examples are x-ray
machines, linear accelerators, and electron
microscopes.

Two particles of opposite charge, usually refer-
ring to the electron and the positive atomic or
molecular residue resulting from the inter-
action of ionizing radiation with the orbital
electrons of atoms.

Nuclides with the same number of neutrons and
protons but capable of existing, for a mea-
surable time, in different quantum states with
different energies and. radioactive properties.
Commonly, the isomer of higher energy decays to
one with lower energy by the process of iso-
metric transition.
                                    16.28

-------
                                                              DARCOM-P 385-i
ISOTOPES:
JOULE:
LATENT PERIOD:
LEAKAGE RADIATION:
LICENSE (SPECIFIC):
LICENSE-EXEMPT MATERIAL
ITEMS:
LICENSED MATERIAL:
LINEAR ACCELERATOR:
LiNL/\8 ENERGY TRANSFER
TLtTT;
MANIPULATOR:
MAN-REM:
Nuclides that have  the  same  number  of  protons  in
their nuclei, hence  the  same  atomic number,  but
that differ in the  number of  neutrons  and  there-
fore in the mass  number.  Isotopes  of  a
particular element  have  almost  identical
chemical properties.  The term  should  not  be
used as a synonym for nuclide.

The unit for work and energy, equal to 10   ergs.

The interval  of seeming  inactivity  between the
time of irradiation  and  the  appearance of  an
effect.

Radiation emerging  from a surface,  a body  of
material, or a region in space.

A document issued by NRC under  10 CFR that gives
the bearer the right, to  procure,  receive,  store,
transfer, use, export,  and  import specified
radioactive items under specific  terms.

Radioactive material  not subject  to NRC regula-
tions, or exempt from NRC licensing under  10 CFR.

Source, special nuclear, or byproduct material
received, stored, possessed, used,  or trans-
ferred under a general  or specific  license -
issued by NRC or an Agreement State.

A device for accelerating charged particles.  It
employs alternate electrodes and  gaps arranged
in a straight line, so  proportioned that when
potentials are varied in the proper amplitude
and frequency', particles passing  through the
waveguide receive successive increments of
energy.

The linear rate of  loss of  energy  (locally
absorbed) over distance by  an ionizing particle
moving in a material medium.  The  usual unit of
LET is keV/pm.

Mechanical hands or some other device for
performing work behind  a barrier or in a
glove box.

A unit of population dose equivalent  or collec-
tive dose equivalent.  The  number  of  man-rems of
dose equivalent is  equal to the  product of  the
population and the  average  dose  equivalent  in
rem common to that  population.
                                     ""6,29

-------
DARCOM-P 385-1

MAXIMUM CREDIBLE ACCIDENT:



MICROWAVE:




MOLECULE:



MONITORING:



MONTE CARLO METHOD:
NATURAL RADIOACTIVITY:
NATURALLY OCCURRING
RADIOACTIVE MATERIALS:
NEUTRINO:
NEUTRON:
NUCLEON:
NUCLEUS (NUCLEAR):
NUCLIDE:
The worst accident in a  reactor  or nuclear energy
installation that, by agreement,  need  be  taken
into account in devising protective measures.

An electromagnetic wave  with  a wavelength of
approximately 1 millimeter to 1  meter  and
corresponding to frequencies  of  about  300 to
300,000 megacycles per second.

The smallest unit of a compound,  consisting of
two or more atoms held together  by chemical
bonds,

Periodic or continuous determination of the
amount of ionizing radiation  or  radioactive
contamination present in an occupied region.

A method permitting the computer solution of
physics problems, such as those  of neutron
transport, by determining the history of a large
number of elementary events by the application
of the mathematical theory of random variables.

The property of radioactivity exhibited by more
than 50 naturally occurring radionuclides.

Radioactive isotopes, such as radium and radon,
that are found in nature but are not classified
as source material.

A neutral particle of very small rest mass
originally postulated to account for the con-
tinuous distribution of energy among particles
in the beta-decay process.

One of three elementary particles, which is part
of all nuclei heavier than hydrogen.

The common name for a constituent  particle of
the nucleus.  Applied to a proton  or neutron.

That part of an atom  in which the  total  positive
electric charge and most of  the mass are
concentrated.

A species of atom characterized by the constitu-
tion of its nucleus.  The  nuclear  constitution
is specified by the number of protons  (Z), num-
ber of neutrons  (N),  and energy content;  or,
alternatively, by the atomic number (Z),  mass
number (A = N + Z), and atomic mass.   To be
                                    16.30

-------
PAIR PRODUCTION:
PARENT:
PERSONNEL MONITOR:
PHANTOM:
PHOSPHORESCENCE:
PHOTOELECTRIC EFFECT:
PHOTON:
                               DAKCOM-P  365-1

regarded as a  distinct  nuclide, the  atom  must  be
capable of existing  for a  measurable time.
Thus, nuclear  isomers are  separate  nuclides,
whereas promptly decaying  excited  nuclear states
and unstable intermediates in  nuclear reactions
are not so considered.

An absorption  process for  x and gamma radiation
in which the incident photon is annihilated in
the vicinity of the  nucleus of the  absorbing
atom, with subsequent production  of an electron
and positron pair.   This reaction  occurs  only
for incident photon  energies exceeding 1.02 MeV.

A radionuclide which, upon disintegration, yields
a specified nuclide, either directly or as a
later member of a radioactive series.

An instrument that measures a radiation quantity
proportional to dose equivalent,  for use by an
individual working in a radiation area.

A volume of material approximating as closely as
possible the density and effective atomic number
of body tissue.  Ideally,  a phantom  should
absorb radiation in the same way tissues does.
Radiation dose measurements made within or on a
phantom provide a means of determining the radia-
tion dose within or on a body under  similar-
exposure conditions.   Some materials commonly
used in phantoms are water, Masonite, pressed
wood, and beeswax.

The emission of radiation by a substance as a
result of the. previous absorption of radiation
of shorter wavelength.   In contrast  to fluores-
cent emissions, the phosphorescent  emissions may
continue for a considerable time after cessation
of the exciting irradiation.

The process by which a  photon ejects an  electron
from an atom.  All  the  energy of the photon is
absorbed in ejecting the  electron  and in  impart-
ing kinetic energy  to  it.

A quantity of electromagnetic energy (E)  whose
value  in joules is  the  product of  it:- freauency
(0)  in hertz and Planck's  constant (h).   The
equation is E = hO.
                                    16.31

-------
PIG;


PRIMARY IONIZATION:
PROPORTIONAL COUNTER:
PROTECTIVE CLOTHING:



PROTECTIVE EQUIPMENT:


PROTON:




PURGING:



PYROPHORIC:

QUALITY FACTOR (Q):
QUENCHING:
RAD:
A container, usually lead,  usea  to  ship or store
radioactive materials.

(1) In collision theory:   the  ionization pro-
duced by primary particles,  as contrasted with
total ionization, which  includes the secondary
ionization produced by delta  rays.

(2) In counter tubes:  the  total ionization
produced by incident radiation without gas
amplification.

A gas-filled radiation detector  tube operated in
that range of applied voltage  in which the
charge collected per isolated  count is propor-
tional to the charge liberated by the original
ionizing event.   The range  of  applied voltage
depends upon the type and energy of the incident
radiation.

The clothing worn by radiation workers to prevent
radioactive contamination of the body or personal
clothing.

Safety devices such as  goggles or clothing used
to do a job safely.

An elementary nuclear particle with a positive
electric charge equal numerically to the charge
of the electron and a mass  of 1.007277 mass
units.

The removal of material  from a  system or pipe by
adding another material, such as blowing with
air.

Igniting spontaneously on exposure to air.

The factor dependent on  linear energy transfer
by which absorbed doses  are multiplied to obtain
(for radiation protection purposes) a quantity
that expresses the effect of the absorbed dose
on a common scale for all ionizing radiations.

The process of inhibiting continuous or multiple
discharge in a counter tube that uses gas
amplification.

The unit of absorbed dose equal  to 0.01 J/kg  in
any medium.
                                    10.32

-------
RADIATION:
RADIATION AREA:
RADIATION. DIRECT:
RADIATION. INDIRECT:
RADIATION, IONIZING:
RADIATION. PRIMARY:
RADIATION. SCATTERED:
RADIATION. SECONDARY;
RADIATION CONTROL OFFICER:
RADIATION HAZARD:
RADIATION PROTECTION
OFFICER:
                                DARCO.M-P 385-1

Energy traveling through space in the form of
waves, particles, or bundles  called photons.

An area or item of equipment  requiring access
control for personnel  protective purposes; an
area or item of equipment presenting personnel
hazards due to radiation or contamination.

Radiation reaching a given location directly
from an emitting source without collision or
energy degradation.  Also called unscattered
or uncollided radiation.

Radiation reaching a given location after having
been scattered at least once.   Also called
scattered radiation.

Radiation composed of particles that are them-
selves ionized (directly ionizing radiation)  or
that are able to ionize other atoms by reaction
with them (indirectly ionizing radiation).

(1) Radiation emitted by a primary nuclear reac-
tion source (as opposed to radiation emitted  by
subsequent nuclear or atomic interactions as  a
result of primary radiation interactions).

(2) Radiation originating within an emitting
source (such as the core of a nuclear reactor).

Radiation reaching a given location after having
undergone at least one scattering.  See also
radiation, indirect.

Radiation emitted by some nuclear or atomic
process as a'result of previous nuclear or
atomic interactions by a primary radiation
source.  Example:  capture-gamma radiation.

An officer, enlisted person, or DA civilian
employee appointed by each major Army commander
to manage the radiation protection program for
the major command.

The presumed risk or deleterious effects
attributable to deliberate, accidental, or
natural exposure to radiation.

A person appointed by the commander  to give
advice on the hazards of ionizing  radiation  and
to supply effective ways to control  these
hazards.
                                    15.23

-------
DARCOM-P 385-1
RADIOACTIVE CONTROLLED
ITEMS:
RADIOACTIVE INDIVIDUALLY
CONTROLLED ITEMS:
RADIOACTIVE MATERIAL:
RADIOACTIVE MATERIAL
CONTROL POINT:
RADIOACTIVE WASTE:
RADIOACTIVITY:
RADIOBIOLOGY:
RADIOCHEMISTRY:
RADIOGRAPH:
All commodities,  components,  and end items
containing radioactive  material  that are
controlled with respect to maintenance,  dis-
posal, and bulk storage.   Items  requiring
additional controls are listed in 10 CFR 30.71.

Items that are assigned national stock numbers
and must be controlled  to the extent that their
integrity and location  are known by the  licensee
or designated agent (control  point) at all
times.

Any material  or combination of materials that
spontaneously gives off ionizing radiation.
This includes natural  elements such as radium,
and accelerator-made radionuclides.

Any Army element (including the RCO) that has
been designated by a .major Army commander to
control radioactive items within the command.

Waste materials that include the following:
a.  property contaminated to the extent  that
    decontamination is  economically unsound
b.  surplus radioactive material whose sale,
    transfer, or donation is prohibited
c.  surplus radioactive material that is
    determined to be unwanted after being
    advertised as surplus
d.  waste that is radioactive due to production,
    possession, or use  of radioactive material.

A natural and spontaneous process by which the
unstable atoms of an element emit or radiate
excess energy from their nuclei as particles or
photons and thus change (or decay) to atoms of a
different element or to a lower energy form of
the original  element.

The branch of biology that deals with the
effects of radiation on biological systems.

The aspects of chemistry connected with  radio-
nuclides and their properties, with the  behavior
of minute quantities of radioactive materials,
and with the use of radionuclides  in the study
of chemical problems.

A shadow picture produced by passing x  rays or
gamma rays through an object and recording the
variations in the intensity of  the emergent rays
on photographic or sensitized film.
                                    16.34

-------
                                                             DARCOM-P 385-1
RADIOLOGY:



RADIOPHARMACEUTICAL:


RADJOSENSJ7IVITY:



REM:
RESTRICTED AREA:
ROENTGEN:
SATURATION (IONIZATION
CHAMBER):
SCATTERING:
f/-V ED SOURCE:
SECONDARY IONI2ATION:

SELF-AESORPTION:
The oranch  of medicine "that  deels  with  the  diag-
nostic and  therapeutic applications  of  radiant
energy, including  x  rays  and radionuclioes.

A pharmaceutical compound that has been tagged
with a radionuclide.

The relative susceptibility  of cells,  tissues,
organs, organisms, or any living substance  to
the injurious action of  radiation.

A special  unit of  dose equivalent.  The dose
equivalent  in rems is numerically equal to  the
absorbed dose in rads multiplied by the quality
factor and  any other necessary modifying
factors.

Any area in a radiation  facility to which access
is controlled by the licensee for purposes  of
protecting  individuals from exposure to radia-
tion and radioactive materials.

One roentgen is the quantity of charge liberated
by x or gamma radiation  and is equal to 2.58 x
10-* coulombs per kilogram of dry air.  It is
equivalent to the energy absorption of x or gamma
radiation of 87.7 ergs/g of air or 96.5 ergs/g
of tissue (0.00877 J/kg and 0.00965 J/kg).

The condition in an  ionization chamber when the
applied voltage is sufficient to  collect all the
ions formed from the absorption of  radiation, but
insufficient to cause ionization  by collisions.

Change of direction  of subatomic  particles or
photons as a result  of a collision  or
interaction.

Any radioactive material that  is  permanently
bonded or fixed in a capsule  or matrix  designed
to prevent the release or dispersal of the mate-
rial under the most  severe  conditions  encoun-
tered  in normal use  or handling.

Ionization produced  by delta  rays.

The absorption of radiation (emitted  by radio-
ac:ive  ator^s) by  the material  in  which the  ctc
ere locateo;-in particular, the  absorption  of
radiation within  a  sample being  assayed.
                                     16.25

-------
DARCOM-P 365-1
SERIES, RADIOACTIVE:
SHIELD:
SIEVERT:
SOURCE, RADIATION:
SOURCE GEOMETRY:
SOURCE MATERIAL:
SPECIAL NUCLEAR MATERIAL:
SPECIFIC ACTIVITY:
SPECIFIC IONIZATION:
A succession of nuclides,  each  of which  trans-
forms by radioactive disintegration  into the
next until  a stable nuclide  results.   The first
member is called the "parent,"  the intermediate
members are called "daughters,"  and  the  final
stable member is called  the  "end product."

A body of material used  to prevent or reduce  the
passage of particles or  radiation.

The SI unit of dose equivalent  equal  to  the
absorbed dose in grays multiplied by the quality
factor and any other necessary  modifying
factors.

Materials or devices that  make  or are capable of
making ionizing radiation, including:
a.  naturally occuring radioactive materials
b.  byproduct materials
c.  source materials
d.  special nuclear materials
e.  fission products
f.  materials containing induced or deposited
    radioactivity
g.  radiographic and fluoroscopic equipment
h.  particle generators  and accelerators
i.  electronic equipment that uses klystrons,
    magnetrons, or other electron tubes  that
    produce x rays.

The shape, size, and configuration of a  radia-
tion source, taken as a  whole.

Uranium or thorium or a  combination of both, in
any physical form, or ores that contain
one-twentieth or more by weight of uranium or
thorium or any combination.   Source material
does not include special nuclear material.

Plutonium or uranium enriched in  isotope  233 or
235, and any other material NRC determines to be
special nuclear material.   Any materiel   (except
source material) artificially enriched by either
isotope.

The total activity of a given nuclide per gram
of a compound, element,  or radioactive nuclide.

The number of ion pairs produced  per unit path
length of ionizing radiation in a medium  (e.g.,
per cm of air or per micron of  tissue).
                                     16.35

-------
SPECTROMETER (NUCLEAR):
STABLE ISOTOPE:

SURVEY (RADIATION);
THIMBLE IONIZATION CHAMBER:
THRESHOLD DOSE:
TISSUE DOSE:
TISSUE-EQUIVALENT
IOKIZATION CHAMBER:
TISSUE-EQUIVALENT MATERIAL:
rRACK:
  r FACTOR:
                                DARCCM-P 385-1
A device or instrument,  usually electroaic,
capable of measuring  the energy distribution of
nuclear radiations.

A nor.radioactive isotope of an element.

An evaluation of the  radiation hazard associated
with the production,  use, or existence of
radioactive materials or other sources of
radiation under specific conditions.   The
evaluation usually includes:
a.  a physical survey of the disposition of
    materials and equipment
b.  measurements or estimates of the levels  of
    radiation involved
c.  predictions of hazards resulting from
    expected or possible changes in materials or
    equipment.

A small cylindrical, or spherical ionization
chamber, usually with walls of organi'c material.

The minimum absorbed dose that produces a
detectable effect.

The absorbed dose received by tissue in a region
of interest, expressed in rads.

An ionization chamber in which the materiel  of
the walls, electrodes, and gas are so selected
as to produce a response to radiation similar to
the response of tissue.

A liquid or solid whose absorbing and scattering
properties for a given radiation simulate as
closely as- possible those of a given biological
material,  such as fat, bone, or muscle.  For
muscle or  soft tissue, water is usually the best
tissue-equivalent material.

The visual manifestation of the path of an
ionizing particle in a chamber or photographic
emulsion.

For a mechanical radiation  source, the fraction
of the workload during which the useful beam  is
pointed toward the area  in  question.
                                     16.37

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DAfCOM-P 385-1


USEFUL BEAM:
VALENCE:
VAN DE GRAAFF ACCELERATOR:
VOLUME. SENSITIVE:
WORKLOAD:
X-RAYS:
The radiation that  passes  through  the  window,
aperture, cone,  or  other collimating device  of
the housing for  a radiation  source.  Sometimes
called "primary  beam."

The number representing the  combining  or dis-
placing power of an atom;  the  number of elec-
trons lost, gained, or snared  by  en atom in  a
compound; the number of hydrogen  atoms with
which an atom will  combine or  which  it will
displace.

An electrostatic machine  in  which  electrical
charge is carried  into the high-voltage terminal
by a belt made of an insulating material moving
at a high speed. The particles are  then accel-
erated along a discharge  path  through  a vacuum
tube by the potential difference  between the
insulated terminal  and the grounded end of the
accelerator.

The portion of a counter  tube  or  ionization
chamber that responds to  a specific  radiation.

A quantity indicating the  average weekly output
of a mechanical  radiation  source.   For example,
for a clinical x-ray apparatus,  the workload can
be specified in  milliampere-minutes  per week,  at
a particular (usually maximum) x-ray  tube
voltage.

Penetrating electromagnetic  radiations whose
wavelengths are  shorter  than those of visible
light.  They are usually  produced by  bombarding
a metallic target with fast  electrons  in a high
vacuum.  In nuclear reactions, it is  customary
to refer to photons originating  in the nucleus
as gamma rays, and  those  originating  in the
extranuclear part  of the  atom as  x rays.  These
rays are sometimes  called roentgen rays afte-
their discoverer, W. K.  Roentgen.

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                                   Appendix A                 DARCOM-P 365-1
                                 BIBLIOGRAPHY
ALARA	A-2
ARMY REGULATIONS	A-2
CONTAMINATION AND DECONTAMINATION  	    A-3
DOSE CALCULATIONS	A-A
EMERGENCY  PREPAREDNESS   	    A-6
ENVIRONMENTAL MONITORING 	    A-9
FACILITY DESIGN     	    A-10
GENERAL HEALTH PHYSICS PROGRAMS    .     .     .          .          .     .    A-12
GENERAL TEXTBOOKS	A-U
INSTRUMENTATION AND DOSIMETRY 	    A-15
QUALITY ASSURANCE/QUALITY CONTROL  	    A-17
RADIATION-GENERATING DEVICES  	    A-18
RADIOCHEMICAL ANALYSIS   	 '   A-19
RECORDS	A-19
SHIELDING-	A-20
STATISTICS	•	A-21
SURVEYING AND MONITORING 	    A-21
I'KAi'NING	A-23
TRANSPORTATION 	    A-24
WASTE MONITORING	A-25
                                      A-l

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DARCOM-P 385-1

                                 BIBLIOGRAPHY




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National Institute for Occupational Safety and Health.  1976.  A^ Guide to
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  Institute for Occupational Safety and Health, Cincinnati,  Ohio.

Odor, D. L.  1975.  "The Health Physicist and Nuclear Power  Plant Design."
  Health Phys. 29:795.

Patty, F. A., ed.  1958.  "Ventilation,"  in Industrial Hygiene end Toxicology.
  Interscience Publishers, Inc., New York-.

U.S. Nuclear Regulatory Commission.  1975.  Design, Testing  and Maintenance
  Criteria for Atmospheric Clean-Lip System Air Filtration and Acsorption  Units
  of Lioht-Weter-Cooled Nuclear Power Plants.  Regulatory Guide 1.52,
  Wcsnington, D.C.

 .•f;rl"ing, 0. P., Jr.,  D. G. Oliver, Jr.,  and E. B. Moore.   1973.  "A  Storage
  and Handling Facility for Californium-252 Medical Sources."  Health Phys.
  25:163.                                                               "~
                                    A-11

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DARCOM-P 385-1

                        GENERAL HEALTH PHYSICS PROGRAMS
American National Standards Institute.  1976.  Glossary of Terms in Nuclear
  Science and Technology.  ANSI-N1.1, Washington, D.C.

Amtey, S. R., and M. D. Allen.  1976.  "Personnel, Space, and Budget Needs of
  a University Radiation Safety Program," in Proceedings cr the 9th Midyear
  Topical Symposium of the Health Physics Society on Operational Health
  Physics.   Rocky Mountain Chapter, Health Physics Society, P.O. Box 3229,
  Boulder,  Colorado 80303.
                  I
Belvin, E.  A., and G. F. Stone.  1972.  "Management and Health Physics
  Interaction in a Large Federal Agency," in Health Physics Operational
  Monitoring, Vol. 3, C. A. Willis and J. S. Handloser, eds.  Gordon and
  Breacn Science Publishers, New York.

Boggs, R. F., W. E. Gundaker, W. M. Brobeck and R. G. Altes.  1972.
  "Health Physics Programs at Low Energy Particle Accelerator Facilities,"
  in Health Physics Operational Monitoring, Vol. 1, C.  A. Willis and
  J. S. Handloser, eas.  Gordon and Breach Science Publishers, New York.

Brodsky, A.  1965.  "Determining Industrial Hygiene Requirements for
  Installations Using Radioactive Materials."  Amer. Ind. Hyg. Ass. J.
  26:294-310.

Ebert, H.,  H. Eriskat, A. Oudig and G. Uzzan, eds.  1981.  Radiation Protec-
  tion Optimization, Present Experience and Methods.  Pergamon Press,
  New York.

Evans, E. A., M. Muranotsu, et al.  1977.  Radiotracer Techniques and Applica-
  tions.  Marcel Dekker, Inc., New York.

Fitzgerald, J. J.  1969.  Applied Radiation Protection and Control.  Gordon
  and Breach Science Publishers, New York.

Ice, R. D.   1971.  "Establishment of a University Radiation Safety Office."
  Health Phys. 20:444.

International  Atomic Energy Agency.  1965.  Provision of Radiological.
  Protection Services.  IAEA Safety Series No. 13, Vienna.

International  Atomic Energy Agency.  1973.  Radiation Protection Procedures.
  IAEA Safety Series No: 38, Vienna.

International  Atomic Energy Agency.  1973.  Safe Handling of Radionuclides.
  IAEA Safety Series No. 1, Vienna.

International  Commission on Radiological Protection.  1971.  Data  for
  Protection Against lonizijig Radiation from External Sources.   ICRP 21,
  supplement to ICRP 15.  Pergamon Press, hew York.
                                    A-12

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                                                              DARCO.M-P 385-1

Lindell, B.  1971.  "Professional Responsibilities of the Health Physicist in
  Relation to the Medical Profession."  Health Phys.  20:475.

McConnon, D.   1972.   "A Health Physics Program for a  Plutonium Fuel  Fabrica-
  tion Facility," in Health Physics Operational Monitoring,  Vol. 1,
  C. A. Willis and J.  S.  Handloser, eds.   Gordon and  Breach  Science
  Publishers, New York.

Miller, K. L.  1976.  "A Radiation Safety Program for a New  Medical  Center,"
  in Proceedings of the 9th Midyear Topical Symposium of the Health   Physics
  Society on Operational  Health Physics.Rocky Mountain Chapter, Health
  Physics Society, P.O. Box 3229, Boulder, Colorado 80303.

National Council on Radiation Protection and Measurements.  1971.  Basic
  Radiation Protection Criteria.  NCRP 39, NCRP Publications, Washington, D.C.

National Council on Radiation Protection and Measurements.  1971.  Protection
  Against Neutron Radiation.  NCRP 38.  Washington, D.C.

