40 CFR Part 61
National Emission Standards
for Hazardous Air Pollutants
EPA 520/1-89-001
Background Information Document:
Procedures Approved for
Demonstrating Compliance
with 40 CFR Part 61, Subpart I
Office of Radiation Programs
U.S. Environmental Protection Agency
Washington, DC
October 1989
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PREFACE
The purpose of this document is to provide information supporting
the procedures developed for NRC-licensed and non-DOE Federal
facilities to use in demonstrating compliance with the radio-
nuclides NESHAP (40 CFR 61, Subpart I). Specifically, this Back-
ground Information Document has the following objectives:
provide the basis for the calculational and
analytical methods approved for determining
emissions; and
provide the basis for the procedures approved for
demonstrating compliance with the dose limits of
-the standard.
This document comprises four chapters. Chapter 1 provides back-
ground information on the history of the rulemaking and summa-
rizes the major provisions of the NESHAP. Chapter 2 describes the
types of facilities covered by the NESHAP. Since the potential
for releasing radioactive materials to the atmosphere depends on
the specific radionuclides used, the processing and handling that
they undergo, and the effluent controls which are used, the in-
formation presented focuses on these factors. Chapter 3 presents
the basis for the methods approved to determine emissions. Par-
ticular emphasis is given to the derivation and use of emission
factors* to estimate the quantities of material handled that
*As used in this document, an emission factor is defined as the
pre-effluent control fraction of a radioactive material that be-
comes airborne. Thus, the quantity of material that is released
to the atmosphere is the product of the quantity that becomes
airborne and the appropriate effluent control adjustment factor.
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become airborne and to the adjustment factors for effluent con-
trols, which may be applied to these quantities to estimate emis-
sions. Within the restrictions given, these emission factors and
effluent control adjustment factors may be used in lieu of meas-
ured release rates in determining compliance with the standard.
Chapter 4 presents the basis for the procedures approved for
determining compliance. References are presented at the end of
each chapter. Additional references are listed in Appendix A.
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TABLE OF CONTENTS
1. INTRODUCTION
1.1 BACKGROUND
1.2 SUMMARY OF THE NESHAP
1.2.1 Applicability
1.2.2 The Standard
1.2.3 Demonstrating Compliance
1-1
1-1
1-2
1-2
1-3
1-3
REFERENCES
1-5
2. FACILITY DESCRIPTIONS
2.1 INTRODUCTION
2.2 NRC MATERIAL LICENSEES
2.2.1 Users and Producers of Radionuclides
for Medical Purposes
2.2.1.1 Radiopharmaceutical Users
2.2.1.2 Radiopharmaceutical Producers
and Suppliers
Sealed Source Manufacturers
2.2.2.1 Manufacturers of Sealed Radiation
Sources
2.2.2.2 Manufacture of Self-Illuminating
Devices
Test and Research Reactors
Non-Light-Water Reactor Fuel Fabricators
Source Material Licensees
Waste Receivers/Shippers and Disposal
Facilities
2.2.2
2.2.3
2.2.4
2.2.5
2.2.6
2-1
2-1
2-1
2-1
2-2
2-5
2-7
2-7
2-8
2-9
2-10
2-11
2-12
2.3 URANIUM FUEL CYCLE FACILITIES
2.3.1 Uranium Mills
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2-13
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2.3.2 Uranium Conversion Facilities
2.3.2.1 Dry Hydrofluor Process
2.3.2.2 Solvent Extraction Process
2.3.3 Fuel Fabrication Facilities
2.3.4 Nuclear Power Facilities
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2-15
2-15
2-16
2-17
2.4 DEPARTMENT OF DEFENSE FACILITIES
2-18
REFERENCES
2-20
3. DETERMINATION OF EMISSIONS
3.1 INTRODUCTION
3-1
3-1
3.2 EMISSION FACTORS
3.2.1 Derivation of Emission Factors
3.2.1.1 Procedures for Obtaining Data
3.2.1.2 Analysis of Data
3.2.2 Approved Emission Factors
3.2.2.1 Gases
3.2.2.2 Liquids and Powders
3.2.2.3 Solids
3-2
3-3
3-3
3-6
3-22
3-22
3-24
3-27
3.3 ADJUSTMENT FACTORS FOR EFFLUENT AIR CONTROL DEVICES 3-28
