ACCN
TYPE
DB
MTI
TRACED
MYDATE
CALLNO
BRANCH
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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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
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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
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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
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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
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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. .
1
S
1
S
1
S
1
S
1
S
1
S
1
S
Sub
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
1
Tib
Col. 1— Air
OiCi'ml)
2- 10"
2- 10- "
1 . 10"-
5. 10- "
1 . 10"-
2. 10- "
1> 10""
6 . 10- "
3\IO-"
8. 10"-
8. 10'1-
5. 10-"
5. 10'"
4 . 10-
5- 10-
4 . 10'
2- 10-
3 • 10
.2- 10'
1 • 10'
6- 10-
1 • 10*
3. 10"
<-'io-»
6- 10"
4. 10"
1 . 10'
5- 10-
9- 10
4 • 10-
2- 10
6- 10"
1 • 10"
4 • 10 '
2- 10"
3- 10"
2- 10"
1 • 10"
1- 10"
3 • 10"
2- 10''
2 • 10"
9- 10"
6 • 10 '
S- 10"
3. 10' '
9 • 10"
2-10"
1 • 10 •
1 • 10'"
2- 10 -
• • 10'"
1 . 10'"
• • 10'"
1 • 10" "
5- 10'"
1 • 10'-
5- 10'"
1 • 10" ~
5 . 10 "
1 • lO'"
6-10 "
1 . 10 "
1 • 10 '
1 . 10 »
1*1
Col. 2-
W»w
[HO'ml)
1 • 10"
1 • 10"
7 • 10'
4 . 10"
7 • 10"
1 • 10'
»» 10"
2> 1C'
2. 10-
4 . 10-
4 > ID'
4V 10"
4 . 10-
2 • 10'
,:
3 -10
3 . 10
1 . 10
1 • 10-
3 . 10"
3. 10"
7 . 10'*
3 • 10'
2 . 10-
3. 10-
3- 10-
1 • 10"
3 • 10-
7 . 10
2 . 10'
2 . 10-
4 . 10-
1 . 10'
?. 10
2 . 10'
1 . 10'
1 . 10'
5 . 10-
$ . 10
2 . 10-
1 . 10'
a. 10-
6. 10
4 . 10
3 • 10'
1 . 10'
1 . 10
1 . 10'
6 . 10
7 . 10
7 . 10
1 . 10'
7 . 10
2 • 10
t . 10
1 . 10"
». to-
1 . 10
* . 10
1 . 10
« . 10
1 . 10
4 . 10
6 • 10
6 • 10
Tit*
Col. 1— AJ>
(pG'ml)
6- 10"
5. 10"'
3. 10"'
2- tO"'
3- 10"'
6v10"'
3 1. 10- "
2> 10"'
1x10- '
3 >• 10- '
3.: 10" '
2 .-. 10- '
2- 10- '
1 . 10 '
1 . 10 •
2. 10 *
S. 10"
9. 10"
7. 10"
3- 10- "
2- 10-"
4. 10"
1 • 10"
1 • 10"
2- 10"
1 • 10"
i. 10- "
2. 10"
3. 10"
t. 10"
6. 10"
2- 10"
S. 10'~
1- 10"
• • 10 "•
9. 10"
7. 10"
4. 10"
«. 10"
1 • 10"
«. 10"
«• to"
3. 10"
3 . 10"
2- 10"
t • 10"
3. 10- "
7. 10"
4 . 10"
4 . 10 "
6 . 10 "
z • tc' ••
3. 10'"
3- 10- •
3 • 10' '
2. 10 •
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4. 10
2- 10 -
4 . 10''
J. 10 '
4 . 10 •
4 ' 10
4 . 10
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Col. 2-
Wtiw
IvO/ml)
3. 10"
4. 10-
2- 10-
1. 10-
3> 10-
4^10*
3x10-'
7x10"
7vlO"
1 <10-1
1x10"
1« 10-'
1 »-IO"
a.-to'-
9 . 10"
». 10"
4 . 10"
4 . 10"
1 . 10"
1 . 10"
2 • 10"
«. 10"
S-..10"
1 . 10"
9 .' 10"
4 /lO"
1 . 10"
2- 10"
9- 10-'
6 • 10-*
2- 10"
4 . 10"
B. 10"
6-10"
4 . 10"
4 • 10"
2-10"
2. 10"
$.10"
4...10"
3. 10"
2- 10-'
t * 10'*
9--. 10- »
5- 10-'
3 . 10"
3. 10"
2-10"
2- 10"
2. 10"
S- 10"
2- 10"
7 • 10-
3- 10-
4 . 10-
3*10-
4 .. 10-
3- 10
4 . 10-
2- 10-
4 . 10
1 . 10
? • 10
2 • 10
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
Taetm-12Sm..