National Council on Radiation Protection and Measurements.  1978.  Opera-
  tional Radiation Safety Program.  NCRP 59, Washington, D.C.

Norwood, W. D.  1975.   Health Protection of Radiation Workers.  Charles C.
  Thomas, Publisher, Springfield, Illinois.

Patterson, H. W., and R.  H. Thomas.  1973.  Accelerator Health  Physics.
  Academic Press, New York.

Remley, M. E.  1972.  "Management Requirements on Health Physics," in Health
  Physics Operational  Monitoring, Vol. 3, C. A. Willis and J. S. Handloser,
  eds.   Gordon and Breach Science Publishers, New York.

Tolen,  J. H.   1972.   "Organization of a Small-Scale Radiation Safey Program,"
  in Health Physics Operational Monitoring, Vol. 1, C. A. Willis, and
  J. S. Handloser, eds.  Gordon and Breach Science Publishers,  New York.

Unruh,  C. M.   1972.   Problems for Radiation Protection Specialists for Safety
  Evaluations.  BNWL-SA-4459, Pacific Northwest Laboratory,  Richiand,
  Washington.

U.S. Code of Federal Regulations.  1982.   Title 10, Parts 0 to  199.   U.S.
  Government Printing Office, Washington, D.C.

U.S. Code of Federal Regulations..  1982.   Title 10, Part 20,  "Standards for
  Protection  Against Radiation."  Washington, D.C.

U.S. Code of Federal Regulations.  1982.   Title JO, Part 30,  "Rules of General
  Aopiicability to Licensing of Byproduct Mater-'al."  Washington, D.C.
                                     A-13

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DARCOM-P 385-1

                              "GENERAL TEXTBOOKS


Adams, A. S., and W. M. Lowder, eds.  1964.  The Natural  Radiation Environ-
  ment.  University of Chicago Press, Chicago, Illinois.

Arena, V.  1971.  Ionizing Radiation and Life.  The C.  B.  Mosby Company,
  St. Louis, Missouri.

Attix, F. H., E. Tochilin and W. C. Roesch, eds.  1968.   Fundamentals.  Vol. 1
  of Radiation Dosimetry, 2nd ed.  Academic Press, New York.
                 1
Attix, F. H., E. Tochilin and W. C. Roesch, eds.  1968.   Instrumentation,
  Vol. 2 of Radiation Dosimetry, 2nd ed.  Academic Press,  New York.

Blatz, H.  1964.  Introduction to Radiological Health.   McGraw-Hill, New York.

Bushong, S. C.  1975.  Radiological Science for Technologists.  The C.V. Mosby
  Company, St. Louis, Missouri.

Cember, H.  1969.  Introduction to Health Physics.  International Series of
  Monographs in Nuclear Energy, Volume 105.  Pergamon Press,  New York.

Eisenbud, K.  1973.   Environmenta1 Radi pact i vity, 2nd ed.   Academic Press,
  New York.

Fitzgerald, J. J.  1969.  Applied Radiation Protection and Control, Vol. I and
  II. Gordon and Breach Science Publishers, New York.

Fitzgerald, J. J.,  G. L. Browne!! and F. J. Mahoney.  1967.  Mathematical
  Theory of Radiation Dosimetry.  Gordon and Breach Science Publisners,
  NewYork."

Gloyna, E. F., and J. 0. Ledbetter.  1969.  Principles of Radiological Health.
  Marce! Dekker, Inc., New York.

Hall, E. J.  1978.   Radiobiology for the Radiologist, 2nd ed.  Harper and Row,
  Hagerstown,  Maryland.

Hentiee, W. R.   1970.  Medical Radiation Physics.  Year Book Medical
  Publishers,  Inc.,  Chicago, Illinois.

Henry, H. F.  1969.   Fundamentals of Radiation Protection.  Wiley-Interscience,
  New York.

Hine, J. G., and G.  L. Browne!!, eds.  1956.  Radiation Dosimetry.  Academic
  Press, New York.

Johns, H. E.,  and J. R. Cunningham.  1974.  The Phys-ics of Radiology. 3rd ed.
  Charles C. Thomas, Springfield, Illinois.
                                     A-1A

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                                                              DARCOM-P 385-1

Knoll, G. F.  1979.  Radiation Detection and Measurement.   John  Wiley and Sons,
  New York.

Lapp, R. E., and H. L. Andrews.  1972.   Nuclear Radiation  Physics, 4th ed.
  Prentice-Hall, Englewood Cliffs, New Jersey.

Lederer, C.M., and V. S. Shirley, eds.   1978.   Table of Isotopes.  John Wiley
  and Sons, New York.

Moe, J. J., S. R. Lusuk, K. C. Shumacher and H. M.  Hunt.   1972.   Radiation
  S_£Jety Technician Training Course.  ANL-7291, Rev. 1, Argonne  National
  Leooratory, Argonne,111inois.

Morgan, K. Z. , and J. F  Turner, eds.  1967.  Principles  of Radiation Protec-
  tion.  John Wiley and Sons, New York.

Norwood, W. D.  1975.  Health Protection of Radiation Workers.  Charles C.
  Thomas, Springfield, Illinois.

Radiological Health Handbook.  1970.  U.S.  Department of Health, Education and
  Welfare, Bureau of Radiological Health.   U.S. Government Printing Office,
  Washington, D.C.

Shapiro, J.  1972.  Radiation Protection.   Harvard University Press, Cambridge,
  Massachusetts.

Shapiro, J.   1981.  Radiation Protection,  A Guide for Scientists and
  Physicians.  Harvard University Press, Cambridge, Massachusetts.


                         INSTRUMENTATION AND DOSIMETRY
American National Standards Institute.  1972.  Criteria for Film Badge
  Performance.  ANSI N13.7, New York.

American National Standards Institute.  1975.  Performance, Testing, and
  Procedural Specifications for Thermo!uminescent Dosimetry:  Environmental
  Applications.  ANSI N5.45, New York.

American National Standards Institute.  1978.  Radiation Protection  Instru-
  mentation Test and Calibration.  ANSI N323, The Institute of Electrical  and
  Electronics Engineers, Inc., New York.

Apt, K. E., and K. J. Schiager.  1975.  "A Passive Environmental Neutron
  Dosimeter."  Health Phys. 28:474.

Becker, K.   1972.  "The Future of Personnel Dosimetry."  .Hes'.ih Thys. £3:729.

Boyns, P.  K.  1976.   The Aerial Radiological Measuring System (ARMS)—Systems.
  Procedures and Sensitivity.   EGG 1183-1691, EG4.G Energy Measurements  Group,
  Las Vegas, Nevada.

                                     A-15

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DARCOM-P 385-1

Brackenbush, L. W., et al.  1980.  Personnel Neutron Dosimetry at Department
  of Energy Facilities.  PNL-3213, Pacific Northwest Laboratory, Ricniand,
  Washington.

Brodsky, A.  1969.  "Personnel Dosimetry," in Handbook of Radioactive Nuclides,
  Y. Wang, ed.  Chemical Rubber Company, Clevelana, Ohio.

Budnitz, R. J.  1974.   "Radon-222 and Its Daughters - A Review of Instrumenta-
  tion for Occupational and Environmental Monitoring."  Health Phys. 26:145.

Budnitz, R. J.  1974.   "Tritium Instrumentation for Environmental and
  Occupational Moni'toring-A Review."  Health Phys.  2_6:165.

Dudley, R. A.  1966.  "Dosimetry with Photographic Emulsions," in Radiation
  Dosimetry, Vol. lid, F. H. Attix, W. C. Roesch, and E. E. Tochilin, eds.
  Academic Press, New York.

Eichholz, G. G., and J. W. Poston.  1979.  Principles of Nuclear Radiation
  Detection.  Ann Arbor Science Publishers, Inc., Ann Arbor, Michigan.

European Nuclear Energy Agency.  1963.  Proceedings of a Symposium on
  Personnel Dosimetry Techniques for External Radiation.  McGraw-Hill,  Inc.,
  New York.

Fitzgerald, J. J.  1969.  "Instruments for Radiation Detection and Measure-
  ment," in Applied Radiation Protection and Control, Vol.   I.  Gordon and
  Breach Science Publishers, New York.

Gessell, T. F., G. de P. Burke and K. Becker.  1976.  "An  International
  Intercomparison of Environmental Dosimeters."  Health  Phys. J0:125.

Gibbs, W. D., and C. C. Lushbaugh.  1969.  "Whole-Body Counter System," in
  Handbook of Radioactive Nuclides, Y. Wang, ed.  Chemical   Rubber Company,
  Cleveland, Ohio.

Hankins, D. E.  1973.   "Progress in Personnel Neutron Dosimetry," in
  Proceedings of the 3rd International Congress of the  International  Radiation
  Protection Association."CONF-730907, National Tecnnical  Information
  Service, Springfield, Virginia.

Hart, J. C.  1972.  "Legal and Administrative Aspects of Personnel  Dosimetry."
  Health Phys. 23:343.

Howard, L. E. Jr., J.  H. Spickard and M. Wilhelmsen.  1971.   "A  Human
  Radioactivity Counter and Medical Van."  Health  Phys.  £1.:417.

Hoy, J. E.  1972.  "An Albedo-Type Personnel Neutron Dosimeter."  Health  Phys.
  24:385.

International Atomic Energy Agency.  1970.  Nuclear  Accident  Dosimetry
  Systems.  IAEA Publication No. STI/PUB/241, Vienna.
                                     A-16

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                                                              DAKCOM-P 385-1

International Atomic Energy Agency (IAEA).   1970.   Personnel Dos'tietry Systems
  for External Radiation Exposures.   IAEA Publication No.  109, Vienna.

Joanes, A. P.  1975.  "A Personal Contamination Monitor Using Large Area
  Geiger Counters."  Health Phys. 28:521.

Kocher, L. F., L. L. Nichols, G.  W.  R.  Endres, D.  V.  Shipler and
  A. J. Haverfield.  1973.  "The Hanford Thermoluminescent Multipurpose
  Dosimeter."  Health Phys. 2J5:567.

Knoll, G. F.  1979.  Radiation Detection and Measurements.  John Wiley and
  Sons, New York.

National Council  on Radiation Protection and Measurements.  1978.
  Instrumentation and Monitoring Methods for Radiation Protection.  NCRP 57.
  Washington, D.C.

Oshino, M.  1973.  "Response of NTA Personnel Neutron Monitoring Film Worn  on
  Human Phantom."  Health Phys. 24:71.

Price, W. J.  1964.  Nuclear Radiation Detection.   2nd ed.  McGraw-Hill  Inc.,
  New York.
Rich, B. L., end B. G. Samardzich.  1973.  "Organizational Aspects of a  Per-
  sonnel Dosimetry Service."  Health Phys.  _24:95.

Sommers, J. F.  1975.  "Sensitivity of the G-M and Ion Chamber Beta-Gamma
  Survey Instruments."  Health Phys. 28, 755.

Technical Education Research Center-S.W.  1980.  Course II, Radiation Detec-
  tion and Measurements.  Waco, Texas.

U.S. Department of Defense, Dense Civil Preparedness Agency.   1977.   Handbook
  for Aerial Radiological Monitors.   CPG-2-6.2.3, Washington,  D.C.

U.S. Nuclear Regulatory Commission (NRC).  1973.  "Film Badge  Performance
  Criteria."  Regulatory Guide 8.3,  Washington, D.C.
                       QUALITY ASSURANCE/QUALITY CONTROL


American National Standards Institute.  1973.  Quality Assurance  Terms  and
  Definitions.  ANSI N45.2.10, New York.

American National Standards Institute.  1974.  Quality Assurance  Requirements
  for the Design of Nuclear Power Plants.  ANSI N45.2.11,  New  York.

American National Standards Institute.  1976.  Administration  Controls  and
  Quality for Operational Place of Nuclear Power Plants.   ANSI  N1S.7,
  New York.
                                     A-17

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DAKCOM-P 385-1

American National Standards Institute and American Society of Mechanical
  Engineers.  1979.   Quality Assurance Prograrr. Requirements for Nuclear
  Facilities.  ANSI/ASME N45.2, New York.

International Atomic Energy Agency.  1979.  Quality Assurance in the Procure-
  ment of Items and Services for Nuclear Power Plants.   IAEA Safety Series No.
  50-SG-QA3, Vienna.

International Atomic Energy Agency.  1979.  Quality Assurance Records Systems.
  IAEA Safety Series No. 50-SG-QA2, Vienna.

U.S. Atomic Energy 'Commission, Division of Reactor Research and Development.
  1973.  Quality Assurance Program Requirements.  RDT Standard F2-2,
  Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 10, Part 50, Appendix B,
  "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing
  Plants."  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1974.  "Quality Assurance Terms and
  Definitions."  Regulatory Guide 1.74, Washington, D.C.

U.S. Nuclear Regulatory Commission.  1975.  "Quality Assurance Requirements
  for the Design of Nuclear Power Plants."  Regulatory Guide 1.64, Rev. 1,
  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1979.  "Quality Assurance for
  Radiological Monitoring Programs (Normal Operations) - Effluent Streams and
  the Environment."-  Regulatory Guide 4.15, Rev. 1, Washington, D.C.

U.S. Nuclear Regulatory Commission.  1979.  "Quality Assurance Program
  Requirements (Operation)."  Proposed Revision 3 of Regulatory Guide 1.33,
  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1980.  "Auditing of Quality Assurance
  Programs for Nuclear Power Plants."  Regulatory Guide 1.144, Rev.  1,
  Washington, D.C.


                         RADIATION-GENERATING DEVICES
American National Standards Institute.  1967.  Immediate Evacuation
  Signal for Use in Industrial Installations Where Radiation Exposure  May
  Occur.  ANSI N2.3, New York.

American National Standards Institute.  1975.  General Safety Standard
  for Installation? Using Non-Mec'-ical X-Ray and Sesi&d &amm£-r,fcy Sources,
  Energies up to 10 MeV.  ANSI N543, also published as National Bureau of
  Standards Handoook No. 114,  Washington D.C.
                                     A-18

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                                                              DARCOM-P  3E5-1
American National  Standards Institute.   1976.  -Radiological  Safety Standards
  for the Design of Radiographic and nuoroscopic  industrial  X-Rey Equipment.
  ANSI N537, New York.

American National  Standards Institute.   1977.   Radiation Safety for X-Rey Dif-
  fraction and Fluorescence Analysis Equipment.   ANSI  No.  N43.2, also publisned
  as National Bureau of Standards Handoook No.  Ill, Washington, D.C.

Devanney, J. A., and C.  J.  Daniels.   1976.  "Radiation Leakage from Electron
  Microscopes."  Health Phys.  20:231.

International Atomic Energy Conimission.   1979.   Radiological  Safety Aspects
  of the Operation of Electron Linear Accelerators.  Vienna.

National Council on Radiation  Protection and Measurements.  1968.  Medical
  X-Ray and Gamma-Ray Protection for Energies up to 10 MeV - Equipment Design
  end Use.  KCRP 33,  Washington D.C.
                            RADIOCHEMICAL ANALYSIS
International Atomic Energy Agency.   1966.   Quick Methods for Radiochemical
  Analysis.   IAEA Technical Report No.  95,  Vienna.
                                    RECORDS
American National Standards Institute.  1972.  Practice for Occupational
  Radiation Exposure Records Systems.  ANSI N13.6-1966, Rev. 1972, New York.

Boiter, H.  P.   1976.  "Radiation Exposure Records Management,"  in Proceedings
  of the 9th Midyear Topical Symposium of -the Health Physics Society on
  Operational  Health Physics.   Rocky Mountain Chapter, Health Physics Society,
  P.O. Box  3229, Boulder, Colorado 80303.

Eastman Kodak  Company.   1978.   Storage and Preservation of Microfilms.   Kodak
  Pamphlet  D-31, Rochester, New York.

Matterazzi, A.  R.  1978.  Archival Stability of Microfilm - A Technical  Review.
  Technical Report No.  18, U.S. Government Printing Office, Washington,  D.C.

U.S.  Department of Energy.  1980.  "Micrographics Management."   DOE
  Order 1300.1, Washington, D.C.
                                     A-19

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DARCOM-P 385-1

                                   SHIELDING
Blizard, E. P., and L. S. Abbot, eds.   1962.   Reactor Handbook,  2nd ed., Vol.
  Ill, Part B, "Shielding."  Interscience Publishers, Inc.,  New  York.

Borak, T. B.  1975.  "A Simple Approach to Calculating Gamma-Ray Skyshine for
  Reduced Shielding Applications."  Health Phys.  2_9:423.

Bozyap, 0., and L. R. Day.  1975.  "Attenuation of 15 MeV Neutrons in Shields
  of Concrete and Paraffin Wax."  Health Phys.  28_:101.

Burson, Z.  G., and A. E. Profio.  1975.  Structure Shielding from Cloud and
  Fallout Gamma Ray Sources for Assessing tne Consequences of Reactor Acci-
  dents.  EGG-1183-1670, EG&G, Inc., Las Vegas, Nevada.

DeAlmeida,  C. E., S.  A. Rosanky, J.  R.  Marbach and P. R.  Almond.  1975.
  "Transmission in Concrete and Scatter Angular Distribution of  25-MV X Rays
  from a Betatron and a Linear Accelerator."   Health Phys. 28:771.

Jaeger, R.  G., ed.  1968.  Engineering  Compendium on Radiation Shielding -
  Fundamentals and Methods, Vol. I.Springer-Verlag, New York.

Jaeger, R.  G., ed.  1970.  Engineering  Compendium on Radiation Shielding -
  Shielding Design and Engineering,  Vol. III.Springer-Verlag,  New York.

Jaeger, R.  G., ed.  1975.  Engineering  Compendium on Radiation Shielding -
  Shielding Materials, Vol. II.  Springer-Verlag, New York.

Jones, T. D., and F.  F. Haywood.  1975.  "Transmission of Photons Through
  Common Shielding Media."  Health Phys. 28:630.

National Council  on Radiation Protection and Measurements.  1976.  Structural
  Shielding Design and Evaluation for Medical Use of X-Rays and Gamms Rays of
  Energies Up to 10 MeV^NCRP 49, Washington, D.C.

Patterson, H. W.,  and R. H. Thomas.  1973.  "Accelerator Shielding," in
  Accelerator Health Physics.  Academic Press, New York.

Price, B.  T., C. C. Horton and K. T. Spinney.  1957.  Radiation Shielding.
  Pergamon Press,  New York.

Rockwell,  T., Ill, ed.  1956.  Reactor Shielding Design Manual, 1st ed.  Van
  Nostrand, Princeton.

Schaeffer, N. M.,  ed.  1973.  Reactor Shielding for Nuclear Engineering,
  TID-25951,  National Technical Information Service, Springfield, Virginia.

Trout, E.  D., 0. P. Kelley and G. L. Herbert.  1975.  "X-Ray Attenuation in
  Steel-50 to 300  kVp."  Health Phys. 29:163.
                                     A-20

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                                                                CO.V-P 385-1

U.S.  Nuclear Regulatory Commission.   1975.   "Concrete  Radiation  Shields  for
  Nuclesr Power Plants."   Regulatory Guide  1.69.   Washington,  D.C.
                                  STATISTICS


Ostle, B., and R.  W.  Mensing.   1975.   Statistics  in  Research.  3rd.  ed.   Iowa
  State University Press,  Iowa.

Speer, D.  R., and  D.  A.  Waite.   1976.   "Statistical  Distributions  as  Applied
  to Environmental Surveillance  Data," in  Proceedings  of  the  9th  Midyear
  Topical  Symposium of the Health Physics  Society on Operational  Health
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  Boulder, Colorado 80303.
                           SURVEYING AND MONITORING
American National  Standards Institute.   1969.   Administrative Practices in
  Radiation Monitoring (A Guide for Management).   ANSI  N13.2, New York.

American National  Standards Institute.   1969.   Guute for Administrative
  Practices in Radiation Monitoring.   ANSI N13.2,  New York.

Diegl, H.   1972.   "Guidelines for Determining  Frequency of Wipe Surveys," in
  Health Physics  Operational  Monitoring, Vol.  I,  C.  A.  Willis and J. S. '
  Handloser,  eds.   Gordon and Breach Science Publishers, New York.

Feldman, A.  1976.   "Factors  and Strategies in Selection of Instruments for
  Radiation Surveys Around Medical  X-Ray Installations," in Proceedings of the
  9th Midyear Topical  Symposium of the Health  Physics Society oh "OperatipneT
  Health Physics.   Rocky Mountain Chapter, Health  Physics Society,
  P.O. Box 3229.  Boulder, Colorado 80303.

Fish, B. R.,  ed.   1964.   Surface Contamination.  Pergamon Press, New York.

Gaeta, N.  A., and  M.  H.  Repacholi.   1975.   "A Standard Survey Procedure for
  Photofluorographic X-Ray Machines."  Health  Phys.  28:763.

lacovino,  0.  M.   1976.   "Radiation Monitoring  Systems in a Changing Regulatory
  Environment,"  in  Proceedings of the 9th Midyear Topical Symposium of  the
  Health Physics  Society on Operations! Health Physics.Rocky Mountain
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International Atomic Energy Agency.  1979.  Radiological Surveillance  of
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  No. 49,   Vienna.
                                     A-21

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IJARCOM-P 385-1

International Atomic Energy Association.  Particle Size Analysis in Estimating
  the Significance of Airborne Contamination.   IAEA Safety Series No.  179,
  Vienna.

Katz, M.  1977.  Methods of Air Sampling and Analysis.   R. R.  Donnelley and
  Sons Company, Crawfordsvilie, Indiana.

McClelland, T. W. , and E. D. McFall.  1976.  "Radiation Monitoring Considera-
  tions for Radiobiology Facilities," in Proceedings of the 9th Midyear
  Topical Symposium of the Health Physics Society on Operational Health
  Physics.  Rocky Mountain Chapter, Health Physics Society, P.O. box 3229,
  Boulder, Colorado 80303.

Moghissi, A. A., and M. W. Carter.  1973.  Tritium.  Messenger Graphics,
  Phoenix, Arizona.

National Council on Radiation Protection and Measurements.  1976. Tritium
  Measurement Techniques.  NCRP 47, Washington, D.C.

National Council on Radiation Protection and Measurements.  1978. A Handbook
  of Radioactivity Measurement Procedures.  NCRP 58, Washington, D.C.

National Council on Radiation Protection and Measurements.  1978.  Instrumenta-
  tion and Monitoring Methods for Radiation Protection.   NCRP 57, Washington,
  D.C.

Noll, K. E., and T. L. Miller.  1977.  Air Monitoring Survey Design.  Ann Arbor
  Science Publishers Inc., Ann Arbor, Michigan.

Olson, 0. L.  1976.  "A Determination of Criteria for a Bioassay Program,"
  in Proceedings of the 9th Midyear Topical Symposium of  the Health Physics
  Society on Operational Health Physics.  Rocky Mountain  Chapter, Health
  Physics Society, P.O. Box 3229, Boulder, Colorado 80303.

Orvis, A. L.  1970.  "Whole Body Counting," in Medical Radionucl ides:
  Radiation Dose and Effects.  USAEC Symposium Series No.  20, CONF-691212,
  National Tecnnical Information Service, Springfield, Virginia.

Unruh, C. M.  1970.  "Radiation Protection Practices for  Tritium - A Manual  of
  Good Practice."  BNWL-SA-3390, Pacific Northwest  Laboratory,  Richland,
  Washington.

U.S. Nuclear Regulatory Commission.  1975.  "Acceptable Concepts, Models,
  Equations and Assumptions for a Bioassay Program.J  Regulatory Guide  8.9.
  Washington, D.C.

Wade, J. E., and G. E. Cunningham.  1967.  Radiation Monitoring:  A  Programmed
  Instruction Book.  Division of Technical Information, Oak Riage National
  Laooratory, Oak Ridge, Tennessee.
                                     A-22

-------
                                                              DARCOX-P 385-1

                                   TRAINING


American National Standards Institute (ANSI).   American National Standard for
  Selection and Training of Nuclear Power Plant Personnel.   ANS1/ANS-3.1,
  American Nuclear Society, LaGrange Park, Illinois.

American National Standards Institute.   1979.   Proposed American National
  Standard for the Qualification and Training  of Personnel  For Nuclear Power
  Plants.  ANSI/ANS-3.1, American Nuclear Society, La&range Park, Illinois.

American National Standards Institute.   1980.   Practices for Respiratory
  Protection.  ANSI-Z88.2, New York.

American Nuclear Society (ANS).  1976.   Proposed American National Standard
  for the Qualification and Training of Personnel for Research Reactors.
  ANS-15.A, LaGrange Park, Illinois.

Brown, B.  1971.  "A New 2-Year Part-Time Modular MSc Source in Health Physics
  at the University of Salford."  Health Phys. 20:663.