3.3.1 Use of Effluent Air Control Devices
3.3.1.1 HEPA Filters
3.3.1.2 Baghouse Filters
3.3.1.3 Sintered-Metal Filters
3.3.1.4 Activated Carbon Filters
3.3.1.5 Douglas Bags
3.3.1.6 The Xenon Trap
3.3.1.7 Venturi Scrubbers
3.3.1.8 Packed-Bed Scrubbers
3.3.1.9 Electrostatic Precipitators
3-28
3-31
3-32
3-33
3-33
3-34
3-35
3-35
3-36
3-37
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3.4 APPROVED SAMPLING AND ANALYTICAL METHODS
3-37
REFERENCES
3-39
4. COMPLIANCE PROCEDURES AND EXEMPTION CRITERIA
4.1 INTRODUCTION -
4-1
4-1
4.2 MODELS USED IN THE COMPLIANCE PROCEDURES
4.2.1 Atmospheric Dispersion Models
4.2.2 Models used to Estimate Exposures
4.2.3 Derivation of Dose Conversion Factors
4-3
4-3
4-9
4-14
4.3 APPROVED. COMPLIANCE PROCEDURES
4.3.1 Procedure 1: Quantity of Material Handled
Concentration Limits
NCRP Screening Procedures
Compliance Model of the
COMPLY Computer Code
4.3.2
4.3.3
4.3.4
Procedure 2:
Procedure 3:
Procedure 4:
4.4 EXEMPTION CRITERIA
REFERENCES
APPENDIX A: ADDITIONAL REFERENCES
APPENDIX B: DERIVATION OF EMISSION FACTORS FROM IFTAH DATA
4-15
4-15
4-28
4-35
4-36
4-37
4-39
A-l
B-l
REFERENCES
B-8
VI1
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LIST OF TABLES
3-1. Summary of Reported and Derived Emission Factors
3-2. EPA-Approved Emission Factors
3-3. Approved Adjustment Factors for Effluent Controls
4-1. Annual Possession Quantities for Environmental
Compliance
4-2. Concentration Levels for Environmental Compliance
3-20
3-23
3-29
4-17
4-29
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CHAPTER 1. INTRODUCTION
1.1 BACKGROUND
On December 27, 1979, the Administrator of the Environmental Pro-
tection Agency (EPA) listed radionuclides as hazardous air pol-
lutants subject to regulation under Section 112 of the Clean Air
Act (FR79). Three National Emission Standards for Hazardous Air
Pollutants (NESHAPS) were promulgated on February 6, '1985, regu-
lating radionuclide emissions from Department of Energy (DOE)
facilities, Nuclear Regulatory Commission- (NRC-) licensed and
non-DOE Federal facilities, and elemental phosphorus plants
(FR85a). Two additional radionuclide NESHAPS, covering radon-222
emissions from underground uranium mines and licensed uranium
mill tailings, were promulgated on April 17, 1985, and September
24, 1986, respectively (FR85b, FR86).
The EPA's basis for the radionuclide NESHAPS (including its deci-
sion not.to propose NESHAPS for certain categories of facilities)
was challenged in lawsuits filed by the Sierra Club and the Nat-
ural Resources Defense Council (NRDC). While these suits were
under adjudication, the U.S. Court of Appeals for the District of
Columbia issued a decision finding that the EPA's NESHAP for
vinyl chloride was defective in that costs had been improperly
considered in setting the standard. Following the Court's order
to review the potential effect of the vinyl chloride decision on
other standards, the EPA determined that costs had been consid-
ered in many rulemakings on radionuclide emissions. On December
9, 1987, the Court accepted the EPA's proposal to leave the
existing radionuclide NESHAPS in place while the Agency reconsi-
der the standards. In the interim, the suits filed by the Sierra
Club and the NRDC have been placed in abeyance. As a result of
the reconsideration, the EPA has promulgated a NESHAP for NRC-
licensed and non-DOE Federal Facilities.
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This Background Information Document deals only with the compli-
ance procedures for implementing the radionuclide NESHAP
covering NRC-licensed and non-DOE Federal facilities (40 CFR Part
61, Subpart I). A separate Environmental Impact Statement pre-
sents the risks posed by these facilities and the Agency's ration-
ale for the dose standard (EPA89).
This Background Information Document explains the basis for the
procedures for demonstrating compliance with the dose limits.
The compliance procedures include methods for determining the
quantities of radionuclides released to the atmosphere and meth-
ods for estimating the resulting doses to the most exposed indi-
vidual.