10
100
100
10
100
100
100
100
10
too
10
10
10
10
100
100
100
too
100
10
100
100
100
100
too
.01
0.1
10
100
100
10
10
.01
100
100
too
100
10
10
100
to
10
1
10
100
10
100
10
10
100
10
1
100
10
10
1
O.I
10
10
100
10
10
100
too
100
10
10
TaUunum-129
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
IOO
10
too
10
10
10
100
100
IOO
10
100
10
10
10
10
10
10
IOO
100
.01
.01
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
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5/93
TABLE 2
EXCERPT: 10 CFR 20, APPENDIX B
11
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
5/93 14
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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
-------
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
'888
1:
f- i
I?'.l£
:FF'
BI
: i if J I! J
; 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
» »°-^feS8^ox 5.
iS5ss|^Ssi i-
Q >-,
t|I * 23s§
°«? =t:».,,
.vMu.j:uv«i<>
•o c Sjc S "i «^/:jC .. - s 3 ox: i
rf.E Ki.&oTcisiitSojjSg-'"!
S g 3cg2>J5^wc.og"».c.
!>.£
I I
8? S
™S «
•^ A
?*a|5|*gr|
|s°'l|§6ei
S -J3 « C£B E fei»
fegS-s^au,
a
«!! il^i* 1 ^lll!fl!l^l!PiH'41l!l!l
>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
-------
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
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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
-------
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
-------
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
-------
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
-------
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
-------
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
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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".
-------
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.
-------
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.
-------
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.
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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.
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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)
JH ;l2.3y):
I: 1/2 atomic .peel, M: <2.f7t>5 NMR (Lj«d{164)
p*: O.Om mat ipect (PorlFS0)
othtr. (Lan;eL52, CurrS4!>. Ham!DS3a, HannG4°)
.. 5730 y
(MTW)
N
(5730 y):
I: 0 atomic «p«ct
mierowav*
; O.liS m»j ipcct (F«lM9a, WarlSJO, ForH54); ion ch (AnjJ49)
O.IS4 (CookCSXM); 0.1 J4 (L«vyP47); 0.15? (PohAJS); mag >p«ct
oOi«r« (M01AS4)
no conv, mag cpect conv (l^evyP47): no Y (RubeSJl)
2.60 y
EC '•"
(MTW)
22,
. (2.60 y):
1: J. M: < 1.744 atomic bum (Unp«ct (WriB53)
other. (CeoW4o. Mor|K<9. LcuHil, BranWo4*. CharPoS)
Y: Y. 1.1744 itmicond ipcct (RobtR6S)
Y, («/Y6.7» IO"^(NaVY43, l.«amR54)
other. (MarlKoS. SinPS9. AlbuD4?. AJtFSS. OooW4o)
PY(6): (Gr4bZ65. DaniHoO*. SubBolb, StevDSl, MullH6S)
DlpcUrii(e): (Stem!?. BloS42, AppHi?. Bh»S65, SchoHST)
(MackPiOa); O.J40 (WonC54);
-------
226 R0
1602y
2.C5rr,
2c.S rr
„ i 4.60 MeV (5%)
\ff\ 4.78 MeV (95%)
19.7m \
y 1C. 19 MeV (4% + 1C)
5.49 MeV (100%)
6.00 MeV (100%)
0.69 MeV (—47%)
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:
%
r i
1
•-».