International Atomic Energy Agency.  1964.  Training in Radiological
  Protection:  Curricula and Programming.  IAEA Technical Report No. 31,
  Vienna.

Medina, L. C., -W. D. Kittinger and R. M. Vogel.  1976.  "Radiation Monitor
  Training Program at Rocky Flats," in Proceedings of the 9th Midyear Topical
  Symposium of the Health Physics Society on Operational Health Physics.
  Rocky Mountain Chapter, Health Physics Society, P.O. Box 3229, Boulder,
  Colorado 80303.

National Council on Radiation Protection and Measurements (NCRP).  1978.
  Radiation Safety Training Criteria for Industrial Radiography.  NCRP 61,
  Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 10, Part 19, "Notices,
  Instructions ana Reports to Workers;  Inspections."  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1979.  Information Relevant to  Ensuring
  That Occupational Radiation Exposures at Nuclear Power Stations Will Be  As
  Low As Reasonably Achievable (ALARA)."  Regulatory Guide 8.8,
  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1980.  "Personnel Qualification  and
  Training."  Second Proposed Revision to Regulatory Guide 1.8,
  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1981.  "Instruction Concerning  Risks  From
  Occupational  Radiation Exposure."  Regulatory Guide 8.29, Washington,  C.C.
                                      A-23

-------
DARCOM-P 385-1

U.S. Nuclear Regulatory Commission.   1981.   "Radiation Protection Training for
  Personnel At Light-Water-Cooled Nuclear Power Plants."  Regulatory
  Guide 8.27, Washington, D.C.

Vetter, R. J., and P. L. Ziemer.   1976.   "Operational  Training  in the Health
  Physics Curriculum," in Proceedings of the 9th Midyear Topical  Symposium of
  the Health Physics Society on Operational  Health Physics!Rocky Mountain
  Chapter, healtn Physics Society, P.O.  Box  3229, Boulder, Colorado 80303.

Williams, S. L.  1976.  "Meaningful  Radiation Worker Training for Temporary
  Craftsmen," in Proceedings of the  9th  Midyear Topical  Symposium of the
  Health Physics Society on Operational  Health Physics.RocKy  Mountain
  Chapter, Healtn Physics Society, P.O.  Box  3229, Boulder, Colorado 80303.
                                TRANSPORTATION
American National Standards Institute.   1973.  Administrative Guide for
  Packaging and Transporting Radioactive Materials.  ANSI N14.10.1, New York.

American National Standards Institute.   1975.  Administrative Guide for
  Verifyina Compliance with Packaainq Reauirements for Shipments of Radioac-
  tlve Materials.  ANSI N14.10.3, New York.

International Atomic Energy Agency.  1979.  Regulations for the Safe Transport
  of Radioactive Materials.  IAEA Safety Series No. 6, Vienna.

U.S. Code of Federal Regulations.  1982.  Title 49, Parts 100-199.
  Washington, D.C.

U.S. Code of Federal Regulations.  1982.  Title 10, Part 71, "Packaging of
  Radioactive Materials for Transport and Transportation of Radioactive
  Materials Under Certain Conditions."   Washington, D.C.

U.S. Department of Transportation.  1977.  A Review of the Department of
  Transportation (DOT) Regulations for Transportation of Raoioactive
  Materials.  Washington, D.C.

U.S. Nuclear Regulatory Commission.  1975.  "Administrative Guide for
  Obtaining Exemptions from Certain NRC Requirements over Radioactive
  Materials Shipment."  Regulatory Guide 7.5, Washington, D.C.

U.S. Nuclear Regulatory Commission.  1975.  "Leakage Tests on Packages for
  Shipment of Radioactive Materials."  Regulatory Guide 7.4, Washington, D.C.

U.S. Nuclear Regulatory Commission.
                                     1975.  "Procedures for Picking Up and
  Receiving Packages of Radioactive Material."  Regulatory Guide  7.3,
  Washington, D.C.
                                     A-24

-------
                                                              DARCOM-P 385-1
U.S. Nuclear Regulatory Commission.  1977.  "Administrative Guide for Verify-
  ing Compliance with Packaging Requirements for Shipment of Radioactive
  Materials."  Regulatory Guide 7.7, Washington, D.C.
                               WASTE MONITORING


Bradley, F. J.  1969.  "Radioactive Waste Disposal," in Handbook of
  Radioactive Mud ides, Y. Wang, ed.  Chemical Rubber Company, Cleveland,
  Ohio.

Cardozo, R. L.  1973.  "The Dispersal of Radioactive Matter by Evaporation."
  Health Phys. 25:593.

Gallagher, F. E., III.  1976.  "A New Facility for Processing and Storage  of
  Radioactive and Toxic Chemical Waste," in Proceedings of the 9th Midyear
  Topical Symposium of the Health Physics Society on Operational Health
  Physics.  Rocky Mountain Chapter, Health Physics Society, P.O. Box  3229,
  Boulder, Colorado 80303.

Gera, F.  1974.  "The Classification of Radioactive Waste."  Health Phys.
  27:113.

Gregory, W. D., and H. D. Maillie.  1975.  "Incinceration of Animal Radioactive
  Waste:  A Comparative Cost Analysis."  Health  Phys. 29:389.

International Atomic Energy Agency.  1965.  Management ofRadioactive Wastes
  Produced by Radioisotope Users.   IAEA Safety Series No. 12, Vienna.

International Atomic Energy Agency.  1966.  Management of Radioactive Wastes
  Produced by Radioactive Users-Technical Addendum.  lAtA Safety Series  No.
  19, Vienna.

Mawson, C. A.  1965.  Management of Radioactive  Waste.  Van Nostrand, New
  York.

Murphy, P. H., and N. S. Anderson.  1975.  "Inexpensive, Convenient  Xenon
  Disposal."  Health Phys. 213:779.

Port, E. A.  1975.  "An Improved Receptacle for  Radioactive Waste."   Health
  Phv_s. 29:801.

U.S. Nuclear Regulatory Commission.  1975.  "Evaluating and Reporting
  Radioactivity in Solid Wastes and Releases  of  Radioactive Materials in
  Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear  Power Plants."
  Regulatory Gu^.e 1.21, Washington, D.C.
* f.S. GOVIRWffiUT niKTIKS OTTlti: I984-44 1-740
                                     A-25

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us
    DC 20W
A Review of
The Department of Transportation
Regulations  for Transportation
of Radioactive Materials
                                          REV1SED 19E3

-------
This document is disseminated under the sponsorship
of the Department of Transoortetion in the interest
of information exchange.  The United States Govern-
ment assumes no liability for its contents or use
thereof.

-------
      A REVIEW OF THE DEPARTMENT OF TRANSPORTATION (DOT)

  REGULATIONS FOR TRANSPORTATION OF RADIOACTIVE MATERIALS
                                                 Rev. Summer 1983
For sole by the Superintendent of Documents, U. S. Government Printing Office,
                     Washington, D. C. 20402

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TO TOE READER:
Tne United States Department of Transportation(U.S.D.C.T.)  promotes
the safe transportation of hazardous materials by all modes.  One
evidence of this commitment to safety IE this review of the U.E.D.O.T.
regulations for the transportation of radioactive materials.  Compliance
with these regulations is a legal responsibility; more importantly,
radioactive materials packaged, labeled, marked and transported in
accord with these regulations have had an excellent safety record.

These regulations also require shippers, carriers, and manufacturers
of radioactive materials to train their staff.  They are required by
the Code of Federal Regulations, Title 49 (CFR 49)  to "instruct each
of his officers, agents, and employees having any responsibility for
preparing hazardous materials for shipment,  as to the applicable regula-
tions . "

It is our hope that this document will increase the safe transportation
of radioactive materials.  Any of the materials may be reproduced and/or
used in the training of your staff.

Comments, suggestions, corrections and/or requests for additional
training aids should be mailed to:  U.S. Department of Transportation,
Information Service Division, DKT-11, 400 Seventh Street, S.W., Wash-
ington, D.C. 2D59D.

Within U.S.D.O.T., the Material Transportation Bureau(KTB)  staff  in-
volved in the publication of this Review, include:

Sponsor and Approval:       L. D. Santman, Director
                            Dr. R. L. Paullin, Assoc. Director, Office
                              of Operations and Enforcernent (03E)
                            Alan I. Roberts, Assoc. Director, Office  of
                              Hazardous Materials Regulation (OHMR)

Technical Review:           Richard R.  Rawl, OHMR

Education Director:         Dr. Virginia T. Litres, OOE

Review Panel:
              OCE -         J. M. Shuler, F.  G.  Punch, R. McKinley

              CHMR -        Thomas Allan, Wendell  Carri3cer

              NRC -         Alfred Grella, U.S.Nuclear Regulatory
                               Oonmission, Office of Inspection and
                               Enforcement

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                           TABLE OF CONTENTS

                                                                  Pooe

     List of Tables and Illustrations                                    (iii)

     Title and Preamble                                                I


I.    BACKGROUND DISCUSSION                                        2

     A.   General                                                     2

     B.   Historical                                                   3


11.    SUMMARY OF RADIOACTIVE MATERIALS TRANSPORTATION
     REGULATIONS                                                   5

     A.   Organizations                                 -               5

     B.   Federal Regulations                                          6

     C.   International Regulations                                      9

     D.   Other Sources of Regulations and Tariffs                        10


III.   SUMMARY OF PRINCIPAL SHIPPER'S REQUIREMENTS IN
     PREPARATION AND OFFERING OF RADIOACTIVE MATERIALS
     FOR SHIPMENT                                                 11

     A.   Definition of Radioactive Material Subject to the Regulations      11

     B.   Best Approach to Using the Regulations                         12

     C.   Special Form Radioactive Materials                            12

     D.   Normal Form Radioactive Materials                            15

     E.   Quantity Limits and Packagings                                16

     F.   Limited Quantities, Instruments and Articles                    18

     G.   Low-Specific Activity (LSA) Materials                          21

     H.   Type A Packaging                                            2k

     1.    Type B Packaging                                            26

     J.    Fissile Radioactive Materials                                 .27

     K.   Highway Route Controlled Quantities                           28

                                    (i)

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                       TABLE OF CONTENTS (Cont.)

                                                                 Poge

     L.    Control of Radiation During Transport - Transport
          Index (T.I.), Vehicle Limits, and Separation Distances            29

     M.   Warning Labels                                            31

     N.   Contamination Control                                     3^

     0.   Other Shipper Requirements                                 35

          I.   Package Markings                                    35
          2.   Shipping Papers                                       37
          3.   Shipper's Certification                                 43
          4.   Security Seal                                         M
          5.   Small Dimension                                      45
          6.   Liquid Packaging Provision                             45
          7.   Surface  Temperature of Package                        45
          8.   Quality Control Requirements


IV.   CARRIER REQUIREMENTS IN HANDLING OF RADIOACTIVE
     MATERIALS PACKAGES                                        46

     A.   Shipping Papers and Certification by Shipper                   46

     B.   Placarding                                               47

     C.   Radiation Exposure Control by Maximum Total Transport
          Index vs. Distance                                         48

     D.   Reporting of Incidents                                      49

     E.   Notification to Pilot (for Aircraft Shipments)                  51


V.   MOST FREQUENTLY NOTED DISCREPANCIES IN RADIOACTIVE
     MATERIALS SHIPMENTS                                         51

     A.   By Shippers                                               51

     B.   By Carriers                                               55


VI.   IAEA REGULATIONS (AS AMENDED)                              57


VII.  DEFINITIONS                                                   58


VIIL  REFERENCES                                                  £3

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                         TABLES AND ILLUSTRATIONS

                                                                       Pope

                                    Tobies

Table 1     Sources of Federal Regulations                                   8

Table 2     Availability of International Regulations                          9

Table 2A   Other Sources of Regulations  and Tariffs                         I I

Table 3     Type A Package Quantity Limits for Selected Rodionuclides        18

Table 4     Activity Limits for Limited Quantities, Instruments and
           Articles                                                      20

Table 5     Radioactive Materials Packages Maximum Radiation Level
           Limitations                                                   2k

Table 6     Shipment Controls for Fissile  Radioactive Materials               28

Table 7     Radioactive Materials Packages Labeling Criteria                 33

Table 8     Removable External Radioactive Contamination Limits           35

Table 9     Most Commonly Used Shipping Names for Radioactive
           Materials                                                    ' 37


                                  Illustrations

Figure I    "What Must I Do to Moke a Safe and Legal Shipment
           of Radioactive Materials?"                                      I

Figure 2    NRC Agreement States as of  July I, 1983                         7

Figure 3    Special Form Radioactive Material                              13

Figure 4    Normal Form Radioactive Material                              15

Figure 5    Typical Type A Packaging Schemes                              25

Figure 6    Typical Type B Packaging Schemes                              26

Figure 7    Package Labels                                                32

Figure 8    A Word of Caution on Terminology                              34

Figure 9    Vehicle Placards for Radioactive Materials                       47

Figure 10   DOE Regional Coordinating Offices for Radiological
           Assistance                                                    50

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         A REVIEW QF THE DEPARTMENT Of TRANSPORTATION (DOT)

     REGULATIONS FOR TRANSPORTATION CF RADIOACTIVE MATERIALS
Figure 1 - "What Must I Do to Make a Safe and Legal Shipment
           of Radioactive Materials?"
   SPEC. COKUIHER ?
Perhaps some persons who have made shipments of radioactive materials may have

felt like the person illustrated above. The purpose of this Review is to summarize

the basic requirements of the Department of Transportation  (DOT) regulations

governing the packaging and shipment of radioactive materials.  It provides guidance

toward more correctly and easily applying those regulations in actual practice.



This Review is a reference and guidance  type document for training  purposes. It is

not intended to be an official interpretation  or restatement of the regulations. Any

material herein may be reproduced without special permission from the Department.
Users of this Review are strongly encouraged to obtain from the Government Printing.

Office the latest copy of the DOT Hazardous Materials Regulations in A? CFR Parts

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 100-177 ond I7B-I75.  Amendments  to the regulations ore published  in the Federal



 Register.








 I.    BACKGROUND DISCUSSION








      A.   General








           Since the beginning of the atomic energy industry about forty years ago,



 there has been an excellent record of safety.  Approximately 2,500,000 packages of



 radioactive materials are shipped per year in the United States.  Thus far, based on



 the best available  information  concerning  accidents  in  transportation,  there have



 been  no known  deaths  or  serious injuries to  the  public  or  to  the  transportation



 industry personnel as a result of the  radioactive nature  of  any radioactive material



 involved in an accident.  This  claim can generally be attributed to the  close attention



 given by the shippers  to the  proper  packaging  of radioactive  materials, and  to the



 effectiveness of the transportation safety standards and regulations.








           Even  with  this excellent  past record  of safety, the  term "radioactive"



 makes many people concerned or fearful. Still very vivid in the minds of many is the



 memory of devastating destruction and  violent deaths caused  by the atomic bomb



 near the end  of World  War  II.  The more recent debate  regarding  the effects of



 nuclear power plants on the environment and the ecology have added to this concern.








           The vast majority  of shipments of radioactive materials involves small or



 intermediate quantities  of  material  in relatively  small  packages.   Many of these



packages involve rodioisotopes which are intended for medical diagnostic or thera-



peutic applications.  They are used by thousands of doctors and hospitals throughout

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 the  United States and abroad.  Many such materials are  quite  often  of very  short



 "half-life." Therefore, for reasons of safety and economy, they  are  supplied by the



 producer to the user by the mosl rapid means of transportation.  Thus, the majority



 of these packages are shipped air  freight or air express  on passenger-carry ing  or



 cargo aircraft  or by other  rapid  delivery services.   Such shipments  are  often



 transported by both aircraft and motor vehicle.








           Other  uses for radioactive  materials include  industrial  applications  for



 inspection and "gauging" operations such as examining the integrity of welded  joints



 or in measuring the  thickness of paper as it is  produced.   Each year there are new



 applications  of nuclear  technology.   These  necessarily  involve the  shipment  of



 radioactive materials.  DOT updates the transport regulations to keep  pace with the



 changing transportation scene and to maintain the existing safety record.








      B.    Historical








           In  the  early  I950's,  the Interstate  Commerce Commission (ICC)) first



established regulations governing  the shipment of  radioactive  materials.    These



regulations were intended to protect property and individuals from excessive  expo-



sure to radiation resulting from the transport of radioactive  materials.  These early



regulations were designed to protect radiation-sensitive cargo, such as photographic



film in the mall, which might  also be transported in proximity to those radioactive



materials packages.   The  ICC established limits on radiation levels  that emanate



from packages.    By protecting  such radiation-sensitive  cargo,  there  was also



protection provided  to the people  who  transported it  or were passengers in the



vehicle or plane.

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           h 196!, the Internationa! Atomic Energy Agency (IAEA) adopted regula-



 tions for the transport of radioactive materials which were based on the ICC  rules.



 These IAEA regulations became the first international regulations for radioactive



 materials and the IAEA suggested that member states adopt the regulations as a basis



 for national requirements.








           In 1967,  o comprehensive  revision of the  IAEA  regulations took place.



 This  revision included consideration of a  new category of  materials—large radio-



 active sources.  This category was considered special because of the large amount of



 radioactivity  involved and  the  heat  that  might be generated by such  very  large



 sources.  The 1967 IAEA regulations served as the basis for a major revision of DOT



 regulations in 1968 governing radioactive materials.  Revisions that were adopted at



 that  time brought DOT'S regulations into essential conformity  with  the suggested



 international standards and, therefore,  harmony existed for  some period of time



 between the domestic and international requirements.  However, in 1973, the IAEA



 made  a  complete  revision of  its  regulations which  included  a new system  for



 classifying radionuclides,  known as  the  A.-A., system.  Elimination  of the "large-



 source" designation was possible because the special characteristics of large sources



 are  now considered routinely  for  all packages containing  greater  than o Type A



 quantity of radioactive material.








          The 1973 IAEA standards include the unilateral and multilateral concepts



 for Type B packages which determine the extent to which each country must approve



a package design.  In  March  1983, DOT adopted regulations which are in essential



conformity with the 1973 edition of the IAEA requirements.  This eliminated most



differences  that existed between international  and domestic regulations since  the



 1973 edition of the IAEA was published.  Although the revisions adopted by DOT are

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 in essential conformance with IAEA's 1973 standards, there are certain exceptions




 between the DOT and the IAEA requirements, which are noted in this Review.








 II.    SUMMARY  OF  RADIOACTIVE  MATERIALS  TRANSPORTATION RECU^A-
      A.   Organizations







           Under the Department of Transportation Act of  1966, the U. S.  Depart-



ment of Transportation has  regulatory responsibility for  safety in the transportation



of all hazardous materials,  including  radioactive  materials.  This includes shipments



by  all modes of transport  in  interstate  or  foreign commerce  (rail,  highway,  air,



water),  and by all means (truck, bus, auto, ocean vessel, airplane,  river barge, railcar,



etc.) except  for postal shipments.  Postal shipments come under the jurisdiction of



the U. S. Postal Service.








           The  ICC formerly  hod  jurisdiction over both  the  safety  and economic



aspects  of  the transport of radioactive materials  by surface modes.  The jurisdiction



over safety was transferred to  the Department of Transportation when it was formed



in April 1967.  However, the ICC still  exercises jurisdiction  over  the economic



aspects  of  radioactive materials transport through  the issuance of operating author-



ities to  carriers.








           Under the  Atomic  Energy Act of  1954, as  amended, the U. S. Nuclear



Regulatory Commission (NRC) also has responsibility for  safety in the possession, use



and transfer  (including transport) of by-product, source, and special  nuclear  mate-



rials.  Except for certain small quantities and specific products, a license is required



from the NRC for possession and use of such materials.  The NRC has promulgated,



in 10 CFR Part 71,  requirements which must be met for  licensees to deliver licensed



                                       5

-------
moteriol to a carrier for transport—If fissile material or quantities exceeding Type A

are involved.  The NRC also  assists  and  advises DOT in  the establishment  of both

national and international safety standards and  in  the  review and  evaluation  of

packaging  designs.   In  1979, NRC  adopted  by  reference  portions of  the DOT

regulations-  Now NRC  inspects its licensees for compliance with DOT regulations

applicable to shippers.



           Several states have entered  into  formal  agreements with  the NRC

whereby the regulatory  authority over  by-product,  source,  and  less than critical

quantities of special nuclear  material has been transferred to the states  from the

NRC.   These "Agreement States"  hove  adopted  uniform regulations pertaining to

intrastate  transportation of  radioactive  materials.  These  regulations require the

shipper to conform to the packaging,  labeling, placarding, and marking requirements

of the U. S. Department of Transportation. Additionally, many states have formally

adopted the DOT regulations and apply  these requirements  to both  intrastate and

interstate transportation.



     B.   Federal Regulations*



           The principal sources of Federal regulations pertaining to the transport of

radioactive materials are listed in Table  I. The regulations of the United States of

America are published by three agencies—the U. S. Department of Transportation,

the U. S. Nuclear Regulatory Commission, and the U. S. Postal Service  This Review

is concerned primarily with those regulations of the U. S. Department of Transporta-

tion, as published in Title 49,  Code of Federal  Regulations, Parts  100-177  and Parts
     Throughout this Review, the  Section references  listed apply  to  pertinent
     Sections in Title 49,  Code  of Federal  Regulations.  Parts  100-199,  unless
     otherwise specified.

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 178-199.   Persons  involved  as  actual  shippers,  package designers,  or  carriers are

advised 10 maintain a current copy of these regulations (all in two bound volumes).

The dated versions are published as of October  1st each year by the Superintendent of

Documents, U. S. Government Printing Office, Washington, D.  C. 20402.  As  of  1982,

the cost was  $9.00  per volume.   Changes to these regulations are published in the

daily Federal Register.  Regulatory changes in the form of amendments or notices of

proposed rulemoking are issued by the Materials  Transportation Bureau (MTB) of the

Department of Transportation.  Another means of keeping abreast of changes  is to

request,  in  writing,  to  be placed on the DOT mailing list for hazardous materials

amendments.   Send requests to:  U. S. Department  of  Transportation, Materials

Transportation Bureau, Information Services Division, DMT-1 I,  Washington,  D. C.

20590.
     Figure 2 - NRC Agreement States as of July 1,1983

                         AGREEMENT STATE PROGRAM
        j      | NON-AGREEMENT STATES
               (Alu Altiki, Hi*iii, drill Zone, District of Columbia, Puerto Rico.)

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                                  TABLE !

                    SOURCES OF FEDERAL  REGULATIONS
Title 49                                                                      !
'          U. S.  Department of Transportation's  Hazardous Materials  Regulations,1
;          Parts i 00-177 and 176-19?                                           '•
I                                                                             i
                                         Main Headings

          49 CFR 106  -   Rulemaking Procedures

          49 CFR 107 -    Hazardous Materials Program Procedures
                                                                             i
          49 CFR 171  -   General Information, Regulations and Definitions        i
                                                                             i
          49 CFR 172  -   Hazardous  Materials  Tables and  Hazardous  Materials!
                          Communications Regulations                          i
                                                                             i
          49 CFR !73  -   Shippers  - General  Requirements  for Shipments  and!
                          Pock agings

          49 CFR 174 -   Carriage by Rail

          49 CFR 175  -   Carriage by Aircraft

          49 CFR 176  -   Carriage by Vessel
                                                                             i
          49 CFR 177  -   Carriage by Public Highway                           j

          49 CFR 178  -   Shipping Container Specifications                     j
                                                                             i
          49 CFR 179  -   Specifications for Tank Cars                          !
Title 10
          U. S. Nuclear Regulatory Commission

          10 CFR 71    -   Packaging of Radioactive Materials  for Transport and
                          Transportation of Radioactive  Materials Under  Certain
                          Conditions
Title 39
          U. S. Postal Service

          Domestic Mail Manual, U. S. Postal Service Regulations, Part 124.  (Post-
          al Regulations for Transport of Radioactive Matter are published in U. S.
          Postal Service Publication ff6, and in the U. S. Postal Manual.)

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      C.    Internalionol Regulations



                                    TABLE 2

               AVAILABILITY OF INTERNATIONAL REGULATIONS
      "Regulations for the Safe Transportation of Rodiooctive Materials," as amend
      ed, Safety Series  //6,1973 Revised  Edition -  International  ATomic Energy!
      Agency (IAEA), Vienna, Austria.  Available from UN1PUB,  I ISO  Avenue of thti
      Americas, New York, New York  10038.