1.2 SUMMARY OF THE NESHAP
1.2.1 Applicabili ty
The NESHAP applies to facilities licensed by the NRC or an Agree-
ment State and all Federal facilities, except those owned or
operated by the DOE, that use or possess unsealed radiation sour-
ces. The NESHAP does not apply to facilities regulated under 40
CFR Part 191 Subpart B, to low-energy accelerators, or to facili-
ties that use or possess only sealed radiation sources. ,
The facilities covered by the NESHAP and the activities at these
facilities involving the use of radioactive materials are quite
diverse. They include both NRC material licensees and facilities
engaged in the uranium fuel cycle. NRC material licensees .in-
clude radiopharmaceutical suppliers and medical users, govern-
ment, academic and private research facilities, test and research
reactors, low-level waste shippers and disposal sites, and other
suppliers and users of radionuclides. In total, the NESHAP
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applies to an estimated 6,000 government, academic, medical, and
industrial facilities.
1.2.2 The Standard
The NESHAP limits annual radionuclide emissions to the atmosphere
from these facilities to such quantities that will not result in
any member of the public receiving an effective dose equivalent
in excess of 10 millirem per year (mrem/yr). Further, not more
than 3 mrem/yr effective dose equivalent may be caused by iso-
topes of iodine. The standard specifically excludes doses caused
by radon-220 or radon-222 and their decay products that are
formed after release.
Facilities covered by the NESHAP are also subject to the report-
ing and approval requirements of 40 CFR Part 61, Subpart I, Sec-
tions 61.104(a) and 61.106(a) of Part 61. However, Sections
61.104(b) and 61.106(b) of Subpart I exempt from these require-
ments any facility that, using the specified procedures, demon-
strates that its total emissions do not cause any member of the
public to receive a dose greater than 10 percent of the limits of
the standard. Further, the approval requirements are waived if,
again using the specified procedures, the emissions from a newly
constructed or modified facility will not cause any member of the
public to receive a dose in excess of 1 percent of the standard.
1.2.3 Demonstrating Compliance
The standard limits doses to the most exposed member of the
public. Dose is a complicated function of the quantity of each
radionuclide emitted; the physical configuration of the facility
releasing the material; the dispersion, transport, and build-up
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of the radionuclides in the soil and foodstuffs; and the proxim-
ity to the facility of individuals and farms producing foodstuffs,
While, in principle, the doses resulting from the release of
radionuclides to the atmosphere can be determined by environ-
mental monitoring, valid measurements are often impossible to ob-
tain. At the levels consistent with the limit of the standard,
concentrations of many radionuclides will be below the minimum
detection level of even state-of-the-art measurement technology.
Even when the concentrations can be adequately measured, it, may
not be possible to distinguish the portion attributable to the
emissions from that which is due to background radioactivity.
Therefore, compliance with the limit of the standard is to be
demonstrated using EPA-approved methods of determining emissions
and EPA-approved procedures for estimating the resulting doses.
The NESHAP includes approved calculational and analytical methods
for determining emissions, and a number of alternative procedures
for determining doses. Other methods and procedures (including
those based on environmental measurements) must be submitted to
the EPA for approval before they are used.
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REFERENCES
EPA89 U.S. Environmental Protection Agency, Risk Assessments;
Environmental Impact Statement for NESHAPS - Radionuclides
Volume 2, Office of Radiation Programs, Washington, DC,
September 1989.
FR79 44 Federal Register. 76738, December 27, 1979.
FR85a 50 Federal Register. 5190-5200, February 6, 1985.
FR85b 50 Federal Register. 15386-15394, April 17, 1985,
FR86 51 Federal Register. 34056-34067, September 24, 1986.
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CHAPTER 2. FACILITY DESCRIPTIONS
2.1 INTRODUCTION
The NESHAP applies to approximately 6,000 NRC-licensed and non-
DOE Federal facilities that possess unsealed -sources of radio-
active materials. The NRC-licensed facilities include material
licensees and facilities engaged in the uranium fuel cycle. NRC-
licensed facilities include facilities licensed by the Agreement
States but exclude low-energy accelerators and facilities
regulated under 40 CFR Part 191, Subpart B.
The major types of facilities covered by the standard are de-
scribed in the following sections. The discussion focuses on the
physical forms of the radionuclides used and the handling and
processing that the materials undergo. These factors are major
determinants of the quantities of materials handled that become
airborne.
2.2 NRG MATERIAL LICENSEES
2.2.1
Users and Producers of Radionuclides for Medical Purpos
The users and producers of radioactive materials for medical pur-
poses constitute by far the largest category of facilities han-
dling unsealed radioactive sources. Approximately two-thirds of
the 6,000 facilities covered by the NESHAP are engaged in some
aspect of the production and distribution of radiopharmaceuticals
or in the medical application of these materials. Medical uses
of radiopharmaceuticals include biomedical research and patient
administration of radiopharmaceuticals for both diagnostic and
therapeutic purposes.