*--»
i
\ \< 2 MeV (~76
^J3.26MeV (~19
0.61 MeV (47%)
, 0.77 MeV (5%)
0.93 MeV (3%)
1.12 MeV (17%)
^ y 1.24 MeV (6%)
TOO U_ V fCOf\
l.j o MeY (s/e)
1 76 MeV (17%)
^2"20 MeV (5%)
2.44 MeV (2%)
21
214Po (RoC1)
Stable
17.69 MeV (100%)
1 MeV (81%)
0.06 MeV (19%)
0.05 MeV (4% 4 1C)
(1.16 MeV (100%)
4d
CL \ 5.31 MeV (100%)
-------
5.26y
60r. >F><" *•"
oyUO /A\/^. 4. - MM?
y\
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(S.J6 y):
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f,, 1.41 (0.12%) (CunMI)i '••" (0.010%) fWol(.J46); 1.41 (0.15%)
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•l! 4 tj (c*/T 0.004) T*Y« eoine (l«fhH6U)
-i, 3.UI (Y 0.0012%) m>| ipcet (WoWtJit)
j'i (-0.0000<%) D--T-I. (MorlHJf)
oiKcr. (AveM5E. Ur«liCi3. Hpm\V4f. KI«mE53. AtpHSii, Ch»tS53
L*-'JS53. L*mH5«, \VicT»«, ColoSii, DihBS 1 . Si<|K)Ot)
YT(»): (0»rsJ6£l. JJr»dESC, KloRii, Ch»tS53. KUmCSJ, L.wJSS3, Wi«T44)
YTpol4rii(0): (MctrSD, WUiiAHSC. KloRSJ)
): ., lx>kV4Zt. C»r««49, AD.R5D. B.yiWOi.. KovtTJO.
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pypol.ri.(O): (J»rP*C, BteStZ. AppWS. IxibVS?. StcIW?, P>(LSt.
i, Dp«cl (HorrMJ)
0.0(2 «)>etrtiiullc »iv»ly«r (
C.073 >b>, ten ch (ME.w»J5T)
0.0(3 ion ch (W1UH4V)
V: lie r (Wl)tvm, Bro.Ail)
-------
392
89
Qf- i.*n (MTU*) 3gSr
i ti 16s
'* Sr |5Z d).
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other* (SUl^ri, R«|W^7j
T- no > (NovcTSl. SlewD37. Sle»
C »I3 (wlU. **m Y) (HerrmCit
5id). m*i ipcct
D39)
LyoWJSb, i«lA67)
Y. Y, O.tOI icUt >p«ct I
( iptcl
conv (Shoril); O.tli »cuu «p«cl |Monli61, H«miJ40;
^1 («/r 0.01) mk( .p.ct eonv, >cuu •?«« (ShuKSl, OolilhMSl)
i... 7) m«j »ptct COBV (B
2B.ly
90
40
Zr
-62 ns
;Sr(2M
•U»r>
t: M t »k. (C)*LS)e)
' T (*< h):
•^ 2, ),: -C.U3 atomic beur.
t~: 1.311 (Andr$«4. JohnOSS). 2.271 (NitRol). 2.2*4
2.273 (LAnfel^64a) m*f cocci '
olhcri (Y>i»Ti7b. BerlUe, l^»nj«Mf. M«jJ3Z, JohnOiS,
l^iLiO. N«1T56)
• peel (YiuTSti, ru.TJ7»)
T, 1.7J (t~ C.SV e/t very Urfe) m«x ipcel eonv (JohnOJS)
1, (tyr/lc- <0.000«) m.t .ptcl conv (RvdH63bJ
other* (CoroSoU. CoroSolI. L»nChH61. AlbuDiS. RydHol,
Dc»M57, CreeJ»6)
-------
7/J,
S30.0y
2.55m
137
Ci (30.0 y):
1: 7/3. v 42.I3>2 atomic btatn; q: +O.OSO ppt double rci (I,u,ejl6<)
p*: A, 1.17< (6.2%). Pj O.JU (73.4%) ma, /Y 0.1)00) .wnlcetw! .p«cl (M«rJ6S)
Y, (lv/ly*K 1C"*) (»«iW«0)
ot>.«ri
Sub
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.. Cr.vCSZ. l
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. OrabCSS.