 2.    International Civil  Aviation Organization (ICAO), Technical Instructions for the
      Safe Transport of  Dangerous  Goods  by Air,  1983 Edition.  Available from
      INTEREG, P. 0. Box 60105, Chicago, Illinois 60660.                           j

 3.    International Maritime Organization  (IMO), formerly  Intergovernmental  Mari-
      time Consultative  Organization (IMCO).   International  Maritime Dangerous
      Goods (IMDG) Code.

 k.    International  Air Transport Asssociation (IATA),  Restricted Articles Regu-
      lations, 25th Edition, plus Supplement and Amendment issued  March  I,  1981,
      effective December  1982.  International Air Transport Association,  2000 Peel
      Street, Montreal, Quebec, Canada H3A 2RA.
           There are a number of international bodies  and organizations which deal

with the transportation of radioactive materials.  The majority of these international

bodies are  sanctioned by or affiliated  with the United  Nations.  These agencies

promulgate  regulations which  are recommended to  member states as a basis for

adoption  of  national regulations.   The  primary agency for the  promulgation of

radioactive materials transport standards is the International Atomic Energy Agency

(IAEA)  located  in Vienna, Austria.  The IAEA has been  the primary body for the

establishment of radiooctive  materials regulations which hove served as the basis of

all other international regulations and requirements.  In the air transport mode, the

International  Civil  Aviation Organization  (ICAO),  an  intergovernmental body,  is

active in regulating  the transport  of dangerous  materials,  including  radioactive

materials.  The  ICAO requirements  have been adopted by nearly all countries and

deal  with the air  carriage  of radioactive and  other hazardous  materials.   The

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International Air  Transport Association (IATA), a body of member air carriers, also



publishes regulations  for  air  transport  of  restricted articles  including  radioactive



materials.   In  the water  mode,  the  International  Maritime  Organization  (IMO),



formerly the International  Maritime  Consultative  Organization  (IMCO),  publishes



regulations which deal with the carriage of radioactive materials by vessel. The IMO



regulations and the ICAO regulations are based on the regulations of  the IAEA, but



are  more explicit  in  the  compliance  actions and requirements for shippers and



carriers.








      D.   Other  Sources of Regulations and Tariffs








           There  are  a number  of other agencies  or  organizations  which  publish



regulations or tariffs  on  the transportation of radioactive materials.  A word  of



caution  is in order  here.  A tariff is not  an official regulation.   These tariffs are



merely a publication by an organization or association which reprints certain Federal



or  international  regulations.   It  shows  the application  and  acceptance  of  those



regulations by the carriers who participate in the tariffs. As such, tariffs are binding



only on the  organization or association or  member carrier.   If a person chooses  to



utilize the widely used Bureau of Explosives (B of E 6000) as a source  of  regulations,



it is suggested that  the subscription be  for the quarterly updated version so that the



current requirements will be used. (See Table 2A)
                                        10

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                                 TABLE 2A

               OTHER SOURCES Or REGULATION'S AND TARirp$
 I.    "Official Air Transport Restricted Articles Tariff No. 6-D" and "Air Transport
      Restricted Articles Circular bO' -  Airline Tariff PuDiishing Co., Washington,
      D~C~:

 2.    "BOE 6000," Hazardous Materials Regulations of the Department of Transport©-
      tion,  including Specifications for Shipping Containers -  Bureau of Explosives,
      Association of American Railroads, Washington, D. C.

 3.    "ATA Hazardous Materials Tariff 111," Department of Transportation Reguio-
      tions  for Governing Transportation of Hazardous  Materials by Air, Motor, Rail,
      and Water,  Including Specifications for Shipping  Containers - American Truck-
      ing Associations, Inc., Washington, D. C.


 NOTE:     Refer to the current edition of the above references-
Ill.   SUMMARY OF PRINCIPAL SHIPPER'S REQUIREMENTS IN PREPARATION
     AND OFFERING OF RADIOACTIVE MATERIALS FOR SHIPMENT
     A.   Definition of Radioactive Materials Subject to the Regulations
              purposes of transportation, radioactive materials are defined as those

materials which spontaneously emit ionizing radiation and hove o specific activity in

excess of 0.002 microcuries per gram  of material.  All materials are to some degree

radioactive.  THE DEMARCATION OF 0.002 MICROCURIES PER GRAM ALLOWS A

DISTINCTION  BETWEEN  MATERIALS NOT  NORMALLY CONSIDERED  RADIO-

ACTIVE AND THOSE WHICH ARE REGULATED AS RADIOACTIVE IN TRANSPOR-

TATION. Materials with a specific activity lower  than 0.002 microcuries per gram

are not  regulated by DOT or IAEA.  These materials ARE NOT SUBJECT TO THE

RADIOACTIVE MATERIAL PROVISIONS  OF  THE DOT  REGULATIONS, however,

they may be subject to use or transfer regulations issued by the NRC or EPA.

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     B.    Best Approoch to Using the Rcgulotions



           A primary consideration for achievement of safety in  the  transportation

of radioactive  materials is the use of proper packaging for the specific  radioactive

material to be  transported.  In order  to determine the  packaging requirements, c

prospective shipper or package designer must answer ALL of mese  questions:



           I.    What  rodionuclides ore  being shipped?   Section  173.435 contains a
                listing of over 250 specific radionuciides.  Certain "ground rules" for
                dealing with unlisted or unknown radionuciides, or with  mixtures of
                radionuciides, appear in  Section 173.433.

           2.    Whet  quantity of the rodionuclides is being shipped?   As you will
                see, the packaging requirements are related to the quantity (activity
                of material).  They are generally  structured about the total quantity
                in a package in terms  of activity (curies,  millicuries,  and micro-
                curies).

           3.    What  is the form of the  rodionuclide?

                c.    is the material in special form?;  or

                b.    Is it in normal form?



           The terminology which has been introduced  in asking these  three questions

is explained in the succeeding paragraphs.



     C.   Special Form Radioactive Materials



           What is meant by "special form" radioactive  material? As illustrated in

Figure 3, "special form" materials are limited to materials which,  if released from a

package, might present a hazard of direct external radiation.  However,  due to their

"high physical integrity," they would present very  little hazard, if  any, as a result of

the spread of loose radioactive material  (contamination).  This high physical integrity

could be the result of a natural property of the material, such as its being in massive,

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nondispersoble solid form, or on ocquired characteristic,  such as being seated into o

very durable capsule (encapsulated).



           Special  form  encapsulations  must  hove a1  least  one  external physical

dimension  which exceeds  5mm.  This minimum  dimension requirements mokes the

capsule more easily seen and recovered in the event of an  accident/incident.



           Special form encapsulations are  required to be  so constructed that  they

can  only  be opened  by  destroying  the  capsule.   This  requirement  is  intended to

prevent the inadvertent loosening or opening of the capsule,  either during transport

or following  an accident.   The "special form" materials are  much less likely to spread

contamination in the  event of package failure.  Therefore,  the regulations generally

allow  substantially  larger  quantities of  such  materials   to be  placed  in  given

packogings than when the materials are in "normal form."
         Figure 3 - "Special Form" RA.M. (173.403 (z) and 173.469 (a))

         May Present a Direct Radiation Hazard if Released From Package, but
         Little Hazard Due to Contamination

         "Special Form" FLA.M. May Be "Natural" Characteristic,  i.e., Massive
         Solid Metal, or "Acquired" Through High Integrity Encapsulation
              Massive
              Solid Metal
High Integrity
Encapsulation
as a Sealed Source
                              Stainless Sloe
                              Outer Capsule
              High Integrity Weld

              Tantalum Inner
              Capsule

              Radioisotope

              Hiph integrity WelQs

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           For purposes of export, a shipper must furnish to the foreign consignee o

certificate of competent authority for the special form material.  Such  a certificate

will  onjy  be issued bv the DOT,  Office  of Hazardous  Materials Regulation, upon

receipt  of o specific petition and only when o certificate is required bv o shipper to

fulfill o need.   Such c need will  be in the case of foreign  shipments only,  such as

pursuant to  paragraph  803  of  the International   Atomic  Energy  Agency  (IAEA)

regulations.  Section 173.476, relating to certain special form requirements, is quoted

below:



           "Section 173.476    Approval of special form radioactive materials.

           (a)   Each  shipper of special form radioactive materials shall maintain on
           file for  at  least one year after the latest shipment, and  provide to the
           MTS on request, a  complete safety analysis, including documentation of
           any  tests,  demonstrating  that the  special  form  material meets the
           requirements of Section 173.469.  An  IAEA  Certificate of Competent
           Authority issued for the special form material may  be used to satisfy this
           requirement.

           (Approved  by the Office of Management and Budget  under OMB control
           number  2137-0516.)

           (b)   Prior  to the first export shipment of  c special  form radioactive
           material from  the  United States, each  shipper   shall  obtain  c U. S.
           Competent Authority Certificate for the specific material.  Each  petition
           shall be  submitted in accordance with Section 173.471(e) and must induce
           the following information:

                (I)   A detailed  descriotion  of the material,  or  if  a capsule,  a
                detailed description of the contents.  Particular reference  must be
                made  to both physical and chemical states:

                (2)   If a capsule is  to be used, a detailed statement of its design
                and dimensions, including complete engineering drawings and sched-
                ules of material, and methods of construction; and

                (3)   A statement of  the tests  that  have  been  made  and their
                results; evidence  based  on calculative  methods to show that  the
                material is able to pass the tests; or other evidence  that the special
                form  radioactive material complies with Section 173.46?.

           (c)   Paragraphs (a) and (b) of this section do  not  apply in those cases
           where A, equals A.,  and the material  is  not described or, the  shipping

           papers as "Radioactive Material Special Form, n.o.s."

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           In determining if on encapsulation passes the test requirements for special

form (Section 173.469), an option is available.  In addition to the leaching acceptance

test which is performed over a seven-day period, some capsules may be acceptance

tested by volumetric means.  If on encapsulation has on internal  void of at least O.I

ml  (.006 in ~ ), it may be leak tested using techniques such as the vacuum bubble or

helium  leak tests.  Any volumetric  leak testing technique may be  utilized, provided

that it has a sensitivity of ot least:
           (I)    1.3 x 10" atm - cm /s for solid contents; or

           (2)    1.3 x 10~ atm - cm /s for liquids or gaseous contents.
All tests are based on air at 77  and one atmosphere pressure differential.



     D.    Normal Form Radioactive Materials



           Illustrated in Figure 4 are "normal  form" radioactive materials which are,

therefore, ANY radioactive materials that do not qualify as "special form."
        Figure 4 - Normal Forms Radioactive Materials 49 CFR 173.403(s)

        Normal Form Materials May Be Solid, Liquid or Gaseous and Include any
        Material .Which Has Not Been Qualified as Special Form

        Type A Package Umlts are A2 Values
                   Waste Material in
                   Plastic Bag
                  Powder in Glass
                  or
                  Plastic Bottle
Liquid In Bottle Within
Metal Container
      Gas in Cylinder

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     E.    Quontitv Limits ond pockogings








           Having  considered  the  type,  guonlity,  ond  form of  the  radioactive




material,  it is now  appropriate to consider  the packaging reauirements.  packoamg




types are "Type A," "Type B,"  "excepted," and "strong,  tigh'," all of which will  DC




explained.








                            THE A, and A., SYSTEM
                            ——^— |       /







           The present  regulations use  A, and A-, values as points  of reference for




quantity   limitations  for  every  radionuciide.   This  system  replaces  the former




Transport Group system that was used for limitations when the radioactive materials



were in normal form.








           Every radionuciide  is now assigned an A, and an A^  value.  These two




values  (in curies) are simply the maximum activity of that roa'ionuclia'e that may be




transported in a TYPE A package.  Table 3 gives  examples of A,  and A-, values.for




typical  rodionuclic-    The A,  value   is the number  of  curies for o  particular




radionuciide when in  Special Form.  The  A? value  is the  number  of curies  if the



radionuciide is not in Special Form—i.e., the material is in Normal  Form.








           In previous regulations, the activity limitations for Special Form was the




same  for all  radionuclides.   Now  the limitation for  each  rodionuclide  depends




primarily on the  penetrating radiation emitted by the material when encapsulated.








           Under the former regulations, every radionuciide was  assigned  to  one of




seven transport  groups (I through VII).  The assignment of a radionuciide to o group

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was based on the radiation hazard that would occur if some of the material was token




into a person's body.  The activity limit for all rodionocTides in a transport group was




established by the rodiotoxicity cr  the most hazardous radionuciide  in the  group.




Under the present  system,  the A~ limit for  each radionuciide  is established  on  the




basis of  the hazard that would result if  that individual radionuciide  was ingested,




inhaled or absorbed through the skin.








           The Limited Quantity, Low Specific  Activity (LSA), Type A, Type B,  and




Highway Route Controlled Quantity provisions in the  regulations all relate to At  and




A-  values as points of  reference for activity limits or thresholds.  In  the  case of




Type B quantities, they are simply defined as a quantity exceeding the appropriate A.




or A- value for the radionuclide(s) of interest.








           For mixtures of radionuclides, certain rules are  specified for determining




whether  the Type A quantity has been exceeded (see Section  I73.£33(b)).  In most




cases,  the "ratio rule"  may be applied.  This involves dividing the activity of each




rodionuclide present by its Ai or A? value (as appropriate) and summing the resulting




fractions.   If the sum is  ! .0  or less, then the  mixture does not exceed  a  Type A




quantity.

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                                  TABLE 3
           A PACKAGE QUANTITY LIMITS FOR SELECTED RADIONUCLIPES
        (ADDITIONAL RAblONucUDgS ARE LISTED IN SECTION  173.435)
SYMBOL OF
^ADIONUCLIDE
'«C
l37Cs
?9Mo
235U
226Ra
201Pb
ELEMENT AND
ATOMIC NUMBER
Carbon (6)
Cesium (55)
Molybdenum (42)
Uranium (92)
Radium (88)
Lead (82)
A, (Cj)
(Special Form)
1000
30
100
100
10
20
A2 (Ci)
(Normal Form)
60
10
20
0.2
0.05
20
NOTE 1:  Quantities exceeding Type A package limits require Type B packaging.

NOTE 2:  Highway Route Controlled Quantities are defined in Section 173.403(1).




          "Type B Packages," "Highway Route Controlled Quantities," and "Fissile

Radioactive  Materials" present more unusual and specific problems for packaging and

carrier's operational  controls.  These  materials are additionally controlled by the

packaging standards as promulgated'by the Nuclear Regulatory Commission in  Title

IOCFRPart71.




     F.   Limited Quantities, Instruments and Articles
          The AI  and A^ values are also used as a basis for defining the package

quantity  limits  for limited quantities and  both  the item and  package limits for

instruments, as  illustrated in Table 4.  Packages  containing materials within these

quantity  limits are excepted from some of the requirements which apply to Type A

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pockoges.  These exceptions include not hoving to provide specification packaging,

shipping papers, certification, marking  or labeling.  However,  there are o number of

conditions  which the  limited  quantity, instrument  or  article  must meet.  They

include:
           1.    Activity  limits per package and, if  appropriate, per instrument or
                article;

           2.    The materials must be pocked in strong, tight packages that will not
                leak  ANY of  the  radioactive material during conditions normally
                incident to transportation;

           3.    The  radiation level at  any point on  the external surface of the
                package cannot exceed 0.5 millirem per hour;

           k.    The  external  surface of the  package must be free of significant
                removable contamination;

           5.    For instruments or articles, the radiation  level at 4 inches from any
                point on  the  surface of the unpackoged instrument or article may
                not exceed 10 millirem per hour; and

           6.    A prescribed description of the contents on o document which is in
                or on the package or forwarded with it.
                                       19

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                                   TABLE 4

   ACTIVITY LIMITS FOR LIMITED QUANTITIES, INSTRUMENTS AND ARTICLES
^oture of Content s-
iolids
Special form
Other forms
-iquids
Tritiated water
< O.I Ci/liter
O.I Ci to I.OCi/l
> 1.0 Ci/liter
Other liquids
jases
Tr ilium-
Special form
Other forms
Instruments and Articles
Instrument .,
and article limits—
IO'2A,
IO'2A2

-
IO"3A2
20 curies
ltr3A,
IO'3A2
Package limits
Al
A2

-
KT'AJ
200 curies
IO'2A,
IO"2A2
Materials
Package limits
KT3A,
IO'3A2

1,000 curies
1 00 curies
1 curie
IO'/
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      C.    Low Specific Activity (L.SA) Moteriols


            Low specific octivity  materials  ore  those  moteriols which  present o

 relatively low hazard as o result of their limited radioactive concentration.  Some of

 these materials are listed by name, such as uranium ores  and concentrates, as well as

 unirradiated natural or  depleted  uranium.   Other materials  must  meet  certain

 limitations related to their radioactive concentration.  For example, tritium oxide in

 aqueous solutions (tritiated  water) cannot  exceed 5.0 millicuries per milliUter.  The

 allowable  radioactive  concentration for other  materials  with uniformly  dispersed

 activity is related to the Aj values of the radionuclides present.  The relation is as

 follows:


           Jf the A2 of the                 The maximum activity PER GRAM

           rodionuclide is;                  of material is;

           not more than 0.05 curie         0.0001 milUcurie

           more than 0.05 to 1 .0 curie       0.005 millicurie

           over 1.0 curie                   0.3 millicurie


           When mixtures of radionuc fides are present, 'they must be subjected to the

"ratio"  rule to determine if the mixture is LSA.  For uniform mixtures of nuclides,

the following formula will determine if the  mixture is defined as LSA:
                           APG,   +   APG2
                           ZOTfi     C75DT
           Where:
                                       21

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           APG, =    the total  activity (in millicorics)  per gram of material of all



                     nuclides present with on A2 value of less than 0.05 curie.








           APG2 =    the total  activity (in millicories)  per gram of material of all



                     nuclides present with an A*  value of more than 0.05 but less



                     than 1 .0 curie.
                     the total  activity (in mil(icuries) per gram of material of all



                     nuclides present with an A* value exceeding 1.0 curie.
           If the above summation for a given uniform mixture is less than or equal



to 1, then the mixture may be classified as LSA.








           Because of their  low radioactive  concentration, these materials usually



can be safely carried without regard to the total activity of the  material  in a single



package.   Most  low-level  radioactive waste  shipments  are  comprised of  LSA



materials.  There are TWO WAYS in which LSA materials can be transported.







     o     Nonexclusive use shipments - "essentially Type A packages"







           The first method, "nonexclusive  use" tronsportatiorv,  requires  that  the



material be transported in essentially  a Type A package.  "Essentially o Type A"



package means a package that must survive the physical  tests, such as the drop and



compression tests for  Type A packages - but which is excepted from some of the



general  Type A  requirements.   The actual  test requirements are  found in Section



173.465. Although the packages are excepted from certain  design requirements, their



integrity must be equal to a Type A.
                                       22

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      o    Exclusive use - "strong, tight pockooe.*'








           LSA  materiols which  are  transported by  conveyances assigned for the



"exclusive use" of the consignor may be shipped in packages that are of less rigorou:



construction.  Users of the exclusive use provision MUST ENSURE  that there will be



no loading or unloading of the material except under the direction of the consignee or



consignor.  The  limitation on loading and unloading, plus the requirement  that the



material be in exclusive use,  safely allows the exception from certain packaging test



requirements. Exclusive use  LSA, therefore, is allowed to be mode  in the so-called



strong, tight package.








          There are no specific test requirements for the strong, tight packages.



However, a  performance criteria must be met—there  can be no release of radio-



active content  during  transportation and  like  any other package  of hazardous



material, the requirements  of Section 173.24  must be met.  Materials which are



consigned as exclusive use LSA shipments MUST  have the packages marked "Radio-



active LSA."  And the vehicle on which they are being transported MUST be placarded



with the RADIOACTIVE MATERIAL placard.
                                      23

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                               TABLE 5

                   RADIOACTIVE MATER IALS PACKAGES
                 MAXIMUM RADIATION LEVEL LIMITATIONS
                     (SEE SECTIONS 173.44 I (cO AND (b)
iRADIATION LEVEL (DOSE) RATE AT ANY POINT ON EXTERNAL SURFACE OF
ANY PACKAGE OF R.A.M. MAY NOT EXCEED:

     A.   200 MILLIREM PER HOUR.
     B.   10 MILLIREM PER HOUR AT ONE METER* (TRANSPORT INDEX MAY|
          NOT EXCEED 10).

UNLESS THE PACKAGES ARE TRANSPORTED IN AN "EXCLUSIVE USE" CLOSEDl
            VEHICLE (AIRCRAFT PROHIBITED) - THEN THE MAXIMUM RADIA-j
(TION LEVELS MAY BE:

     A.   1000 MILLIREM PER HOUR ON THE ACCESSIBLE EXTERNAL PACK-I
          AGE SURFACE.
     B.   200 MILLIREM PER HOUR AT EXTERNAL SURFACE OF THE VEHICLE.
     C.   10 MILLIREM PER  HOUR AT TWO  METERS** FROM  EXTERNAL!
          SURFACE OF THE VEHICLE.
     D.   2  MILLIREM PER HOUR IN ANY POSITION OF THE VEHICLE WHICH IS)
          OCCUPIED BY A PERSON.

     *    3.3 feet.
     **   6.6 feet.
     H.   Type A Pockoging



         In Figure 5, there is on illustration of "Typical Type A Packaging Schemes."

Type A packaging is that which must be designed in accordance with the applicable

general packaging -requirements as  prescribed in the regulations (Sections  173.24,

173.411,  173.412), and which must be adequate to prevent the loss or dispersal of its

radioactive contents and to maintain its radiation shielding properties if the package

is subjected  to normal conditions of transport.  The regulations prescribe (Section

173.465)  the performance criteria to simulate normal and rough handling conditions

of  transport.  Typically, the Type A packaging prescribed in the regulations  is the

performance-based  DOT Spec.  7A (Section  178.350) Type A general packaging for

which each shipper must make his own assessment and certification of the particular


                                  2k

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 packooe design against  the  performance requirements.  Tne regulatory  framework,

 therefore, provides  for the use  of Type A packaging  without  prior specific approval

 by DOT of  the package  designs vie  the use of DOT Spec. 7A performance specifica-

 tion.  Additionally, foreign-made Type A packages  are acceptable  internationally,

 provided they  are  so  marked as Type A  and comply with the requirements of  the

 country of  origin.   It should be noted that  the  shipper  of  each DOT  Spec. 7A is

 required to maintain on file for at  least  one year after the latest shipment, and be

 prepared to provide to  the Department, o  complete certification and supporting

 safety  analysis demonstrating that  the construction  methods, packaging design,  and

 materials of  construction are  in compliance with  the  specification (see  Section

 173.415). The  information in this file must show, through any of the methods given in

 Section  173.461, that all of the  requirements of Sections 173.24, 173.463  and  173.465

 are met.  The file must also relate the contents of the package(s) being shipped to the

 contents which were used for testing purposes.




          Figure 5 - Typical Type A Packaging

          Package Must Withstand Normal Conditions (173.465) of Transport
          Only Without Loss or Dispersal of the Radioactive Contents.
             Flberboard Box
Steel Drum
     Wooden Box

 Typical Schemes
Dot Specification 7A
 Type "A" Package
NOTE:
     A useful  reference  in  evaluating whether certain  DOT specification package
     designs meet the requirements of DOT Spec. 7A is listed  in  Section VIII,
     Reference 10.  It is the shipper's responsibility to ensure that all details  of the
     package which  is offered for transport  complies  with the  Spec.  7A require-
     ments.  This includes ensuring that the package evaluation is complete (Section
     I73.4l5(a)) and that  the packaging and contents offered for transport have been
     included in the evaluation.
                                       25

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     1.     Type B Pockoging



           Type B Packaging  (see  Figure 6),  must meet the general packaging  re-

quirements and all of the performance standards for Type A packages. In addition, it

must withstand certain serious accident  damage test  conditions.   After the tests,

there must  be only  limited loss of shielding  capability and essentially no  loss of

containment.   The performance criteria which the  package  designer  must use to

assess Type B packaging against these empirically established  hypothetical accident

test conditions of the transport are prescribed in the Nuclear Regulatory Commission

regulations (!0 CFR 71.73) and include the following:
           1.    A 30-foot free drop onto an unyielding surface.

           2.    A puncture test  which is a free drop (over 40 inches) onto a six-inch
                diameter steel pin.

           3.    Thermal exposure at 1,475 F for 30 minutes.

           4.    Water  immersion  for eight  hours (for  fissile materials packaging
                only).
   Figure 6 - Typical Type B Packagings

   Package Must Stand Both Normal (173.465) and Accident (10 CFR Part 71)
   Test-Conditions Without Loss of Contents.
                    IB Gauge Steel Drum or Outer Cover
                                          Inner
                                          Containment
                                          Vessel
  Steel Outer Drum
  Shielded Inner Container
  Thermal Insulation
  Between Containers

          3" Min.-AI! Around
          Top & Bottom
                                                            Exterior Grade 3/4"
                                                            Douglas Fir Plywood
                             Lag Screws
Inner
Containment
Vessel
Rods
                                       26

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           Except  for o  limited number of specificotion Type B pockogings (e.g.,



DOT-6M) described in the regulations, all  Type B package  designs  require  PRIOR



APPROVAL of the U. S.  Nuclear Regulatory Commission or Department of  Energy



(DOE).  (See Section  173.471 for standard requirements and  conditions pertaining to



NRC approved packages and Section 173.7 for DOE certified pacKoges..)