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2.2.1.1 Radiopharmaceutical Users
The types of facilities that use radionuclid.es for medical pur-
poses include hospitals, clinics, and biomedical research facili-
ties. The radionuclides used directly in patient -therapy and
diagnosis are termed "radiopharmaceuticals," while those used in
research are referred to as "radionuclides." For simplicity, the
term "radiopharmaceuticals" will be used to refer to the radio-,
active materials used in both patient administration and research.
The radiopharmaceuticals used at medical facilities occur in all
three basic physical states: solid, liquid, and gas. The phys-
ical state of a particular radiopharmaceutical product is deter-
mined by (1) the chemical form of the radionuclide and (2) the
solution or other mixture, if any, in which the radionuclide is
dispensed. Both the radionuclide and the substance in which it
is mixed are chosen to suit specific therapeutic, diagnostic, and
research purposes.
The mixing of the radionuclide with some other substance means
that the physical state of a radiopharmaceutical product may be
different than the physical state of the radionuclide itself. In
this document, discussions of the form of a particular radionu-
clide refer to the radionuclide product. The physical states of
these products are important in assessing the potential for air-
borne release.
Most radionuclides used in medical facilities occur in liquid
form. These liquids may be administered either orally or intra-
venously. Orally administered radionuclides are usually in the
form of aqueous solutions. Many of these chemicals are ionic
salts and thus occur in liquid form as saline solutions. Radio-
nuclides that are administered intravenously may occur as solu-
tions, colloids, or suspensions.
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Solutions consist of molecules of solids or gaseous substances
dissolved in a liquid. Colloids involve the dispersion of larger
particles (on the order of 10 nanometers to 1 micrometer in diam-
eter) in a liquid medium; the larger particles are prevented from
aggregating and settling by being coated with a layer of gelatin
(as is done with gold-198). Suspensions are similar to colloids
but involve the radionuclide labeling of still larger particles
(greater than 10 micrometers in diameter) of substances such as
human serum albumin.
Gaseous radionuclides usually occur naturally in elemental form
(e.g., xenon-133), and are administered to patients as a pure gas
or as a gas diluted by air. Patients normally inhale the gas
from a bag or from a gas "generator" through a respirator.
Solid radionuclides occur as gelatin capsules containing liquid
solutions of the radionuclide chemical. In some cases, the solu-
tion is absorbed in dry filler material. Solid radionuclides are
administered orally to patients.
The number of radionuclides with medical applications is exten-
sive and increasing. In the areas of diagnosis and therapy, the
most commonly used radiopharmaceuticals include chromium-51;
cobalt-57, -58, and -60; gallium-67 and -68; technetium-99m;
iodine-123, -125, and -131; selenium-75, xenon-127 and -133; and
thallium-201. Biomedical researchers employ tritium, carbon-14,
phosphorus-32, and sulfur-35 extensively. The radiopharma-
ceuticals used in medical applications may be obtained from
radiopharmaceutical manufacturers or independent radiopharmacies,
or they may be produced on site from radiopharmaceutical
generators. Because of the relatively short half-lives of the
radionuclides used in medicine, shipments from vendors are
received frequently (weekly or daily), and storage times are
minimal.
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Radiopharmaceuticals purchased from vendors may be in the form of
pre-packaged dose kits, radiopharmaceutical generators, or bulk
supplies from which individual doses are extracted and prepared.
Handling of prepackaged dose kits may involve no more than remov-
ing the material from the package and administering the radio-
pharmaceutical to the patient either orally or by intravenous
injection.
Handling of materials obtained in the form of bulk stocks or
radiopharmaceutical generators is more involved. In general,
these materials are received and stored in a central area where
individual doses are prepared. In the case of liquids, dose pre-
paration involves extracting the required quantity from the stock
solution by syringe or pipette and diluting the material in a
suitable sterile medium. These operations are conducted in a
fume hood, and the dose is administered to the patient either
intravenously or orally.