-------
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C*SS
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other. (U.rtiWii,
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1.5 m (SutDSS)
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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,
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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)
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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)
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1
¥ p4 81%, EC 19%
(MUKA58)
f," -fcO%, EC -40%
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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)
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ch*m. K»cii, cross porno
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i i
i
1 i
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I i
1 J
f } 1.41 max (0.12%), 0.114 ma*
, | f»»«%)
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MrSS(a,n) (UvJMa,
UvJ41)
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-------
266
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Sr»°
Sr"
Sr"
Sr
Sr*
Half-life
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(GeldmDTM)
1
% 7.02 IKKrAJIb.
AldL5)j
A -»V»t5 (MTW)
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(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)
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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)
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: 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)
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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)
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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
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OCCUPATIONAL SArETY AND HEALTH
l»10.M
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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
-------
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
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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
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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
r ee>n>mu*f»e? o' * 0*fmn DI Ucami* bv t*>< Comfvvttto
9Ui6*t *•>.-l tx trviM< p*MO0'C*t'T at *pp/
to •ccommoon*
Coow o^ pu&hth«d ywiOrt m*y be otniirwd bv r*owMt (ftdiating the divittom
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
' M*t»f,-*ii jno Pi»nt ^roirctton
&. 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
-------
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
-------
-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
-------
- 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
-------
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
-------
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
-------
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).
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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)
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
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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
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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
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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
1.27
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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
-------
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
1.30
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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
1.35
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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
-------
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.
1.37
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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
1.38
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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.
1.39
-------
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.
1.40
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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
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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
-------
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
-------
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
-------
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
-------
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^
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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).
2.19
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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
2.21
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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
2.22
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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
2.24
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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.
2.25
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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.
2.26
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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
2.29
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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.
2.30
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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
2.31
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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
2.32
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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
2.33
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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
2.34
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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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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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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
2.44
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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.
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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
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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.
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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.
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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
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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.
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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
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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
3.10
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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
3.12
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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.
3.13
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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
3.14
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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.
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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.
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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
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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
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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.
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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
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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
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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
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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
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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.
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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
4.16
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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.
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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
4.18
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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
4.19
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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
4.21
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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
4.22
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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
4.24
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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.
4.26
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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:
4.27
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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.
4.28
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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
4.29
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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.
4.30
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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.
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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
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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
-------
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
-------
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.
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
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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
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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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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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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
-------
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
-------
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
-------
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.
8.14
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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
8.16
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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
8.17
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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
8.18
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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
8.19
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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.
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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:
8.23
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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-
8.24
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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
8.27
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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.
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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.
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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.
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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
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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
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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.
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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.
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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
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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
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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:
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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.
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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
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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
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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
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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
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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
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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
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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.
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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.
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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
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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
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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.
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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
11.31
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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).
11.32
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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?
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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
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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.
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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
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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.
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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
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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
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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.
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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.
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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
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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
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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
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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.
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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
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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.
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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
13.12
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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:
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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.
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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
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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.
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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,
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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."
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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.
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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
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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
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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:
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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
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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
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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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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
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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.
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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
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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
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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
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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.
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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.
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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.
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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.
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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
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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.
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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
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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
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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
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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,
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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.
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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
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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.
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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.).
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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
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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.
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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).
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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
15.15
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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
15.17
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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,
15.23
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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?
15.26
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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?
15.27
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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?
15.28
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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
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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
-------
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
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tel lur ium
Atomic
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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
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lanthanum
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prometh ium
sama- ium
europium
gaool in ium
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dysprosium
holmium
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th u 1 i urn
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lutetium
hafnium
tantalum
tungsten
rhenium
osmium
ir idium
platinum
gold
mercury
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lead
bismuth
polonium
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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
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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
-------
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.
-------
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
-------
DARCOM-P 385-1
BIBLIOGRAPHY
ALARA
International Commission on Radiological Protection. 1973. Implications of
Commission Recommendations that Doses be Kept as Low as Readily Achievable.
ICRP 22, Pergamon Press, New York.
Kathren, R. L., et al. 1980. A Guide to Reducing Radiation Exposures to As
Low As Reasonably Achievable (ALARA).DOE/EV/1830-T5, National Technical
Information Service, Springfield, Virgi n i a.