     J.    Fissile Radioactive Materials








           In  addition  to considerations for the  radioactive  content,  shippers of



fissile  radioactive material must also take into account certain other packaging and



shipment  requirements  to ensure  against  nuclear  criticality due  to the  fissile



(fissionable)  nature of the  materials.   The design  of the  packaging  for  fissile



radioactive material, the transport index to be assigned (if Fissile Class II), and any



special procedures for packaging are prescribed in 49 CFR 173.AS 1 through 173.459 of



the  DOT  regulations and  in 10  CFR 71 of the USNRC regulations.  Each fissile



radioactive materials package design (except for the DOT Spec. 6L, 6M, and Spec. 20



PF-1, 20 PF-2, 20 PF-3, and 21 PF-1 and 21 PF-2) must be reviewed  and approved by



the USNRC prior to  its first use.   The  packaging  must be  such to ensure  against



nuclear criticality  (an unplanned  nuclear  chain reaction)  under both  normal and



hypothetical accident test conditions, and prevent loss  of contents in transportation.



Fissile  radioactive  material  packages  are  classified into one of  three  groups,



according to  the degree of control  which must be  exercised to assure  nuclear



criticality safety, as shown in Table £.

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                                   TABLE 6


         SHIPMENT CONTROLS FOR FISSILE RADIOACTIVE MATERIALS
                               (SECTION 173.455)
 I.    Fissile  Class I  - Packages may be transported in unlimited numbers (Transport
      Index is based only on external radiation levels).

 Z.    Fissile  Class II  -  Number  of  packages limited by  aggregate maximum of
      transport indexes of 50  (50  unit rule).   No single  package  may  exceec a
      transport index of 10. Transport index shall be based on criticality or external
      radiation level  basis, whichever is most restrictive.

 3.    Fissile  Class III - Shipments of packages which do not meet the requirements of
      Fissile  Class I  or II.  Controlled by specific arrangements between the shipper
      and carrrier. (See Section I73.457(b).)
      K.   Highway Route Controlled Quantities



           Certain quantities of radioactive materials  known as "Highway Route

Controlled Quantities" are subject  to additional  controls during transportation.  A

Highway Route Controlled Quantity is defined as an amount of material in a single

package which exceeds either:  (I) 3,000  times  the A, quantity, for  special form

material; (2) 3,000 times the A2 quantity, for normal form materials;  or (3) 30,000

curies,  WHICHEVER IS LEAST.   Packages containing a Highway Route Controlled

Quantity of radioactive material are subject  to specific routing controls which apply

to the highway carrier.  The carrier must operate on preferred routes that are in

conformance with Section 177.825.  The carrier must report to the shipper the route

used  in making  the shipment.    The  shipper  is  required  to report  the  routing

information to the Materials Transportation Bureau (MTB) per Section I73.22(c).



           In determining if  a package contains  a Highway Route Controlled Quan-

tity of material, first identify the rodionuclide being transported.   After identifying
                                      28

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 the rodionociide, determine if it  is in special form or normal form.  If the material is
 in special  form, multiply the A.  value for the rodionuclibe by 3,000.  Compare this
 answer with  30,000  curies.  The lower of the two values  is  the Highway Route
 Controlled Quantity  for that  roriionuclide in special  form.   If  the  contents  of  the
 package being shipped exceeds the Highway Route Controlled Quantity,  the package
 must be transported under the specific route control requirement.   For  example,
 suppose  a  shipper  has a  package of Cobalt 60 in special  form.  The A,  value  for
 Cobalt 60 is seven  curies--? x 3,000 =  21,000 curies; 21,000 curies is less than 30,000
 curies.   Therefore, 21,000 is the Highway Route Controlled Quantity for Cobalt  60.
 Packages containing 21,000 or more curies of Cobalt 60 in special form are subject to
 specific  routing  controls.  Treat  other radionuclides similarly in deciding if  the
 package is  subject to specific routing controls.   Remember,  the Highway Route
 Controlled Quantity relates to the content of a package—not to the sum of contents
 of all packages in a shipment.
      L.    Control of Radiation During Transport -  Transport  Index CT.1.1. Vehicle
           Limits, and Separation Distances                                    "~~~"~
           The  regulations prescribe  that  the  maximum permissible dose rate for
packages of radioactive materials offered for transport shall not exceed 200 millirem
per hour at any  point on the external surface of the package, and the transport index
may not exceed 10.  The highest dose rate at one meter away from any  accessible
exterior lurfoee of the package  equals the "Transport Index," or T.I.  If the shipper
assures that the package will be transported  on a conveyance as "exclusive use," a
higher maximum dose rate is allowed.  The radiation level limitations are summarized
in Table 5.  (See page 24)

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           To  control the rodiotion level resulting  from accumulations of multiple



numbers of packages in the transportation environment,  the regulations  require that



the  carrier shall  maintain certain prescribed  separation  distances between  radio-



active materials packages and other areas occupied by  persons and/or photographic



film (since film max ^e fogged by radiation). For film,  these separation distances are



based  on the storage time  and the transportation  index; and for  separation from



people, the distances are based solely on the T.I. No package offered  for transport (on



other than "exclusive use" vehicles) may have o T.I. exceeding  10. However, the T.I.



per package limit is decreased to 3.0 for packages carried aboard passenger-carrying



aircraft.  T.l.'s of 10 and 3 are based on standards for  limiting personnel exposure,



and to prevent  "fogging" of "fast" photographic film.








           The total of the T.I. of all packages  in any single transport  vehicle or



storage location generally may not exceed 50. Exceeding the 50 T.I.  per  vehicle limit



is authorized only for certain specific types of shipments which are carried under the



special requirements of "exclusive use" vehicles  (Section 173.403(0), which impose



additional  responsibilities on the shipper.  The shipments that qualify most often  for



the 50 T.I. exception are "exclusive use" shipments of low specific  activity radio-



active materials (Sections I73.425(b) and (c)) or the occasional "hot" package that



cannot meet  the  10  T.I. limit (Section 173.44 Kb)).   In either case, the special



arrangements between the shipper and carrier must  satisfy all requirements, includ-



ing those of Sections 173.403(0 and  173.44 Kb).








           The regulations provide graded tables of stowage distances for  stowage in



accordance with  the cumulative tronsport index.  These  tables are found  in  the



carrier sections  of  the regulations (Sections 174.700,  175.701  through  175.703,



176.708, and 177.842).
                                        30

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           The transport index  (T.I.)  system, together  with tables of separation



 distances, provides control by the carrier over the  radiation exposures to personnel



 handling the packages, and to casually exposed persons in the vicinity  of accumula-




 tions of packages.







           The T.I. system is also designed to provide the means to assure critical!ty



 safety.  It limits the amount of fissile materials in one  location under nonexclusive



 use conditions (the 50 T.I. ^pper limit). This prevents conditions that would  support a



 nuclear "chain reaction," or "go critical."  For such fissile materials,  the shipper must



 determine,  in  accordance with regulatory  criteria (Section  I73.^55(b)), the appro-



 priate T.I. based on  nuclear  criticality safety.  For purposes of transportation, the



 shipper then must assign to the package the T.I. value.  Use of the higher T.I. value



 based on either the nuclear criticality safety criteria or the radiation level limitation



 (as described earlier) is required.








           During transportation, the carrier still  makes reference to the T.I. and



 stowage tables, even though the T.I. for fissile materials may be based on criteria



 other than the external radiation  levels.  For this reason, the absence of measurable



 external radiation from certain types of fissile radioactive materials packages would



 not necessarily constitute an "over label ing" violation.








      M.   Warning Labels








           Each package of radioactive material, unless excepted, must be labeled



(Section  172.403) on two opposite sides, with  a distinctive warning label.  Each of the



three label types bears  the unique trefoil  symbol  (Figure 7) recommended by the



 International  Commission on  Radiation  Protection (ICRP)  in 1956.   It  has  been

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 odopted by  the American  National Standards  Institute as  the  standard  radiation

 symbol (N2.1-1969).  The labels alert persons that the package contains radioactive

 materials  and that the package may require special  handling.  A  label with  an all

 white background color indicates  that  the external  radiation level is low and no

 special handling is required.  If the upper half of the label is yellow, the package  may

 have an external radiation  level or fissile properties  requiring consideration during

 transportation.  If the package bears a yellow label with three stripes,  the  transport

 vehicle must be  placarded RADIOACTIVE.  Placarding is discussed  in more detail  in

 Section IV. The criteria which the shipper must consider in choosing the appropriate

 label are listed in Table 7.
   Figure 7 - Package Labels

   Radioactive-White I
      (See §172.436)
      RADIOACTIVE \ /.
Radioactive-Yellow II
   (See §172.438)
V\RADIDACTIV[ 11.
Radioactive-Yellow III
   (See §172.440)
 ••\RADIOACTIVt II \/-f
           For  all  labels, vertical  bars on  each  label are  in  red.   Each label  is

diamond-shaped, four inches on each side, and has a block solid-line border one-fourth

inch from the edge.  The background color of the upper half (within the black line) is

white for the "I" label.  It  is yellow for the "II" and "III" labels.



           The  regulatory provisions in Sections  I72.403(f) and (g) applicable to the

use of these labels are:
                                         32

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            Each package required by this Section to be labeled with o RADIO-
            ACTIVE label must hove  two of these labels  affixed  to  opposite
            sides of the package.

            The following applicable  items of information must  be entered in
            the  blank  spaces  on the  RADIOACTIVE  label by  legible  printing
            (manual or  mechanical^ using o durable  weather resistant means of
            marking.

            "Contents"  - The name of the radionuclide, as token from the listina
            of  rodionuclides in Section  173.435  (symbols which  conform  to
            established radiation  protection  terminology  are authorized,  i.e.,
            °9     60
            " Mo,   Co, etc.).  For mixtures of radionuclides, the most restric-
            tive radionuclides  on  the  basis of radiotoxicity  must be  listed as
            space on the label allows.

            "Activity"  - Units shall be expressed in appropriate  curie units, i.e.,
            curies (Ci), millicuries (mCi) or microcuries (uCi)  (abbreviations are
            authorized). For a fissile  material, the weight  in grams or kilograms
            of the  fissile  radioisotope also may be  inserted  in  addition  to the
            activity.

            "Transport Index" - (See Section 173.403(bb)).
                                     TABLE 7

           RADIOACTIVE MATERIALS PACKAGES LABELING CRITERIA
                                 SECTION 172.403
 Transport Index
    a. i.)
Radiation Level at
Package Surface
     (RL)
Fissile
Criteria
Lapel    . ,
Category-
    N/A
      0.5 millirem per
       hour
       (mrem/h)
Fissile Class I Only
No Fissile Class I) or III
White -
    T.I. <. 1.0
0.5 mrem/h <  RL  <_50
Fissile Class 1,
Fissile Class II
with T.I. ^ 1.0,
No Fissile Class Ml
Yellow - II
    1.0  
-------
           At this point, it is appropriate to offer the following word of caution:
              Figure 8 - Caution — Do Not Confuse the Following:

              Radioactive Materials Package Labels (172.403 and
              172.436 Through 172.440)
                Radioactive Whlte-l
                Radioactive Yellow-ll
                Radioactive Yellow-Ill
              With:
                Fissile Classes I, II, or III (173.455)
      N.   Contominotion Control



           The  regulations prescribe limits (Section 173.443)  for control of remov-

able (non-fixed)  radioactive  contamination,  as shown  in  Table 8.   In  general,  the

contamination levels MUST  be  kept  as  low as  reasonably  achievable  and  the

significant  contamination  level  LIMIT  is applicable  to  any  package  offered  for

transportation.  It also applies to any transport vehicle which  is being released after

having been used to transport either an "exclusive use" load under  the  provisions of

Section I73.443(c) or a bulk shipment of LSA materials (Section !73.425(c)).



           The  limits shown in Table 8 (and Section 173.443) are for  the  activity

measured on a wipe taken on the package surface.  Measuring techniques other than

wiping may be used in accordance with Section 173.443.

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                                   TABLE 8

    REMOVABLE EXTERNAL RADIOACTIVE CONTAMINATION  - WIPE LIMITS
Contominonl
aeto/gammo-emitting rodionuclides;
all rodionuclides with half-lives
less than ten days; natural
uranium; natural thorium;
uranium-235; uranium-238;
thorium-232; thorium-228 and
thorium-230 when cr toined in
ores or physical concentrates 	 	
All other alpho-emitting
radionuc I ides. ........ 	 	

Maximum Permissible
Limits
uCi/cm
1C'5
ic-s
2
dom/cm
22
2.2
uCi/cm «= microcuries per square centimeter.
cipm/cm = disintegrations per minute per square centimeter.
     0.    Other Shipper Requirements



           As a brief review, the shipper must (I) select the proper packaging for the

specific contents; (2) consider the radiation level limits; (3) consider the contamina-

tion  limits; and (4) label correctly.   In  addition,  the shipper must  also ensure

compliance with the following:



           I.  . Package  Markings  - The outside  of the package must  be marked

          with (a) proper shipping name; (b) identification number  as shown in the

           list of hazardous materials (see Section 172.101); and (c) the appropriate

          specification number (see Section 173.24(c)( I )(D) OR Type B or fissile

          packaging certificate number, when  applicable.   Most of the pertinent

          regulatory requirements  for marking of all hazardous materials packages

          are found in Sections 172.300 through 172.308. The special requirements

          for radioactive materials are auoted below:
                                      35

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                "Section 172.310    Rodiooctive materials.

                      (a)   in  addition  to any  other  markings  required  by this
                      subpart,  each package containing radioactive materials must
                      be marked as follows:
      Gross Weight
      Type A
       and
      Type B
     Exports
     USA (with
     specification
     or certificate
     identification)
(1)   Each package of radioactive materials in excess of
I 10 pounds  (50 kilograms) must  hove its  gross weighl
plainly and durably marked on the outside of  the pack-
age.

(2)   Each package of radioactive materials which con-
forms to the requirements  for Type A or Type B packag-
ing (Section 173.403 of this subchapter)  must be plainly
and durably marked on the  outside of  the  package  in
letters at least  1/2-inch (13 mm.) high, with the words
"TYPE A" or "TYPE B," as  appropriate.  A packaging
which is  not in compliance with these requirements may
not be so marked.

(3)   Each package of radioactive material destined for
export shipment must also  be marked "USA" in conjunc-
tion with the specification  marking, or other package
certificate identification (see Sections 173.471, 173.472,
and 173.473 of this subchapter.)"
           The proper shipping names for radioactive materials are listed in Table 9.



           Spec.  7A  must also  be marked in accordance  with Section  178.350-3.

Where a duplication  of marking requirements exists  (such as  between the Section

I78.350-3(a) requirement  to mark the package "Radioactive Material" and when the

marked shipping name contains the words "Radioactive Material"), the markings need

not be duplicated.
                                       36

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                                    TABLE 9
MOST COMMONLY USED SHIPPING NAMES FOR RADIOACTIVE MATERIALS*
(FROM HAZARDOUS MATERIALS TABLE. SECTION 172.101)

Radioactive Material, Limited Quantity, n.o.s.**
Radioactive Material, Instruments, and Articles*"**
Radioactive Material, Fissile, n.o.s.
Radioactive Material, Low Specific Activity or LSA, n.c.s.
Radioactive Material, Special Form, n.o.s.
Radioactive Material, n.o.s.
Uranium Hexafluoride, Fissile (Containino more than 1%
U-ZJ5)
Uranium Hexafluoride, Low Specific Activity

UN 2910
UN 29! 1
UN 29 1 8
UN 2912
UN 2974
UN 2982
UN 2977
UN 2978

*    Refer to Section  172.101 for other proper shipping nomes.
**   n.o.s. means "not  otherwise specified."
***  Underlined words are not part of the proper shipping name.
           2.    Shipping  papers  - As  with other hazardous  materials  shipments,

           certain essential elements of information must be included on the shipping

           papers (see Sections 172.200 through 172.204).



                The information required on the shipping papers  is important to the

           carrier and consignee.   It also is of great value to emergency response

           personnel in the event of an accident.



                a.    Requirements (Section I72.202(o)(l))



                     NOTE: Enter in order listed below:



                     (I)   Proper shipping name from Section  172.101;
                     (2)   Hazard  class  (see  Section  I72.202(a)(2)),  hazard class

                     from Column  3, Section 172.101, except when the hazard class

                     is contained in the shipping name;

                                      37

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(3)    Identification  number  (see  Section I72.202(a)(3)  from



Column 3A, Section  172.101);








(A)    Net quantity of material by weight or volume  as stated



in Sections  I72.202(a)(4) and (c).  For most radioactive mate-



rials packages, if is not required to list the weight or volume.



The requirements of Section i72.203(d) provide better indico-



tions  of  potential  hazards  and  controls  required.   These



requirements  include  the package  contents as measured  in



curies and the transport  index.  A listing of  weight or volume




measurements for radioactive materials is usually needed only




for establishing transportation charges;








(5)    Radionuclide(s) contained in package (abbreviations are




allowed).  For a mixture of radionuclides,  only thos»  radio-



nuclides  which comprise 1% or more of the total activity  in




the package must be listed;








(6)    Physical and chemical form of material, or statement



that the  material is "special form"  (if it is  special  form).  A



generic description of material, such as protein, carbohydrate,



enzyme,  or organic salt, is authorized if exact chemical form




is difficult to specify;








(7)    Activity in curies  (Ci), miUicuries (mCi), or  microcuries




(uCi).  If the  package contains a "Highway Route Controlled



Quantity," those words  must  also be  shown  on  the shipping




papers;
                  38

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      (8)   Category of RADIOACTIVE laocis applied to package;








      (9)   Transport  index of the package if labeled RADIOACTIVE



      Yellow-ll or RADIOACTIVE Yellow-Ill;








      (10)  The information required in Section 172.203(d)(l)(vi) must



      be included if the shipment is "fissile" radioactive material;








      (I I)  The identification markings shown on the package must



      appear on the shipping paper if  the package  is approved  and




      certified by the Nuclear Regulatory Commission or  the  De-




      partment of Energy, OR is certified by DOT or other National



      Competent  Authority for mternati&na! shipment.








      (12)  Other information as required by the mode of transpor-




      tation  or subsidiary  hazard of  the  material.   (See  Section




      172.203.)








b.    Other Information and Examples of Shipping Paper Entries








      The regulations  require that certain specific descriptive infor-



mation  must  be  included  on shipping papers.   While there is no



specification for  shipping paper  format,  the first  three  entries of



the description must  be in  a specific order (see above).  Other



descriptive information is  allowed, such as the functional  descrip-




tion of  the product.  However, other information must not confuse




or detract from the required descriptions of the 'nazarous materials.
                       39

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     The  following  ore  some  exomp»e  entries  of  different  ways



shipments con be described on shipping papers:








     (I)    One (I) box, Radioactive material, special form, n.o.s.,



     UN 2974,  Rodiographic camera, lridium-192,  60  Ci,  Radio-



     active Yellow-ll, 0.6 transport index,  USA/9028/BCJ), Cargo



     aircraft only.







     NOTES:



     Physical and chemical form is  not listed  since  material  is



     "special form."








     The  Hazard class is  not  listed following  the  proper shipping



     name since it is contained in the shipping name.








     (2)    One (I)  carton, Radioactive material, n.o.s.,  UN  2982,



       Co, 30  mCi, liquid, cobalt in 50 ml 5%  hydrochloric  acid



     solution, transport  index  1.8, Radioactive Yellow-Ill and cor-



     rosive.







     (3)    One  (1)  box,  Thorium nitrate,  Radioactive  material,



     UN 2976,  15 kg,  Th  natural, solid (powder), thorium nitrate,



     1.3   mCi, Radioactive White-!  and  Oxidizer labels,  Cargo



     aircraft only.








     NOTES:



     Since the material  is specifically  listed  in Section 172.101,



     there is no "n.o.s." in the  proper shipping name and the hazard



     class Radioactive material is entered.

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Although  this materiel  meets the  definition  of  LSA (Section



173.403),  it  must be packaged ond shown on shipping papers as



specifically  listed material.  It must  meet packaging require-



ments as an oxidizer as required in Section 173.4 I?.







(k)    Three   (3)  drums,  Radioactive  material,  LSA,  n.o.s.,



UN 2912,  non-compacted solid debris  and waste,     Cs,    Co,


    90
and   Sr,  solid as inorganic salts or elemental, 0.04, 0.01, ond



0.005 mCi total, respectively;  Drum  Nos. 731, 680,  and 541.



See  attached forms  for  details.   Exclusive  use instructions



attached.








NOTES:



This entry is appropriate for describing drums that are shipped



as part of an "exclusive use" vehicle shipment.  Drums must be



marked "Radioactive LSA" and the vehicle must be placarded



RADIOACTIVE.  Package labels  are  not  required but are not



forbidden. The detailed contents of  each drum would be on a



sheet attached to the sheet with the basic description.







(5)    Thre*  (3) cartons, Radioactive Material, n.o.s., UN 2982,



Material to  be used in physical chemistry research project at



university.
Carton  No. 1, catalytic  specimen,   S, 70 mCi, solid, metal



oxide matrix, Radioactive White-1 label. 60 Ib.

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      Gorton No. 2,  Togged solvent,   Cl, 3 mCi,  liquid  non-flam-



      mable organic, Radioactive Whlte-l label. 50 Ib.






                                     5°       55
      Carton No. 3, converter element,  Te and   Fe, 30 mCi  and



      20 mCi, solid, steel parl, T.I.  1.6, Radioactive Yellow-ll label.



      80 Ib.







      NOTES:



      This is an example of how  one basic entry car, be used along



      with three different packages.  Detailed information  is given



      on the content, labels, and T.I. of each package.







c.    Documentation for Limited  Quantity Packages, Instruments or



      Articles, and Articles  Manufactured from Natural or Depleted



      Uranium or Natural Thorium.







      These  items are addressed in Sections  I73.42I-I73.424 and are



excepted from the detailed  shipping paper description.  They must



be documented  for transport as  required by Section  173.421-1  by



including a notice in, on, or forwarded with the package.  The notice



must  include the name and address of the consignor or consignee and



a specific statement which is selected on the basis  of the proper



shipping name for the package.  The  following example  illustrates



the notice on a shipping paper. The specific statement required by



Section 173.42!-I is shown in quotes.







      One (I) carton, Ajax  Model 123 Monitor, "This package con-



      forms'to  the conditions and  limitations specified in  49 CFR





                       42

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           173.422  for  excepted   radiooctive  materiel,  instrument,

           UN 291!." 45 !b.



3.    Shipper'^ Certification - The shipping papers must include o certifi-

     cate signed by the shipper.  This certification must  appear on the

     paper that  lists the required shipping description.



     o.    The following statement  is required by Section  I72.204(a) and

     must be used for all hazardous  materials shipments except for those

     by air.
           "This  is to certify that the above-named  (or  herein-named)
           materials are properly classified, described, packaged, mark-
           ed, and  labeled,  and are in proper condition for transportation
           according to the applicable regulations of the Department of
           Transportation.
     b.    For air transportation, the following language may be included

     on shipping papers in place of the statement in example (a)  above.
          "I hereby  certify that the  contents  of  this consignment are
          fully and accurately described above  by  proper shipping name
          and  are classified, packed, marked, and  labeled, and in proper
          condition for carriage by air according to applicable national
          governmental regulations."
          The requirements and limitations for carriage  of  radioactive

     materials aboard  aircraft are prescribed  in  Sections  !75.75(a)(3),

     175.700 through 175.705.  The following statement is required:

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           For packoges not acceptable for transportation on passenger-
           carrying aircraft:

           "This shipment is within the limitations prescribed for Merer,
           aor oif croft/cargo-only aircraft."  (delete non-applicable)

           For  pockoges  acceptable  for  transportation  on  passenger-
           carrying aircraft:

           This shipment  is within the limitations prescribed for passen-
           ger aircraft/corqa onlu aircraft,"  (delete non-applicable)
           Since  radioactive  materials  (other  than Limited  Quantities)

      can be  carried  on  o passenger-carry ing aircraft  only  if  they are

      intended for use in  research or medical applications, o statement to

      that effect must be included in the signed certification for shipment

      by passenger-carry ing aircraft.



A.    Security Seal  -  The outside of each Type A or Type B  radioactive

materials package must incorporate a feature, such as a seal,  which  is (a)

not readily breakable  and which, while intact, (b) will be evidence that the

package has  not  been illicitly opened  (Section  273.412(b)).   For  this

requirement,  the  package designer may  need to be skilled and creative.

This  is espeacially  true   for  packages,  such  as  fiberboard  cartons and

wooden boxes.  The regulations also require that "inner shield  closures

must be positively closed to prevent  loss of contents."  A padlock is not

effective as both a security seal and a closure mechanism.  Most padlocks

are not even a good security seal,  let alone a closure device.  It is usually

not possible with  most types of padlocks to ascertain if  they hove been

illicitly opened.   The best  approach  toward  meeting the dual  require-

ments, especially for Type B packages, is:  (a) serially numbered lead wire

seals,  IN COMBINATION WITH (b) such closure mechanisms as slotted

screw-in plugs, bolted  flanges,  and  positive action shutter mechanisms.

-------
 5.    Small  Dimension - The smallest  outside  dimension of any  rodio-



 octive materials package (other than excepted  quantities) must be four



 inches or greater (Section 173.412(a)).








 6.    Liquid Packaging Provision (Section  I73.4l2(n)) - Liauid radioactive



 material  must  be  packaged  in  a  leak-resistant  inner  container.    In



 addition,  the packaging must be adequate to prevent loss or dispersal of



 the  radioactive contents from  the inner container if the package were



 subjected  to the  30-foot  drop test prescribed in Section 173.466; and



 enough absorbent material must be provided to absorb at least twice the




 volume of the radioactive liquid contents.  Care  should be exercised by




 the  package  designer  to assure  that  the positioning  of  the absorbent




 material about the liquid-containing vessel is such that  the "absorber will




 absorb" in the event of leakage from the vessel.  For packages with liquid



 contents exceeding  50 cm , on alternative is provided to the use  of the



 absorbent in that a secondary outer leak-resistant containment vessel may



 be utilized.  The outer containment vessel must  hove  the ability to retain




 the radioactive contents under normal conditions of  transport, assuming




 the failure  of the  innermost primary containment vessel.  The package



 also requires a marking indicating the upward  position of  the  inside



 packaging (Section  172.312).








 7.   Surface Temperature  of Package  -  Maximum surface  temperature



 limits on  packages, resulting from radioactive thermal decoy energy  of



the contents, are prescribed  in Section 173.442.  The limit  is  either I22°F




or, in the case of exclusive use shipments,  I80°F.

-------
           8.    Quality Control Requirements (Sections  173.474 ond  173.675) - The


           regulations also  prescribe certain  quality control  requirements for  the


           construction  of  radioactive  materials pockogings  (Section 173.A7A)  and


           before  eoch shipment of a package  (Section  173.475).  With regard to


           packages of liquids containing  in excess  of Type A  quantity, destined for


           shipment by air,  an additional requirement (Section  I73.475(g)) is imposed


           such that  the containment system  of eoch package offered for shipmen*


           must be  tested  to assure that it  will remain  leak-free  in  a specified


           ambient reduced atmosphere  (0.25 atmosphere).





                For Type A packages of the fiberboard box variety, sealing tape and


           the consignor's  labels  offer opportunities  for  compliance  IF they  can


           provide positive evidence that the package has not been opened.
IV   CARRIER  REQUIREMENTS IN  HANDLING OF RADIOACTIVE  MATERIALS
     PACKAGES	        ~
     Up to this point, this Review of the regulations has been concerned principally


with the  regulatory  requirements  applicable to the  shipper, and/or  the  package


designer.  The reason for this is simple.  Most of the regulatory requirements for the


assurance of safety  in the transport  of  radioactive  materials  are  directed towards


safety  through proper packaging.  Thus,  the majority of these requirements apply to


the shipper.  In transport of radioactive materials, the principal carrier responsibili-


ties are as follows:





     A.    Shipping Papers ond Certification by Shipper





           Carriers may not knowingly accept for transport packages of radioactive


materials  which  have  not  been properly described  and certified  by  the shipper

-------
pursuant to Section  172.204.   This  certificote is  relied upon  by the corrier, as

evidence that  the packaging is in accordance with the  regulatory requirements-  In

the case of air «hipments, one signed copy of the shipping paper must accompany the

shipment (Section  I75.35(a)).  Th* originating air carrier must  retain o second copy

(Section I75.30(a)(2)).



          Carriers may prepare and carry  with the shipments appropriate bills of

lading, waybills, etc., based on the  information derived from the shippers' shipping

papers (Sections 174.24, 175.35, 176.24,  and 177.817).   For shipments by vessel, o

Dangerous Cargo Manifest or storage plan is also required (Section 176.30).



     B.   Placarding



          The carrier  must  apply  the RADIOACTIVE placard  to  the  transport

vehicle  (rail or highway)  if ANY radioactive material package on board bears o

"Radioactive YELLOW-11!" label (Section 172.440).   The format for  the placard is

illustrated In Figure 9.  The requirements for placarding are in Section  172.504 and

Table I footnotes of that Section.

         Figure 9
                             ADIOACTIV
        (The background color for the black trefoil in the upper half of
         this 12" x 12" placard is yellow.)
                                      47

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           Highwov Route Controlled Quontity Plocording



           Vehicles transporting  ANY  package  which contains  c highway  route

controlled quantity must also display the  square  white background as specified in

Section 172.507.



           A Word of Caution - Plocording of LSA Radioactive Materials



           Questions occasionally arise  about the  placarding requirements  for  "Ex-

clusive  Use" (sometimes  referred to as  "full-load" or "sole-use") shipments of  low

specific  activity radioactive  materials (LSA).  Under the shipper requirements of

Sections I73.4l5(b) and (c), transport vehicles must be placarded by the shipper with

the placard which is normally  required  to be applied by the  carrier (pursuant to

Sections  I72.504(a) or  172.556).  Some persons hove apparently misinterpreted the

provisions of Section 173.415(b)(7) assuming  that if the shipment of  LSA materials

bears no packages with Radioactive  Yellow-Ill labels, then placarding is not  required.

This is  not the  case.  In  fact, the packages in such  shipments  are excepted ONLY

FROM specification packaging, marking, and labeling pursuant to Section 173.A25(b).

The  requirement to placard  is, therefore,  imposed on the  shipper rather  than the

carrier.   This  is  consistent  with  the higher external radiation levels  of Section

173.44Kb) which are allowed  for such shipments.  To some  degree, the requirement

for the  shipper to placard the vehicle is imposed instead of requiring the  radioactive

material warning labels on each package.
     C.    Radiation Exposure  Control  by  Maximum  Total  Transport  index  vs.
           Distance
           For any  group o.f "yellow-labeled"  packages  in o  single conveyance or

storage location, the carrier must  assure  that  the total transport  index  does not

                                       48

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 exceed 50.  The carrier must assure that  such groups of yellow-labeled packages are



 kept separated from  undeveloped film shipments and areas  normally  occupied by



 persons.  The minimum separation distance must be in accordance with o table of



 distances based on the graded  total transport index (Sections I74.700(c),  175.701



 through 175.703,  176.708, and  I77.842(b)).  This  separation  reduces  the  rate  of




 exposure to radiation.








      D.   Reporting of Incidents








           The carrier must assure that DOT  and the shipper are notified in the




 event of fire, breakage, spillage, or suspected radioactive contamination  involving a




 shipment of radioactive material.  Also, carriers must assure that vehicles, areas, or




 equipment in  which radioactive  material may have spilled are not placed  in service




 again until  they hove been surveyed and decontaminated  (Sections 171.15, 171.16,




 174.750, !75.45(a)(4), !76.48(b), and 177.861 (a)).








           The reporting  requirement  cited above  is not  necessarily  a means of



 receiving technical assistance in  radiological monitoring in the event of a transporta-




 tion incident.  To obtain technical assistance, carriers may call upon the services of




 local or state  radiological authorities.








           Federal assistance in resolving a radiological emergency may be provided



 If requested by state  or  local authorities.   As  with emergencies  of nature (floods,



 fires, tornados,  etc.),  the responsibilities  for  resolutions  of  radiological or other




 hazardous materials emergencies belong basically with state and local governments.




 Federal involvement is limited unless requested. When technical advice or assistance




 is needed for  a radiological or other hazardous materials accident and the  state or




local radiological authority is  not known,  help can be obtained  by contacting the

-------
 Chemical Transportation Emergency Center  (CHEMTREC) al (800) 424-?300.  For

 emergencies  involving  radioactive  materials,  CHEMTREC will refer the  problem to

 s+ate or local authorities and/or to the Deportment of Energy's (DOE) Radiological

 Assistance Coordinating  Offices.   The DOE radiological experts will determine the

 nature  of  the  problem.   Based upon their  assessment,  they  will  provide  advice,

 arrange for state or  local assistance, or dispatch o DOE team of radiological experts

 to assist in the emergency.  The  map of  the  DOE Regions showing  the telephone

 numbers of the  Coordinating Offices is  shown  in Figure  10.  These offices can also be

 called directly for radiological assistance.
Figure 10 - Regional Coordinating Offices for Radiological Assistance and
            Geographical Areas of Responsibility
REGIONAL
COORDINATING
OFFICE
NftROOKHAVEN
Y AREA OFFICE
L OAK RIDGE
p OPERATIONS
1 OFFICE
| SAVANNAH
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•y OPERATIONS
j OFFICE
L AIR-UOUEROUE
f\ OPERATIONS
r OFFICE
L CHICAGO
5) OPERATIONS
r OFFICE
L IDAHO
3} OPERATIONS
•^ OFFICE
L SAN FRANCISCO
T\ OPERATIONS
< OFFICE
RICHIAND
BS OPERATIONS
J OFFICE
OFFICt
ADDItii
•** »!>•« »«•»»
r » ««• c
tt.allltl »1X
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«UIN t. C Hill
r.e.m MB
U.MKUIMUI.
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r.e 101 uc
MCMLMIC
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AlitiUNCt
(Sit) M57IOD
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(CIS) S25-7I15
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ISD9) 37E-73E1
                                        50

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      E.    Notification TO pilot (to- Aircraft Shipments)








           The aircraft operator must provide, in writing,  the  pilot in command of




the aircraft the following information:  (a) the name, (b) type of label, (c) quantity,




and (d) location of any "hazardous material," such as radioactive materials packages.



The cargo load manifest  must be conspicuously marked to indicate the presence of



such packages (49 CFR 175.33).








V.    DISCREPANCIES IN RADIOACTIVE MATERIALS SHIPMENTS








      This discussion  is intended to serve as an aid to both shippers and  carriers.




Noncompliance in radioactive materials  shipments is  generally either  of  a  safety




related nature, i.e., improper packaging, excessive radiation or  contamination, or an




administrative nature, i.e., improper shipping paper description, illegible labels, etc.




Items of noncompliance which deserve special attention in order to avoid are:








      A.    By Shippers








           1.    Excess radiation levels (Section I73.403(bb)) - Fortunately,  this item




           is not noted frequently.   It is one  of the  most serious type of shipper



          violations  of  the safety requirements for  transportation of radioactive



          materials.  Excessive radiation levels on packages of radioactive mate-



          rials indicate inadequate planning, procedures, and/or practices.  They are



          generally the result of:








                a.    An unsatisfactory monitoring of the entire package.








                b.    Using inadequate radiation measuring instruments.
                                       51

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      c.    Using instruments that are not properly calibrated.








      d.    A failure  to  properly secure o shielded  closure mechanism, o



      faulty closure mechanism, etc.








      e.    Use  of pockogings  for  materials  for which  they were not



      designed.  For  example,  putting radioactive materials into o con-



      tainer that  exceeds the shielding  capabilities  of  the  package,  or



      placing materials into o  container  thai is not compatible with the




      material's physical or  chemical properties.    For  "special  form"



      sources, this results in a potential  for excessive  radiation  levels.




      For dispersible, normal form radioactive materials, o hazard may  be




      present due to both excessive radiation, as well  as possible dispersal



      of  loose contamination.  The very  short half-life of some radioiso-




      topes occasionally presents problems in transportation.   Some sup-



      pliers  load more  than the total quantity  allowed  for shipment,  so



      that the  radiation  levels at  the actual time of shipment  will  be



      within the limits,  taking into account radioactive decay. This is a



      violation  if the package  is offered for shipment too soon, that  it,



      before the material has decoyed to legal limits.








2.    Improper  Packaging - This is also o most  serious safety  item.  It is




closely  related to the  excessive  radiation level  in that on improper




package  may  not  incorporate  sufficient thickness of shielding  for the




material.  Another  example  of  improper  packaging  is the use of  a



packaging not  authorized in the regulations or under an  exemption  or



other specific approval.
                             52

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      Safe*» may be affected even when or approvea package is  j

 F IT  IS NOT  IN ITS  PROPER CONDITION  AS  REQUIRED  RY ITS

 DESIGN.   Good  quality  control practices by  shippers of  radioactive

 materials are  of paramount importance.   A  relevant  requirement of

 Section 173.22 reads as follows:
      "The person shall  determine that the packaging or container has
      been manufactured, assembled, and marked ..."
      The above provision is cited as o reminder to shippers.  No package

 will perform  during  transportation as intended by  its original  design,

 unless it  is in its proper design condition.  It  must be "as good as new"

 when offered  for transportation.  The  quality control requirements of

 Sections 173.474 and  173.475 serve to clarify these aspects.



 3-    Lock  of Security  Seal  (Section 173.412(b)) - This requirement  is

 sometimes misunderstood by shippers of radioactive material.  It is really

 o performance type requirement, wherein—
      "The outside of the packaging incorporates a feature, such as o seal,
      that is not readily breakable, and that, while  intact, is evidence  that
      the package has not been opened."
      On some types of packages, i.e., steel drums, hinged lid boxes, etc.,

provision for a security seal is fairly simple.  On many other types, i.e.,

wooden boxes,  fiberboard  cartons, much more  thought  and ingenuity in

designing a seal  to meet the requirements will be necessary. The use of

padlocks as a security seal may not, in all cases, be appropriate.  Many

types of padlocks  may be opened and closed again without knowledge of

the consignee.
                            53

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4.    improper Lobels - Incorrect labeling of radioactive materials pock-



ages is a common deficiency.  The most frequent error is "overlabeling."



A YELLOW-III  label  is used where o WHITE-! or  YELLOW-1!  label  is



prescribed.  The trree labels are used to  indicate the degree of control



the package requires.   "Overlabeling" signals  incorrect hazard  warning



information.








5.    Illegible/Incorrect Label Notations  -  This item  should  speak for




itself.  Needless to say, shippers should exercise care to  insert legible,




durable entries on the labels. These entries coll for noting the "contents,"




"activity," and "transport  index."   The name of the radionuclide  or  its




abbreviation must be entered on the Contents line exactly as on the list  of



radionuclides in Section 173.435.  Clearly indicate the exact number  of




curies,  millicuries or  microcuries  on the  Activity line.   Finally, the




"transport  index" must be  rounded  up  to  the  next  highest  tenth   as



prescribed in Section 173.403(bb).








6.    Improper or  Incomplete Shipping Paper Description  -  The  basic



requirements for the shipping paper description are prescribed in Subpart



C of  49 CFR  Part  172.  The  first,  second, and third  elements on the



shipping  paper must always  be  the applicable  "proper shipping  name,"




"hazard class," and "identification number," exactly as listed in the table




in Section  172.101.   Any other required description or  information (not




inconsistent therewith) may  follow.  However, there are at present  16




different proper shipping names for radioactive materials in the hazardous



materials table.   The most  commonly used  names are listed in Table 9.




The most appropriate name  must be  used.   Care  should be  exercised  to

-------
      properly  enter  the other  information  as  required  by Section  I72.203(d).




      When  the  shipment  involves o  radioactive  material  prohibited  from




      transport by passenqer-carrying aircraft, the description shall also include




      the words "cargo aircraft only."








      7.    Inadequate Provision for Liquid Contents - The regulations prescribe




      certain additional packaging  requirements  for  liquid radioactive mate-




      rials.  As required by Section 173.4 I2(n)  and Section (73.466(a)(l),  these




      are basically the  "performance" requirements.  The  package must  with-




      stand  o 30-foo1-drop test without  loss  of liquid contents.  Absorbent




      material  must be  present to absorb the  liquid  contents in  the  event of
     breakage of the primary liquid container for contents which do not exceed



     50 cm .   A double  containment  vessel  system  that will  survive  all



     applicable testing  may  be used  as an  alternate  packaging  when  the



     contents  exceed 50 cm  .   The package orientation marking  (example:



     "This side up 'h ")  is also required.








B.   By Carriers








     L    Acceptance of Consignments Without Shipper's Certification  -  The



     regulations applicable to the carrier specify that shipments of regulated



     hazardous materials may NOT be accepted  unless  accompanied by the



     appropriate certification.  The shipper must certify that the material has



     been  properly packaged,  marked and  labeled  in accordance  with  the



     regulations. This certification must be signed by the  shipper.  It is a legal



     representation to the carrier that the safety requirements of the shipment



     are in order.   Needless to  say,  originating  carriers must  not accept



     hazardous materials which are offered to them without this certification.







                                  55

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2,    P'oilure to Prepore Proper Snippinc &gper Description - Each of the



carrier  regulations permit carrier-prepared manifest, way bills,  etc., to



include  the proper shipping description and on indication of the type of



label applied. In many cases, carriers,  m preparing  their shipping papers,



do  not  accurately copy these essentio1  items of information from the



shippers' papers.







3.    Acceptance  of  Radioactive Materials Consignments Exceeding the



50  Transport Index  Maximum  Per  Vehicle  and Inadequate Separation



Distances -  In many  cases, carriers  either do not appear to be aware of



this limitation or  they  blatantly fail to  follow it.   Each of the carrier



regulations contain a table which prescribes certain segretation distances



and in  certain cases, stowage  times  for  accumulations  of radioactive



materials packages.  These  distances  and times are based  on the  total



transport index. These segregation controls are intended to provide a safe



separation  distance of  radioactive materials packages from areas occu-



pied by  persons or photographic film.







A.    Failure to Properly  Placard  Transport Vehicles  -  For any rail or



highway vehicle transporting  any quantity of radioactive  materials pack-



ages bearing the radioactive YELLOW-III label, the carrier is required to



display the RADIOACTIVE placard.  Intentional failure to  placard vehicles



is a very serious offense.








5.    Inadequate Vehicle Safety - Other safety regulations,  such as  those



of the Bureau of  Motor Carrier Safety, 49 CFR Parts 390-397, play an



important role in the safe transport of radioactive and  other hazardous



materials.  Even  though shippers and carriers may  be in  full compliance





                             56

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           with the Hazardous Materials Regulations, an unsafe vehicle can increase



           the risks to the public and the transport  workers.  Inspections have often



           disclosed vehicle  safety violations of greater possible  consequence than




           the hazardous  materials violations.   Be  certain that the  transporting



           vehicle is in proper condition before accepting a shipment.








VI.   IAEA REGULATIONS (AS AMENDED)








      The IAEA (International Atomic Energy Agency) has published the "Regulations




for the Safe Transport of Radioactive Materials, Safety Series No. 6,  1973 Edition (as




amended)".  These regulations have now been accepted and adopted (either wholly or




in part)  by  most  nations as  their standard for  both national  and international




regulations.  The United States revised the  DOT regulations  as of  July 1,  1983, to




achieve a substantial conformity with the 1973 IAEA standards.








      During September 1980, the IAEA convened a  panel of experts from its member




countries for the purpose of  considering proposed changes to  the  IAEA regulations.




These proposed changes had been submitted in advance of the convening of the panel



by each member country.  As  a result of that panel,  drafts of comprehensive revisions



to Safety Series No. 6 were were subsequently prepared.  The  second draft  was



submitted to each IAEA member country and various international  organizations in



May 1981.  Official comments on it were then submitted  to IAEA in 1981.  Another



meeting of the review panel was then convened to  finalize the revisions.  The IAEA




regulations in Safety Series No. 6 should be revised in 1983.  Following these changes,



DOT will propose changes, if needed.
                                       57

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VII.  DErINITiQN!S







     The following definitions ore derived  from the Code  of  Federal  Regulations,



Title 49 - Transportation, Section 173.403.







     CLOSED TRANSPORT VEHICLE  - A vehicle equipped with a securely attached



     exterior  enclosure,  which during normal  transport  restricts the  access of



     unauthorized persons to the cargo space containing the  radioactive material.



     The enclosure may be permanent or temporary,  may be  of  the  "see-through"



     type and must limit access from the top, side, and ends.  (Section I73.403(c))








     FISSILE CLASSES - The groupings into which radioactive material  packages are



     classified according to the controls needed to provide nuclear criticality safety



     during transportation.  (Section 173.455)








     FISSILE  MATERIAL  -  Piutonium-238,  plutonium-239,  plutonium-241,  uro-



     nium-233,  uranium-235,  or  any  material  containing  any of  the  foregoing.



     (Section  173.403(0)







     LIMITED QUANTITY RADIOACTIVE MATERIALS - A quantity of radioactivity



     which  does exceed the limits specified in Section 173.423.  Limited quantities



     and certain  radioactive  instruments  and articles (Sections  173.421  through



     173.424)  are  excepted from specification packaging, shipping paper and certifi-



     cation, marking  and labeling  requirements, but  are still subject  to certain



     requirements as specified in Sections 173.421 through  173.424.
                                       58

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LOW SPECIFIC  ACTIVITY MATERIAL  -  Material  in  which the activity  is



essentially uniformly distributed and in which the estimated overage concentra-



tion per  gram  of  contents does not  exceed the  specifications of  Section



 I73.403(n).








SPECIAL  FORM RADIOACTIVE MATERIALS -  Those materials which, by



nature of their  physical form  or encapsulation,  if  released from a package



might present some direct radial ion hazard but would present little hazard from



the possibility of contamination. (Section I73.403(z))








NORMAL FORM RADIOACTIVE  MATERIALS - Those materials which do not




meet the  requirements of Special  Form Radioactive  Materials.   (Section




I73.403(s))








TRANSPORT  INDEX -  A  number  placed  on a  package  of  "Yellow-Label"



radioactive  materials by the shipper to denote the  degree of control to be




exercised by  the carrier,  i.e., to determine the number of  yellow-labeled



packages which  may be placed in a  single vehicle or storage location.  The




transport index  is either the highest  measured dose rate of radiation at three



feet from the surface  of  the  package,  or a number assigned  for  criticality



control purposes. (Section !73.403(bb))








TYPE "A" PACKAGING - Packaging  which is designed in accordance with the




general  packaging requirements of  Sections  173.24 and 173.412, and which  is




adequate to prevent the loss or  dispersal of the  radioactive contents and to



retain the  efficiency of its  radiation shielding properties if  the package  is



subject to the tests prescribed in Section  173.465.

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     TVPC Hg» PACKAGING  -  Pockaging which meets the sTondords  for Type "A"



     pockog ing ond, in oddition,  meets the stondords  for the hypothetical  accident



     conditions of transport as prescribed in 10 CFR Part 71.








     The  following definitions,  though  not  derived   from ,the  Code  of   Federal



Regulations, Title 49, are held as generally accepted meanings of the terms listed.








     ALPHA PARTICLES - One of the three primary forms of radioactive emissions



     from  radioactive  atoms.   Alpha particles  are  positively  charged  particles



     emitted  from the nucleus of a radioactive atom  ond  hove o mass and charge



     equal to the nucleus of a helium atom (2 protons  * 2 neutrons).  Alpha particles



     hove very little penetrating  ability and, therefore, are chiefly internal  radiation



     hazards. They travel very short distances in air and are shielded very easily.








     BETA PARTICLES - One of three primary forms of radioactive emissions  from



     radioactive  atoms.   Beta  particles are negatively  charged  particles emitted



     from the nucleus of a radioactive atom and hove a mass and charge equal to



     that of an electron.   They usually travel greater distances in air than alpha



     particles, have an  intermediate  penetrating ability, but still can be easily



     shielded with common materials.








     GAMMA RAYS - One of three  primary forms of radioactive emissions  from



     radioactive atoms.  Gamma rays are not particulate (as opposed to alpha and



     beta particles), but are short-wave length electromagnetic radiations from the



     nucleus of radioactive  atoms.  Except  for their origin  (the nucleus of  the atom



     rather  than the outer shell),  they are identical  in characteristics to X-rays.



     Gamma  rays  are the  most penetrating  form of radiation  and  travel great



     distances in air before absorption.  They require heavy shielding materials, such



     as lead, to attenuate the radiation.