Mauro, J. J., and J. Porrovecchio. 1976. "Numerical Criteria for In-Plant 'As
Low As Is Reasonably Achievable,'" 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.
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Johns, H. E., and J. R. Cunningham. 1974. The Phys-ics of Radiology. 3rd ed.
Charles C. Thomas, Springfield, Illinois.
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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.
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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,
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Budnitz, R. J. 1974. "Radon-222 and Its Daughters - A Review of Instrumenta-
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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
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Eichholz, G. G., and J. W. Poston. 1979. Principles of Nuclear Radiation
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European Nuclear Energy Agency. 1963. Proceedings of a Symposium on
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Fitzgerald, J. J. 1969. "Instruments for Radiation Detection and Measure-
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Gessell, T. F., G. de P. Burke and K. Becker. 1976. "An International
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Gibbs, W. D., and C. C. Lushbaugh. 1969. "Whole-Body Counter System," in
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Hankins, D. E. 1973. "Progress in Personnel Neutron Dosimetry," in
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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
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Hoy, J. E. 1972. "An Albedo-Type Personnel Neutron Dosimeter." Health Phys.
24:385.
International Atomic Energy Agency. 1970. Nuclear Accident Dosimetry
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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
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Kocher, L. F., L. L. Nichols, G. W. R. Endres, D. V. Shipler and
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Knoll, G. F. 1979. Radiation Detection and Measurements. John Wiley and
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National Council on Radiation Protection and Measurements. 1978.
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Oshino, M. 1973. "Response of NTA Personnel Neutron Monitoring Film Worn on
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Price, W. J. 1964. Nuclear Radiation Detection. 2nd ed. McGraw-Hill Inc.,
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Rich, B. L., end B. G. Samardzich. 1973. "Organizational Aspects of a Per-
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Sommers, J. F. 1975. "Sensitivity of the G-M and Ion Chamber Beta-Gamma
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Technical Education Research Center-S.W. 1980. Course II, Radiation Detec-
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U.S. Department of Defense, Dense Civil Preparedness Agency. 1977. Handbook
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U.S. Nuclear Regulatory Commission (NRC). 1973. "Film Badge Performance
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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
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A-17
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American National Standards Institute and American Society of Mechanical
Engineers. 1979. Quality Assurance Prograrr. Requirements for Nuclear
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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.
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U.S. Nuclear Regulatory Commission. 1974. "Quality Assurance Terms and
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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
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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.
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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.
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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 -
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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 -
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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
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Energies Up to 10 MeV^NCRP 49, Washington, D.C.
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Price, B. T., C. C. Horton and K. T. Spinney. 1957. Radiation Shielding.
Pergamon Press, New York.
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Nostrand, Princeton.
Schaeffer, N. M., ed. 1973. Reactor Shielding for Nuclear Engineering,
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Trout, E. D., 0. P. Kelley and G. L. Herbert. 1975. "X-Ray Attenuation in
Steel-50 to 300 kVp." Health Phys. 29:163.
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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
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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
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Feldman, A. 1976. "Factors and Strategies in Selection of Instruments for
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Fish, B. R., ed. 1964. Surface Contamination. Pergamon Press, New York.
Gaeta, N. A., and M. H. Repacholi. 1975. "A Standard Survey Procedure for
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International Atomic Energy Agency. 1979. Radiological Surveillance of
Airborne Contaminants in the Working Environment/ IAEA Safety Series
No. 49, Vienna.
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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.
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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.
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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.
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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
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Gera, F. 1974. "The Classification of Radioactive Waste." Health Phys.
27:113.
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Waste: A Comparative Cost Analysis." Health Phys. 29:389.
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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
-------
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
-------
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
-------
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
-------
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.
-------
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
-------
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
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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,
-------
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
-------
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
-------
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
-------
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
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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.
-------
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.
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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.
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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
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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
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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
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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
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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.
-------
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
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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
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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
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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
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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
-------
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
-------
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.
5/93 11
-------
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.
5/93 12
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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:
5/93 13
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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.
5/93 26
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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.
5/93 27
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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
5/93 31
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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
5/93 33
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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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