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CURIE - An expression of the quantity of rodiooctive material  in terms of the



number  of  atoms which disintegrate (decoy) per second.   A curie (CO is that



quantity  of  radioactive material which  decoys  such that 37  billion atoms



disintegrate per second with each  disintegration  resulting in  the emission of



alpha or beta particles and/or gamma rays.  One thousandth  of o  curie  is a



milMcurie (mCi); one millionth of a curie is a microcurie (uCi).








RADIATION LEVEL  -  A term sometimes used instead of radiation "dose rate"



or "exposure rate."  It  generally refers to the  effect of radiation on matter,



that is, the  energy imparted to and absorbed by matter due to emitted radiation




per unit of  time.








MILLIREM (One one-thousandth of a rem) - The rem is o unit  sometimes used



to  express  radiation  level or dose  rate  (millirem  per  hour).   Technically



speaking,  the  rem  is an  expression of "radiation dose  equivalent"  which



considers  the biological effect  of the absorbed radiation.   Do not confuse




millirem with curie.








ENCAPSULATION - The term used  to denote  an additional fabrication tech-



nique often used in preparation of radiation sources, wherein the basic material



is physically placed within sealed, high physical integrity capsules or envelopes



to provide further assurance that in the event a package breaks and the capsule



escapes,  there  would be little possibility of a spread of radioactive contamina-



tion.








NUCLEAR  CRITICALFTY  -  This term  denotes the  occurence  of a chain



reaction with fissile radioactive materials. The purpose of the Fissile Classes is



to prevent  the  occurrence of nuclear criticality during the transport  of Fissile




                                  61

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      Materials.  (Controlled nuclear  criticality is the abjective within a nuclear

      power reactor.)



      RADIOISOTQPE  AND RADIONUCLIDE  - For the purpose of transportation,

      these terms are synonymous with "Radioactive Materials" and identify specific

      isotopes of chemical elements that have radioactive properties.



      RADIOTQXICITY -  A term used to denote the relative hazards of the various

      radionuclides, that is, their internal radioactive effect within the body.



      WIPE SAMPLE - A  test for loose  or removable radioactive contamination on

      surfaces (also sometimes referred to as a "smear" test).
     This material may be reproduced without  special permission from this office

and is available from:
                                Department of Transportation
                                Research and Special Programs Administration
                                Materials Transportation Bureau
                                Information Services Division (DMT-11)
                                Washington, D. C. 20590
                                       62

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VIII. REFERENCES



     To supplement this review, the following publications relating to transportation

of radioactive materials are listed:



     I.   IAEA "Regulations for  the Safe Transportation of Radioactive Materials,

          Safety Series  No. 6, 1973 Edition (as  amended)," International  Atomic

          Energy Agency, Vienna, Austria.
          Availability;     Unipub, Inc.
                          1180 Avenue of the Americas
                          New York, New York  10038
     2.   "The Safe Transport of Radioactive Materials/' edited by R. Gibson, 1966.
          Availability;     Pergamon Press, Inc.
                          44-01 21st Street
                          Long Island City, New York  11101
     3.    "Environmental Survey of Transportation of Radioactive Materials To and

          From Nuclear Power Plants," U. S. Atomic Energy  Commission  (NRC),

          Wash-1238, December 1972.



     4.    "Draft Environmental  Impact Statements  on Transportaiton of  Radio-

          active Materials by Air and Other Modes,"  (NUREG-OI70), U. S. Nuclear

          Regulatory Commission, Office of Standard  Development, March 1976.



     5.    "Survey of Radioactive Material Shipment  In the United States,"  BNWL-

          1972, Battelle Pacific Northwest Laboratories, Richland, Washington.

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6.   "Evaluation of Radiation Emergencies and Accioents—Selected Criierio

     and Data," Technical Report Series No. 152, IAEA, Vienna, Austria, I97A.



7.   "Certification of ERDA Contractors' Packaging With Resped to Compli-

     ance  with DOT  Specification 7A  Performance Requirements,"  -  Two

     reports by Mound Laboratory, Monsanto Research Corporation, as follows:



     Phase II Summary Report - June 12,  1979, MLM 2228.

     Phase II Summary Report (Supplement No. I)  -  April 15, 1976, MLM 2228,
     (Suppl. I).
     Availability;     National Technical Information Service
                     U. S. Department of Commerce
                     Springfield, Virginia  22161
                     (703) 557-4650
                     * US GOVfcONMtNl iwmllNC OHCf l»8S'-6!«-OK> - 000 1 i

-------



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-------
                U.S.  Nuclear  Regulatory  Commission
               and Agreement State Offices  09/09/90
USNRC REGION I
USNRC REGION II
USNRC REGION III
USNRC REGION IV
USNRC REGION V
Alabama
Arizona
Arkansas
California
215-337-5000
475 Allendale Road
King of Prussia, Pennsylvania 19406

404-331-5500
101 Marietta Street, NW
Suite 2900
Atlanta, Georgia 30323

312-790-5500
799 Roosevelt Road
Building 4
Glen Ellyn, Illinois 60137

817-860-8900
611 Ryan Plaza Drive
Suite 1000
Arlington, Texas 76011

415-943-3700
1450 Maria Lane
Suite 210
Walnut Creek, Calif. 94596

205-261-5315
Aubrey Godwin, Director
Division of Radiological Health
Department of Public Health
Room 510, State Office Building
Montgomery, Alabama 36130

602-255-4845
Charles F. Tedford, Director
Arizona Radiation Regulatory Agency
925 South 52nd Street
Phoenix, Arizona 85040

501-661-2301
Greta Bicus, Director
Div. of Radiation Control and Emerg. Man.
Department of Health
4815 West Markham
Little Rock, Arkansas 72205

916-445-0931 (License Insp.)
Jack McGurk, Chief
Radiological Health Board
State Dept.of Health Services
714 P Street, Office Bldg. #8
Sacramento, California 95814

-------
Colorado
Florida
Georgia
Idaho
Illinois
Iowa
Kansas
303-331-8480
Robert Quillin, Director
Radiation Control Division
Department of Health
4210 East llth Avenue
Denver, Colorado 80220

904-487-1004
Mary E. Clark, Ph.D.
Office of Radiation Control
Dept. of Health & Rehabilitative Services
1317 Winewood Blvd.
Tallahassee, Florida 32399-0700

404-894-5795
Thomas E. Hill, Acting Director
Radiological Health Section
Department of Human Resources
878 Peachtree Street, Room 600
Atlanta, Georgia 30309

208-334-5879
Ernie Ranieri, Prog. Man.
Radiation Control Section
Department of Health and Welfare
Third Floor, 450 West State Street
Boise, Idaho  83720

217-785-9947
Steve Collins
Division of Radioactive Materials
Department of Nuclear Safety
1035 Outer Park Drive
Springfield, IL  62704

515-281-3478
Don Flater, Director
Environmental Health Section
Bureau of Radiological Health
Iowa Dept. of Public Health
Lucas State Office Building
Des Moines, Iowa 50319

913-296-1560
John Erwin, Manager
Bureau of Radiation Control
Dept. of Health & Environment
Building 740, Forbes Field
Topeka, Kansas 66620

-------
Kentucky
Louisiana
Maryland
Mississippi
Nebraska
Nevada
New Hampshire
502-564-3700
Donald R Hughes,  Manager
Radiation Control Branch
Cabinet for Human Resources
275 East Main Street
Frankfort,  Kentucky 40601

504-925-4518
William H.  Spell, Administrator
Nuclear Energy Division
Office of Air Quality & Nuclear Energy
P.O. Box 14690
Baton Rouge, Louisiana 70898-4690

301-631-3300
Roland G. Fletcher, Administrator
Radiological Health Program
Dept. of the Environment
2500 Broening Highway
Baltimore,  Maryland 21224

601-354-6657/6670
Eddie S. Fuente,  Director
Division of Radiological Health
P.O. Box 1700
State Board of Health
Jackson, Mississippi 39215

402-471-2168
Harold Borchert,  Director
Division of Radiological Health
State Department of Health
301 Centennial Mall South
P.O. Box 95007
Lincoln, Nebraska 68509

702-885-5394
Stanley Marshall, Supervisor
Radiological Health Section
Dept. of Human Resources
505 East King Street, Room 203
Carson City, Nevada 89710

603-271-4588
Diane Tefft, Program Manager
Radiological Health Program
P.O. Box 148
Concord, New Hampshire 03301

-------
New Mexico
New York
North Carolina
North Dakota
Oregon
Rhode Island
South Carolina
505-827-2940 (Ext. 279)
Richard Mitzelfelt
Radiation Protection and Licensing
Environmental Improvement Division
Harold-Rennel Building
1190 St. Francis Drive
Santa Fe, New Mexico 87504

518-458-6461
Karim Rimawi, Director
Bureau of Environmental Radiation Protection
State Health Department
University Plaza, Western Avenue
Albany, New York 12205

919-733-4283
Dayne H. Brown, Director
Division of Radiation Protection
N.C. Dept.  of Env. Health and Natural Resources
701 Barbour Drive
Raleigh, North Carolina 27603

701-224-2348
Dana Mount, Director
Div. of Environmental Engineering
N.D. State Department of Health
P.O.BOX 5520
1200 Missouri Avenue
Bismarck, North Dakota 58502-5520

503-229-5797
Ray D. Paris, Manager
Radiation Control Section
Division of Health
Dept. of Human Resources
1400 South West Fifth Avenue
Portland, Oregon 97201

401-277-2438
Charles C. McMahon, Acting Supervisor
Div. of Occupational and Radiological Health
Rhode Island Dept. of Health
206 Cannon Building
3 Capitol Hill
Providence, Rhode Island 02908

803-734-4700
Heyward Shealy, Chief
Bureau of Radiological Health
State Dept. of Health and Env. Control
J. Marion Sims Building
2600 Bull Street
Columbia, South Carolina 29201

-------
Tennessee
Texas
Utah
Washington
615-741-7812
Michael H. Mobley, Director
Division of Radiological Health
Terra Building, 150 Ninth Ave. North
Nashville, Tennessee 37219-5404

512-835-7000
David K. Lacker, Chief
Bureau of Radiation Control
Texas Department of Health
1100 W. 49th Street
Austin, Texas  78756-3189

801-538-6734
Larry Anderson, Director
Bureau of Radiation Control
State Department of Health
P.O. Box 1-6690
288 North, 1460 West
Salt Lake City, Utah 84116-0690

206-753-3468
T.R. Strong, Chief
Department of Health/ Radiation Protection
Mail Stop LE-13
Air Industrial Park-Building 5
Olympia, Washington 98504

-------
                            SUPPLIERS
Instrument

Eberline Instrument Corp.
504 Airport Road
P.O. Box 2108
Santa Fe, NM  87504-2108
800-678-7088
505-471-3232
Ludlum Measurements, Inc.
501 Oak St.
Sweetwater, TX  79556
915-235-5494
Canberra Industries, Inc.
1 State St.
Meriden, CT  06450
800-243-4422
Dosimeter Corp.
P.O. Box 42377
(11286 Grooms Rd.)
Cinti., OH  45242
513-489-8100
800-543-4976
EG&G ORTEC
100 Midland Rd.
Oak Ridge, TN  37830
615-482-4411
EG&G Instruments
Nuclear Products Group
P.O. Box 486
Lenoir City, TN  37771
615-986-4212
800-251-9750
Tennelec Inc.
601 Oak Ridge Turnpike
P.O. Box 2560
Oak Ridge, TN   37830
615-483-8405

-------
 victoreen
 6000  Cochran Road
 Cleveland,  OH  44139-3395
 216-248-9300
 fax 216-248-9301
 Sources

 The Nucleus,  Inc.
 P.O.  Box  2561
 761 Emory Valley Rd.
 Oak Ridge,  TN  37830
 615-482-4041
 615-483-0008
Services

Teledyne  Isotopes
50 VanBuren Ave.
WestWOOd, NJ  07675
201-664-7070
fax 201-664-5586
Diagnostic Engineering, Inc.
347 Stanley Ave.
Cinti., OH  45226
513-271-3737
General Dynamics
14 Holmes Street
Mystic, CT  06355-2644
203-572-8971
Environmental Dimensions, Inc.
3939A San Pedro Drive, N.E.
Albuquerque, NM  87110
505-881-9427
Duratek Corp.
6700 Alexander Bell Drive
Columbia, MD  21046
30; -2:50- 2340

-------
Contamination Control Equipment

Defense Apparel (D.A. Services Inc.)
247 Addison Road
Windsor, CT  06095
800-243-3847
203-285-0808
fax 203-688-5787
Lane's Industries, Inc.
12704 NE 124th- Street
Kirkland, WA  98034
206-823-6634
Nuclear Power Outfitters (NPO)
Division of PPI Industries
P.O. Box 737
Crystal Lake, IL  60014
815-455-3777
Frham Safety Products, Inc.
P.O. BOX 101177
318 Hill Ave
Nashville, TN  37210
615-254-0841
Totes
10078 E. Kemper Rd
Loveland, OH  45140
513-583-2327
Bortek Systems
P.O. Box 355
Royersford, PA   19468
215-948-9696
Nilfisk of America, Inc.
20566 White Bark Drive
Strongsville, OH   44136
          or
224 Great Valley Parkway
Malvern, PA   19355
800-645-3475

-------
Pertex, Inc.
P.O. Box 579
Franklin, MI  48025
313-737-6900
ORR Safety Equipment Company
11379 Grooms Road
Cinti., OH  45242
513-489-0800

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          RADIATION SAFETY AT SUPERFUND SITES (165.11)




                              WORKBOOK




                               CONTENTS






                                                                    Page



RADIATION SURVEY INSTRUMENTS




      1.  Exposure Rate Meters/Dosimeters  	  1



      2.  Count Rate Meters	  3




      3.  Bench Counters	  6



CHARACTERISTICS OF UNKNOWN SOURCES AND DOSE ASSESSMENT	  9



PROBLEM SESSION:  DECONTAMINATION	11




RADIOACTIVE MATERIAL PACKAGING AND LABELING	13




SITE WORK DAY



      1.  Initial Entry and Count Room  	14




      2.  Contamination Survey Station  	-. . 22




      3.  Simple Soil and Water Sampling Protocol	24

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                         RADIATION SURVEY METERS

                      Lab 1: Exposure Rate Meters/Dosimeters

I.     Objectives

      At the end of this lab the student shall be able to:
      A.    Operate a dose rate meter.
      B.    Perform simple dose rate calculations.
      C.    Measure dose using a pocket dosimeter.
II.    Dosimeter Calibration and Use

      Calibration Source	   Type

      Measure dose using a pocket dosimeter.
      Projected Dosimeter Reading	mR

      Dose Rate: Inner Ring	mR/Hr Outer Ring  	mR/Hr


      Start Time   Dosimeter Reading   Stop Time   Dosimeter Reading
III.   Dose Rate Measurement

      Source

            Type 	
             Activity	[j.Ci

      Dose Rate  (w/ASP-1 using HP270 probe)

             Contact      	

             1 in.        	

             6 in.        	

             1ft.         __	
5/93

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                        RADIATION SURVEY METERS
      Calculate the following exposures
                 8 hrs                Iwk            i month        1 year
                                    (168 hrs)         (730 hrs)       (8766 hrs)
 Contact
 lin.
 6 in.
 1 ft.
5/93

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                          RADIATION SURVEY METERS

                             Lab 2:  Count Rate Meters

I.      Objectives

       At the end of this lab the student shall be able to:

       A.    Operate a count rate meter.
       B.    Perform count rate measurements using alpha, and beta/gamma detectors.
       C.    Calculate detector efficiencies relative to source size and distance from detectors.


II.     Count Rates/Detector Efficiency

       Instruments:  Eberline ASP-1 with HP-210L Detector

       Instrument I.D.                     Calibration date
       Background (max.reading over 30 sec.)

       (note:  lei = 2.2 x 1012 dpm)
       Strontium-90 (2-inch diameter)

       Source I.D.	Source Size	Activity (dpm)_

       Meter Reading (slow response, max.reading over 30 sec.)	

       Detector Efficiency (meter reading-bkgd/activity)	
       Technetium-99  (2-inch diameter)

       Source I.D.	Source Size	Activity
       Meter Reading (slow response, max.reading over 30 sec.)_

       Detector Efficiency	
5/93

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                          RADIATION SURVEY METERS

      Technetium-99 (1-inch diameter)

      Source I.D.	Source Size	Activity	

      Meter Reading (slow response, max.reading over 30 sec.)	

      Detector Efficiency	


III.   60 Second Integrated Measurement (integrate response, reading after 1 minute.)

      Technetium-99 (2-inch diameter)

      Source I.D.	Source Size	Activity	

      Expected Integrated Value 	

      Actual Integrated Value	


IV.   Count Rates/Detector Efficiency

      Instruments:  Eberline ASP-1 with AC-3 Probe

      Instrument I.D.	 Background	
       Response to gamma source

       Source I.D.	      Reading	


       Response to beta source

       Source I.D.	      Reading	


       Response to alpha source

       Source I.D.	    Activity	

       Meter Reading	

       Detector Efficiency	

                                                                                         i
5/93                                      4

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                          RADIATION SURVEY METERS




       Thorium-230 (2-inch diameter)




       Source I.D.	   Activity	




       Meter Reading	
      Detector Efficiency
      Instruments:  PAC-4G-3 Gas Proportional Meter




      Instrument I.D.            Calibration date
      Background
      Response to gamma source




      Source I.D.	     Activity




      Meter Reading	
      Detector Efficiency
      Response to beta source




      Source I.D.	    Activity




      Meter Reading	
      Response to alpha source




      Source I.D.	    Activity




      Meter Reading	
5/93

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                          RADIATION SURVEY METERS

                               Lab 3:  Bench Counters

I.     Objectives

      At the end of this lab the student shall be able to:
      A.     Operate bench counter instruments.
      B.     Perform air sample volume calculations.
      C.     Perform half life calculations.


II.    Air Sample Data Sheet

      Sample Date:	         Sample No.
      Time:	
      Location:	
      Sampled by:_
      Air Sample Data

      Air Sampler #:	
      Flow Rate:	cfm
      Time: Start	                                              M
       Stop	
       Sample Duration:	min
Sample Volume Calculation

       (Flow Rate  -=£-) (Sample Duration   min) () (^ = Samplg Voiume
                  nun
                         ft3.  .          .  . ,28.32«, , 1000m/,                 .
                         •*-—)  (         mm) (	—) (—-—)  =             ml
                         min                 i3       (!
Sample Fraction Calculation
                              A.    n r,    (r,¥          ,
              Sample Fraction = -1  =	 = — |  = (      )2 =
                               lr
5/93

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                         RADIATION SURVEY METERS


Volume of Count Sample

             (Sample Volume ml) (Sample Fraction) = Volume of Count Sample (ml)
                                             ) =
                        ml
III.   Count Room Data Sheet
      Count Date:.

      Time:	
      Counted by:.
      Instrument:
      Background:.

      Efficiency:	
      Time of Count


      Count Duration


      Total Counts
                         Count #1
                    Sample No.
                    Serial No.:
           _cpm
          Count #2
      Count Rate


      t
             mm.
Air Activity Calculation


      (Sample  cpm -Background  cpm)l	ICi	\ 11 x 106

          (Sample  Volume~ml)( Eff)    (2.22 x 1012 dpm) (    Ci
                                              uc
                                 = Air Activity  —
                                              ml
                  cpm -
cpm)
106uc
                uc
               (   ml) (0,    )     1,2.22 x 1012 dpm)
5/93

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                       RADIATION SURVEY METERS


Half Life Calculation




                                  A = Ao e~
                                   A
                                   — = e
                                   Ao
                                  tn— =-A
                                    Ao
                                since
                                  Ao
                                      tn (AlAo)
5/93

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      CHARACTERISTICS OF UNKNOWN SOURCES AND DOSE ASSESSMENT

I.     Objectives

      At the end of this exercise the student shall be able to:
      A.     Determine the general area dose rates in the vicinity of the unknown boards and
             establish any necessary stay times.
      B.     Locate all radioactive sources on the unknown board and chart their location.
      C.     Determine what types of radiation(s) is being emitted from each  source.

II.    Exercise

      A.     Initial Survey

             1.      Inspect the dose rate meter (ASP-1 W/HP270 probe)

                    a.     battery check
                    b.     calibration sticker check
                    c.     response check

             2.      Perform a general area survey

                    a.     starting with meter on the  top scale, shift down  in scale until you
                          obtain a reading
                    b.     survey the area, staying 12  inches away from any  object

                    Note: The 12 inches is established in OSHA's regulation on the
                    requirements for a radiation area in 29 CFR 1910.96.

             3.      Document the results of the survey

      B.     Draw a map of the  unknown board

             1.      Note any unusual characteristics of the board

      C.     Survey the board for hidden sources

             1.      Select the survey meter

               Note: We recommend the ASP-1 W/HP260 be used. This probe will conveniently
               scan for  both beta and gamma and will also pick up high-energy alphas; alphas
               which are > 3 meV.
5/93

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      CHARACTERISTICS OF UNKNOWN SOURCES AND DOSE ASSESSMENT           ^

             2.     Inspect the survey meter

                   a.    battery check
                   b.    calibration sticker check
                   c.    response check

             3.     Scan the board  as if you are cutting grass in a manner which will reduce
                   overlaps and minimize large gaps

             4.     Document source locations on the map

      D.     Evaluate each source location for types of emission

             1.     Beta or gamma  or both

                   a.    using the ASP-1 W/HP270 probe, evaluate each of the source locations
                   b.    at each location start with an open window to locate the source and
                         obtain the highest possible reading
                   c.    hold the  probe firmly in place and slide the window shut

                   Note:  Any  decrease in readings can be attributed to beta radiation being
                   shielded out.                                                             m

                   d.    compare  your three readings; open  window, close  window,  and
                         background; you can determine if you are dealing with beta, gamma,
                         or both

                   e.    document the results

             2.     Alpha

                   a.    Inspect the alpha survey meter (ASP-1 w/AC-3 probe)

                         -   battery check
                         -   calibration sticker check
                         -   response check

                   b.    perform  survey with the alpha survey  meter

                   c.    document the results
5/93                                     10

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                          RADIATION SURVEY METERS

                    PROBLEM SESSION:  DECONTAMINATION

Problem Session (Decontamination exercise)

Instructions:  For each scenario,  use your judgement  and the  information provided in the
decontamination lecture to select the method(s) and techniques for decontamination.  All of the
radioactive material involved in the incidents is Uranium-238.  There will be a group discussion
upon completion of the problem session.

Manual Reference Section:  Chapter 7 of Darcom P385-1

      Personnel:

             Methods - Table 7.1
             Procedures (techniques) - Appendix 1

      Equipment and Material:

             Methods - Table 7.2 and Table 7.4
             Procedures (techniques) - Appendix C
      1.     An employee accidently contaminated the back of his/her right hand while removing
             protective gloves after working in a contaminated area.
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                          RADIATION SURVEY METERS

      2.      A 55-gallon radioactive material drum accidently fell off a transport vehicle onto a
             smooth, stainless steel truck scale.  A dry, powder-like radioactive material was
             spilled over a 20-square-foot area.  A response team arrived and uprighted the drum,
             put the  breached drum in an overpack  drum, secured and  posted the  area  as
             contaminated, and transported the overpack drum to a storage area.  The bulk of the
             spilled material is right in the middle of the contaminated area.
      3.     A  laboratory worker accidently spilled some  fine form radioactive material on a
             smooth  porcelain laboratory countertop.   All of the material is contained  in a
             1-square-foot area on the countertop.
      4.     A mechanic working on a contaminated vehicle got some of his tools contaminated
             and wants to have them decontaminated so he can take them back to his shop. The
             tools were a screwdriver, wrench, and pliers.  The tools also became slightly greasy
             from working on the contaminated vehicle.
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              RADIOACTIVE MATERIAL PACKAGING AND LABELING

I.     Objectives

      At the end of this lab the student shall be able to:
      A.     Determine the package type based on curie amount of given radionuclides.
      B.     Given the dose rates of a package  containing radioactive materials, determine the
             appropriate radioactive warning label.


II.    Type Quantity Exercise
      Indicate the quantity type for each of the radionuclides.  All radionuclides are normal form
      unless otherwise indicated.
      Radionuclide        Activity                 Type Quantity

      1.     Ca-14        63 Ci                   	


      2.     Mo-99       1980 mCi               	


      3.     Ra-226      18 Ci (Special Form)     	
1.    Soil samples were taken from an area thought to be contaminated by small amounts of Sr-90
      and Co-60.  The samples will be sent to an off-site lab for radiological assaying. The dose
      rate at the surface of the sample jar was 0.4 millirem per hour and not detectable at 1 meter.
             a. What type of radioactive label would be affixed to the sample package?

                Radioactive Label:	

2.    The developer for a newly planned residential community has been advised by some senior
      citizens that the land in which he planned to build the community on was an old land fill.
      Soil samples were extracted from depths of 1 to 3 feet.  The dose rate at the surface of the
      sample jar was 18 millirem per hour and 1.6 millirem per hour at 1 meter.   The dose rate
      at the surface  of the sample package on the other hand  was 16.S millirem per hour and 1
      millirem per hour at 1 meter.

             a. What type of radioactive label would be affixed to the sample package?

                Radioactive Label:	

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                                     SITE WORK DAY

                          Lab 1: Initial Entry and Count Room                            ™

I.     Objectives

      At the end of this lab the student shall be able to:
      A.     Don and doff protective clothing.
      B.     Post suspected  contaminated area.
      C.     Perform initial entry monitoring techniques.
      D.     Setup count room.
      E.     Perform count  rate measurements.


II.    Post the  Contaminated Area

      A.     String ropes

             1.     yellow and magenta rope or ribbon

             2.     waist high

             3.     sturdy enough to hold  signs

      B.     Put up radiation signs

             1.     select appropriate inserts

                    a.     contamination area

                    b.     airborne radioactivity area

                    c.     radiation area

                    d.     dress out requirements

      C.     Lay down step off pads

              1.     the step off pad is outside of the contaminated area

             2.     ropes do not cross the step off pad

             3.     tape the step off pad to the floor so it doesn't slide around

      D.    Exit  containers

              1.     containers needed for respirators, laundry, and waste

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                                     SITE WORK DAY

             2.     containers need  to be  well supported to handle the  weight of the material
                    placed in them (usually plastic bags are used as containers)

      E.     Set up a decon area

             1.     usually a table top will be sufficient

             2.     place a protective covering on the table to prevent contaminating the table

             3.     establish three "zones" on the table

                    a.     contaminated area

                    b.     to be surveyed area

                    c.     clean area

             4.     set out decon equipment

                    a.     absorbents

                    b.     decon spray

                    c.     soap and  water

      F.     Step off support area

             1.     bags should be available to receive contaminated material being brought out
                    of the contaminated area

             2.     frisking station

             3.     staging of additional supplies; extra smears, absorbent, tape, etc.


III.   Set up the Count Room

      A.     Select instruments

             1.     alpha frisker  (ASP-1 w /AC-3 probe)

             2.     beta/gamma frisker (ASP-1 W/HP210L probe)

             3.     alpha bench counter  (SAC-4)


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                                    SITE WORK DAY

             4.     beta/gamma bench counter (BC-4)                                       U

      B.     Perform instrument checks

             1.     battery check

             2.     calibration sticker

             3.     efficiency determination

                    a.      use a 230Th source for the efficiency determination of alpha detection
                           instruments

                    b.      use  a  "Tc source  for the efficiency determination of beta/gamma
                           instrument

      C.     Lay out the count room into areas for handling both clean and contaminated items.
             The following needs  should be addressed:

             1.     smear/air sample handling area

             2.     clean area for notes and maps

             3.     clean waste bags                                                       ^

             4.     contaminated  waste bags

      D.     Assemble all of the  support supplies which  should be available at this station, the
             following items  should be considered:

                tape       paper         clipboards
                tweezers   pens         calculator
                bags       blotter paper  calculation forms
                gloves      paper  towels  source sets
                smears     decon spray   rad rope
                clock      envelopes    rad signs

IV.   Prepare Survey Equipment

      A.     Map

             1.     draw a map of the area

             2.     if the area  has not been seen yet, prepare the materials required to draw a
                    map during the survey (clip board, paper, pen, etc.)

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                                     SITE WORK DAY



      B.      Contamination survey materials




              1.     number smears




              2.     prepare envelopes




      C.      Radiation survey materials




              1.     select instrument




              2.     inspect instrument




                    a.     battery check




                    b.     calibration check




                    c.     response  check




      D.      Air sample materials




              1.     load filter into the air sampler




              2.     prepare air sample envelope




              3.     have a bag ready to remove the air sampler from the contaminated area




V.    Anti-Contamination Clothing




      A.      Select anti-c's




              1.     minimum dress is one full set of anti-c's




                    a.     plastic shoe covers




                    b.     cotton glove




                    c.     coveralls




                    d.     hood or skull cap




                    e.     heavy shoe covers




                    f.     rubber gloves




                    g.     tape gloves and plastic shoe covers




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                                     SITE WORK DAY

             2.     add additional anti-c's based upon need                                        I

                    a.     wet

                    b.     high contamination

                    c.     airborne contamination

      B.     Don anti-c's (in the same order listed in step 1 above)


VI.   Enter Area and Perform Initial Entry Surveys

      A.     Start the air sampler

             1.     enter area taking  dose rates

             2.     move to the place where the air sample is to be drawn

             3.     lay down a clean  cloth or piece of plastic; away from the floor

             4.     set the air sampler down on the cloth and start it

             5.     record start time

      B.     Take radiation survey of the rest of the room

             1.     take radiation  readings

             2.     record dose rate

                    a.     general area dose rates are taken  12 inches from any surface and at
                           waist level
•
                    b.     contact dose rates are marked with an asterisk (*)

             3.     remove the radiation survey instrument from the contaminated area

                    a.     if the radiation instrument has not come into contact with anything in
                           the  contaminated  area  it can be placed in  the  section marked "to be
                           surveyed"

                    b.     if the instrument has come into contact with something it must be
                           wiped down before it is placed in the "to be surveyed" area


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                                    SITE WORK DAY

      C.     Perform a contamination survey of the room

             1.     scan the room for high probability contamination locations and high traffic
                    areas

             2.     take smear samples in those areas

                    a.     each  smear should be  ~ 100 square  centimeters,  or 4 inches by
                          4 inches square

             3.     mark the smear  locations on the map using sequentially numbered  circles
                    (i.e., 0, ©
             4.     ensure each  smear is physically separated from the others to prevent cross
                    contamination

                    a.      folding smear papers

                    b.      individual smear envelopes

                    c.      paper dividers

             5.     place all of the smears in an envelope

             6.     pass the envelope out of the area

      D.     Secure the air sampler

             1.     turn the air sampler off and record the time

             2.     using  a clean  pair  of gloves, remove the air  sample and place it  in  its
                    envelope

             3.     pass the air sample out of the area

             4.     reassemble the air sampler and remove it from the area

                    a.      wipe down the air sampler

                    b.      place air sampler in the "to be surveyed" area
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                                   SITE WORK DAY

      E.      Exit the contaminated area

             1.     Remove anti-c's

                   a.      all tape

                   b.      rubber shoe covers

                   c.      rubber gloves

                   d.      hood

                   e.      coveralls

                   f.      plastic shoe covers

                   g.      cotton liners

      F.      Frisk

             1.     go to the nearest frisker and perform a whole body frisk

                   a.      probe should be held 1A - l/i inch from surface being monitored

                   b.     move probe at 2 - 3 inches/second

                   c.      entire frisk should last 2-3 minutes

                   d.     pay particular  attention to high probability areas such as hands and
                          feet, elbows, knees, face, or any area which was uncovered


VII.   Evaluate the Smears

      A.     Count the smears

             1.      do a field evaluation of the smears for both alpha and beta/gamma using
                    friskers

             2.      document the results in both cpm and dpm

             3.      if the smears have very low levels of contamination on them, a frisker may
                    not show it; therefore, you may wish to bench count the smears



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                                    SITE WORK DAY

                    a.     bench counters are more accurate for determining the exact number of
                          emissions from a source over a specific period of time

                    b.     bench counters allow you to extend the count time which dampens the
                          effect of the random nature of radiation emissions

      B.     Count the air sample

             1.     mark the air sample size using a 2 inch planchet and pen

             2.     cut the air sample out with a pair of scissors

             3.     count the air sample for alpha contamination using  a one minute count time

             4.     record the alpha counts and calculate the curie content in microcuries per
                    milliliter

             5.     count the air sample for beta contamination using a one minute count time

             6.     record the beta counts and calculate the curie content in microcuries per
                    milliliter

      C.     Survey the equipment which must be removed from the contaminated area

             1.     take smears on each piece and  count the smears on both alpha and beta-
                    gamma friskers

             2.     if no contamination is found on the smears, pick up each piece and frisk it for
                    fixed contamination


VIII.  Break down the areas and put the equipment awav
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                                    SITE WORK DAY

                          Lab 2:  Contamination Survey Station                                *

I.     Objective

      At the end of this lab the student shall be able to:
      A.     Detect very low levels of contamination.

II.    Exercise

      A.     Perform instrument preparation (RM-14S or ASP-1 W/HP260 probe)

             1.     calibration check

             2.     battery check

             3.     high voltage (900 volts)

             4.     response check

             5.     measure (30 sec.) background radiation and record result

      B.     Make a drawing of the item to be surveyed                                           ^

      C.     Frisk low contamination boards

             1.     mark  all areas  on  the drawing that  are  contaminated  (100 counts above
                    background)

             1.     turn item over to verify areas of contamination

             3.     refrisk any  areas that are contaminated but were not detected

      D.     Frisk high contamination board

             1.     mark all areas on the drawing that are contaminated (100 counts above your
                    background)

             2.     turn item over to verify areas of contamination

             3.     refrisk any  areas that are contaminated but were not detected
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                                    SITE WORK DAY

      F.     Frisk articles

             1.     boxes

                    a.     mark all areas on the drawing that are contaminated (100 counts above
                          background)

                    b.     open box to verify areas of contamination

                    c.     refrisk any areas that are contaminated but were not detected

             2.     flashlights

                    a.     mark all areas on the drawing that are contaminated (100 counts above
                          background)

                    b.     do not disassemble flashlights

      G.     Answer the following questions.

             1.     What may account for the sporatic increases in cpm when a probe is not in
                    contact with contamination?
             2.     Is it always appropriate to rely solely on your meter reading and not the audio
                    speaker  of a detector when surveying for  contamination?   Explain  your
                    answer.
             3.     What may cause a meter reading greater than 100 counts above background
                    but no counts above background based on smear data?
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                                    SITE WORK DAY                                    4

                     Lab 3: Simple Soil and Water Sampling Protocol

I.     Objectives

      At the end  of this lab the student shall be able to:
      A.     Perform soil  sampling.
      B.     Perform water sampling.
      C.     Set up radiological area when a contaminated sample is discovered.


II.    Soil Sampling

      A.     Purpose

             1.     To determine whether the soil is radioactively contaminated.
             2.     To characterize the radioactive components of the soil by laboratory isotopic
                    analysis.
             3.     To effectively control the radiological hazard.

      B.     Soil sampling preparation

             1.     Protective clothing                                                    ^

             2.     Minimum protective clothing requirements shall include rubber shoe covers
                    and waterproof gloves

             3.     Monitoring and survey equipment

             4.     Portable alpha probe instrument

             5.     Portable beta-gamma probe instrument

             6.     Portable dose rate instrument

             7.     Smear paper  (swipes,wipes)

             8.     Small plastic  bags

             9.     Pen and data sheet for documentation

      C.     Sampling equipment and supplies

              1.     Sampling apparatus (i.e., core drill)

             2.     Sample container(s) (i.e., 1-liter polyethylene bottle with cap)

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                                    SITE WORK DAY

             3.     Metal tray

             4.     Medium mesh screen (sieve - about 1 ft. by 1 ft.)

             5.     Trowel or spoon

             6.     Plastic bags (medium and large)

             7.     Plastic sheeting

             8.     Absorbent materials (i.e., cotton rags or diapers)

             9.     Adhesive tape

             10.    Mild soap or detergent

             11.    Bucket of water

             12.    Barrier rope or tape, and stanchions

             13.    Appropriate work site posting (i.e., EPA work area)

             14.    Knife or scissors

             15.    Yellow and magenta radiation rope

             16.    Underground Radioactive Materials signs

             17.    Do Not Enter Without Approval signs

             18.    Surface contamination signs

      D.     Simple soil sampling procedure

                1.   Don protective clothing

                    a.     Put on minimum shoe covers and gloves

                2.   Set up work area

                    a.     Take area and surface readings to ensure no surface contamination or
                          exposure dose rates are  significant (Ref: US EPA SOSGs)

                    b.     Establish perimeter  work area


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                                    SITE WORK DAY

                   c.     Install barriers and posting (EPA work area)

                   d.     Spread plastic sheeting over sample location, and cut hole in exact
                          location where sample will be drawn

                   e.     Set up staging area and place equipment near sample hole

                   f.     Set up small cleaning area on plastic  sheeting (about 2 ft. by 2 ft.)
                          Surround with absorbent material

                   g.     Place bucket of water, detergent,  large  plastic bag,  and some
                          absorbent material near cleaning area but not inside cleaning area

             3.    Prepare to obtain sample

                   a.     Prepare sample apparatus to draw sample

                   b.     Ensure monitoring instruments are near sample hole and operating

                   c.     Have dampened absorbent material handy

             4.    Obtain sample

                   a.     Draw sample and monitor with portable contamination instruments
                          while sample is being obtained (Primarily beta-gamma,  spot check
                          with alpha)

                   b.     If background levels increase, it  is a good indication the sample is
                          radioactive. Use dose rate instrument to monitor and continue to draw
                          sample.

                   Note 1:  If exposure rate increases to 3-5 times above background, work can
                   continue, but a health physicist should be consulted.

                   Note 2:  If exposure rate reaches 1 mR/hr or above, stop work and consult
                   with a health physicist. If not, continue as follows.

                   c.     Continue to draw sample while carefully wiping sampling apparatus
                          with dampened absorbent material. Place absorbent material in large
                          plastic bag as waste.

                   d.     Place sample on metal tray  and spread  out into very thin layer.  If
                          sample is too chunky, break  up  with trowel or use medium mesh
                          screen to sift sample onto metal tray.


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                                    SITE WORK DAY

                    e.     Use beta-gamma and alpha contamination instruments to monitor soil
                          sample, and document results.

                    Note 3:  If there is no indication of detectable contamination, work can
                    continue as normal (non-contaminated).

             5.     Treatment of contaminated sample

                    a.     If contamination is detected, carefully transfer soil sample on tray into
                          sample container using trowel and cap container.

                    b.     Transfer sample to cleaning  area.   Carefully change  gloves.  Dip
                          absorbent material in bucket of water, slightly wring but leave wet and
                          decontaminate outer surfaces of sample container inside cleaning area.
                          Dry outer surface of container using dry absorbent  material.

                    Note 4:  Place wet and dry absorbent material into large plastic waste bag
                    after each  single use.

                    c.     Survey  outer  surface of container  for loose surface contamination
                          using smear paper and  checking  smear paper  with  portable  beta-
                          gamma and alpha instruments. If outer surfaces remain contaminated,
                          repeat step b., and resurvey.  Document results.

                    d.     If outer  surfaces are contamination free, place sample in medium
                          plastic bag,  tie off  top, and tape  seal.   If outer surfaces  remain
                          contaminated after decontamination, place sample in medium plastic
                          bag, tie off top, tape seal, and repeat using second medium plastic bag
                          (provides double containment).

                    e.     Ensure outer surface of plastic bag containment is contamination free,
                          and prepare  sample for  shipment to  analytical  laboratory using
                          prescribed radioactive materials shipping  methods.

             6.     Termination of soil sampling work

                    a.     Monitor unused materials to verify  they are contamination free, and
                          remove from work area. Document results.

                    b.     Monitor remaining materials, and decontaminate as necessary.  Those
                          that can be verified as contamination free shall be removed from work
                          area. Those that remain contaminated shall be  disposed of in large
                          plastic waste bag, or sealed in  large plastic bag for subsequent use.
                          Document results.
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                                  SITE WORK DAY

                   c.     Carefully roll or fold up plastic sheeting and dispose of in large plastic
                         waste bag

                   d.     Perform surface contamination monitoring of entire work area surface.
                         Contaminated dirt can be placed in large plastic waste bag. Document
                         results.

      E.     Securing the work area/Setting up radiological area

            1.     Setting up perimeter boundaries

                   a.     Remove the work site perimeter barriers and postings.  Monitor to
                         ensure they are contamination free

                   b.     Replace perimeter barriers with yellow and magenta radiation rope

            2.     Classify area based on survey results

                   Note 5: The remaining contained contaminated items can be secured and left
                   inside the area for  subsequent site use or appropriately shipped for proper
                   disposition.  In any case, the outer surfaces of bags shall be contamination
                   free.

            3.     If nothing is left inside the boundaries and the ground surface is contamination
                   free, the area can be posted as  1-UNDERGROUND RADIOACTIVE
                   MATERIALS  and 2-DO NOT ENTER WITHOUT  APPROVAL.

            4.     If  contained   plastic bags  are  left inside  and  the ground  surface is
                   contamination   free,  the  area  can  be posted  as  1-RADIOACTIVE
                   MATERIALS  AREA, 2-UNDERGROUND RADIOACTIVE MATERIALS
                   and 3-DO NOT ENTER WITHOUT APPROVAL.

            5.     In addition to a. and b.  above, if the ground surface remains contaminated,
                   it will be classified as a surface contamination area and appropriate posting
                   will be  required (ie.  CONTROLLED SURFACE CONTAMINATION
                   AREA).

                   Note 6:  All postings shall be installed about waist  high.  All postings shall
                   be visible from all accessible sides to the site.

      F.     Final monitoring

            1. Remove protective clothing,  place in plastic waste bag, and seal

            2. Perform whole body personnel monitoring prior to leaving work site

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                                    SITE WORK DAY

      G.     Considerations

             1.     Be conscientious of cross contamination at all times

             2.     Change gloves regularly, especially when performing different tasks (consider
                    wearing two pair of gloves)

             3.     Work carefully to prevent the spread of contamination

             4.     Monitor yourself whenever you suspect you may have become contaminated

             5.     Bring extra protective clothing, such as rubber shoe covers, in the event they
                    get ripped, torn, or otherwise rendered useless

             6.     Dispose of all used smear paper  as contaminated waste, or place in  small
                    plastic bags to go to laboratory for further analysis

             7.     Ensure no free standing water is noticeable in contaminated waste bag(s)


III.   Water Sampling

      A.     Purpose

             1.     To determine whether the water is radioactively contaminated

             2.     To characterize the radioactive components of the water by laboratory isotopic
                    analysis

             3.     To effectively control the radiological hazard


      B.     Water sampling preparation

             1.     Protective clothing

                    a.      Minimum protective clothing requirements  shall include waterproof
                           shoe covers (preferably boots) and  waterproof gloves (preferably mid-
                           forearm length)

                    b.      Waterproof apron

             2.     Monitoring and survey equipment.

                    a.      Portable alpha probe instrument

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                                    SITE WORK DAY

                    b.      Portable beta-gamma probe instrument

                    c.      Portable dose  rate  instrument (water may  be suspected  of being
                           infested with debris)

                    d.      Smear paper (swipes,wipes)

                    e.      Small plastic bags

                    f.      Pen and data sheet for documentation

             3.     Sampling equipment and supplies

                    a.      Sampling apparatus (optional depending on type and depth of desired
                           sample, i.e., extension pole equipped with sample bottle stopper)

                    b.      Sample container(s)  (i.e., 1-liter polyethylene bottle with cap)

                    c.      Plastic  bags (medium and large)

                    d.      Plastic  sheeting

                    e.      Absorbent materials (i.e., cotton rags or diapers)

                    f.      Adhesive tape

                    g.      Mild soap or detergent

                    h.      Bucket of water

                    i.      Appropriate work site posting (i.e., EPA work area)

                    j.      Yellow and magenta radiation rope,  and stanchions to hold rope

                    k.      Do Not Enter Without Approval signs

                    1.      Radiological Control Area signs (e.g., radiologically contaminated
                           water,  surface  contamination)

      C.     Simple water sampling procedure

             1.     Don protective clothing

                    a.      Put on  minimum shoe covers, gloves, and apron                          4


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                                    SITE WORK DAY

             2.     Set up work area

                    a.     Take area and surface readings to ensure no surface contamination or
                          exposure dose rates are significant.  (Ref: US EPA SOSGs)

                    b.     Establish perimeter work area, and install posting (EPA work area)

                    c.     Spread plastic sheeting near sampling location

                    d.     Set up small cleaning area on plastic sheeting  (about 2 ft. by 2 ft.).
                          Surround with absorbent material

                    e.     Place bucket of  water,  detergent,  large plastic  bag, and  some
                          absorbent material near cleaning area but not inside cleaning area

                    f.     Set up staging area near  sampling location  and place remainder of
                          equipment

             3.     Prepare  to obtain sample

                    a.     Ensure  monitoring  instruments  are  near  sampling  location  and
                          operating

                    b.     Have dry absorbent material handy

                    c.     Have open sample bottle and cap handy

             4.     Obtain sample

                    a.     (1) For surface sample:  skim top of water surface using open bottle,
                          collect sample, and cap bottle

                          (2) For underwater sample:  turn bottle upside down, submerge below
                          water surface (being careful not to let water level reach top of glove),
                          turn bottle right side up and collect sample, bring out of water and cap
                          bottle

                    b.     Use dry absorbent material to  dry outer surfaces of sample bottle.
                          Place absorbent material and bottle on plastic sheeting,  change gloves

                    c.     Survey sample bottle for contamination

                          (1)  If no contamination is detected by dry smears or direct readings
                          on sample or any used materials, you can assume  materials to be


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                                    SITE WORK DAY

                           clean. Occasionally spot check for contamination, if none is detected,       i
                           continue as normal.

                           (2)  If any contamination at all is detected on outer surfaces of bottle,
                           dispose  of smears  and used absorbents in  large  plastic  bag  as
                           contaminated waste. A direct reading increase with beta-gamma probe
                           may indicate sample itself is contaminated.  Document results.

                    d.     If sample container is found to be contaminated on the outer surfaces,
                           it must be decontaminated.  Transfer sample to cleaning area, change
                           gloves, and prepare to decontaminate.

                    e.     Use a little detergent in  the bucket of water  and dip  some absorbent
                           in it. Hand wring over bucket, then wipe outer surfaces of container.
                           Dry  and  resurvey  for  contamination.   Dispose  of  smears and
                           absorbents as contaminated.

                    f.     After outer surfaces are found to  be contamination free  using dry
                           smears,  decontamination is complete.  Change gloves.  Increase in
                           radiation level from a direct beta-gamma reading on the side of the
                           sample would indicate sample itself is contaminated.  Place sample in
                           medium plastic bag and seal bag  using adhesive  tape.   Document
                           results.                                                                I

                    g.     Ensure outer surfaces of plastic bag containment is contamination free,
                           and prepare  sample  for  shipment to  analytical  laboratory  using
                           prescribed radioactive materials shipping methods.

             5.     Termination of water sampling work

                    a.     Monitor unused materials to verify  they are contamination free, and
                           remove from  work area.  Document results.

                    b.     Monitor remaining materials,  and decontaminate as necessary.
                           Those that can be verified as contamination free  shall be removed
                           from work area.  Those that remain contaminated shall be disposed of
                           in large plastic waste bag, or sealed in large plastic bag for subsequent
                           use. Document results

                    c.     Blot wet  areas  on plastic sheeting  using  dry absorbent material
                           Carefully roll or fold up plastic sheeting and dispose of in large plastic
                           waste bag

                    Note 7:  If any free standing water is noticeable in contaminated waste bag,
                    place some extra dry absorbent material in bag to absorb water.

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                                    SITE WORK DAY

                    d.     Perform surface contamination monitoring of entire work area surface
                           to ensure no residual contamination is left on the ground
                           Document results

             6.     Securing the work  area/Setting up radiological area

                    a.     Install yellow and magenta radiation rope around entire water body
                           being sampled

                    b.     Post appropriate radiological control area signs
                           Post Do Not Enter Without Approval signs.  Signs should be about
                           waist high,  and seen from all  accessible approaches.

                    c.     Secure  contaminated  waste  inside posted area, or properly transport
                           for disposal as contaminated waste

             7.     Final monitoring

                    a.     Monitor protective clothing, remove and properly dispose

                    b.     Perform whole body personnel monitoring prior to  leaving work site

             8.     Considerations

                    a.     Be conscientious of cross contamination at all times

                    b.     Change gloves regularly, especially when performing different tasks,
                           consider wearing two pair of gloves.

                    c.     Work carefully to  prevent  the  spread  of contamination  and be
                           conscience of where potentially contaminated water is dripping

                    d.     Monitor  yourself whenever  you suspect  you  may have  become
                           contaminated

                    e.     Bring extra protective clothing, it might come in handy

                    f.     Do  not leave any free-standing water in contaminated waste bag
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U.S. Environmental Protection Agency
Region 5, Library  -L-12J)
77 West Jackson Boulevard, 12th Floor
Chicago,  IL  60604-3590

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