Preparation of doses from radiopharmaceutical generators, of
which molybdenum-99/technetium-99m generators are the most com-
mon, involves elution of the product from the generator and divi-
sion of the elute into individual doses. The procedures for
eluting a. generator depend on whether it is a wet or dry column
design. In a wet column generator, an evacuated extraction vial
is attached to the end of the generator column with a sterile
needle. Using the vacuum within the vial, the solvent is pulled
from the generator reservoir through the column and into the
vial. The procedure for a dry column generator is similar. How-
ever, since dry generators do not have a reservoir of solvent,
solvent must be added to the column prior to elution. The charge
vial is attached to one end of the generator, and then the evacu-
ated extraction vial is attached to the other end. The 'solution
is drawn through the generator column and collected in the elu-
tion vial. These elution procedures and dose divisions are
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conducted in a fume hood, with the generator shielded to prevent
external irradiation of the technicians.
Handling of radionuclides for biomedical research is more varied
than that of radiopharmaceuticals used for patient administra-
tion. Depending on the specific radionuclides used and the goal
of the experiment, the materials may simply be extracted from
bulk stocks and administered, or the radionuclide may be subject-
ed to additional chemical or physical processing.
2.2.1.2 Radiopharmaceutical Producers and Suppliers
Radiopharmaceutical manufacturers produce the radionuclide-
labeled compounds, diagnostic kits, and radionuclide generators
used in biomedical research and medical diagnosis and therapy.
The radiopharmaceutical products may be shipped directly to medi-
cal users, or they may be shipped to independent radiopharmacies
where individual doses are prepared from the bulk supplies or
generators and distributed to medical users. Individual radio-
pharmaceutical manufacturers may specialize in only a few widely
used radiopharmaceuticals or may produce many of the radionuc-
lides used in biomedical research and patient diagnosis and
therapy.
The radionuclides used in radiopharmaceuticals are produced
either in nuclear reactors or in accelerators. Radiopharmaceuti-
cal manufacturers may operate their own production facilities or
may purchase the bulk radionuclides from an outside vendor. In
producing the bulk radionuclides, a suitable target is first prepar-
ed and then bombarded with neutrons or positive ions in the re-
actor core or accelerator. Once irradiation is complete, the
target is removed from the production device, and the product is
recovered and purified in a hot cell by appropriate chemical
processing.
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The production of the labeled compounds used in radiopharmaceuti-
cals and biomedical research is essentially a wet chemistry proc-
ess. Depending on the specific radiopharmaceutical, workers con-
duct these operations within laboratory fume hoods or gloveboxes.
The final products are generally assembled and packaged in as-
sembly line operations.
Radiopharmaceutical generators are designed and produced as
closed aseptic systems using some type of chromatographic column.
Typically, this chromatographic column consists of an inorganic
ion exchange resin to which the generator (parent) radionuclide
is bound. As the parent radionuclide decays, the decay product,
which has different chemical/physical properties, is produced.
The decay product is eluted from the column by the user at
specified intervals. Generators are manufactured in a hot cell,
where the parent radionuclide is packed in the column, and the
column of the generator is surrounded by absorbent materials and
shielding. The absorbent materials minimize the consequences of
accidental breakage; the shielding reduces the radiation exposure
of users. Once the generator is loaded, final assembly and
packaging are carried out on an assembly line.
Independent radiopharmacies are a relatively recent phenomenon.
Generally located in large cities, these facilities serve as
distribution facilities. Radiopharmacies purchase bulk stocks
and generators from radiopharmaceutical manufacturers and provide
hospitals and clinics with individually prepared doses on an as-
needed basis. The dose preparation procedures at these facili-
ties do not differ from those at medical facilities that obtain
their radiopharmaceuticals directly from the manufacturers.
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2-2.2 Sealed Source Manufacturers
While facilities that use only sealed radiation sources are not
covered by the NESHAP, the industrial facilities that produce
sealed sources are subject to the standard. The facilities de-
scribed in this section fall into two broad classes: those that
manufacture encapsulated alpha, beta, or gamma-emitting radiation
sources; and those that manufacture self-luminous devices.
2.2.2.1 Manufacturers of Sealed Radiation Sources
Sealed radiation sources are widely used in medical, industrial,
and residential applications. Medical applications include
gamma-emitting devices used in diagnostic arid therapeutic proce-
dures and sources used in patient implants. Industrial applica-
tions include nondestructive imaging and inspections, static
eliminators, industrial gages, irradiation devices, and well-
logging devices. The main radionuclides used in these devices
iridium-142, krypton-85, americium-241, cesium-137, and cobalt-60
Smoke detectors, using alpha-emitting americium-241 sources, are
the most widely used sealed sources in residential applications.
The manufacture of sealed sources is essentially a repackaging
and redistribution process. Bulk radionuclides, in the form of
pellets or foils, are received from a vendor in an approved ship-,
ping package. The shipping package is opened, and the required
quantity of the radioactive material is removed and transferred
to a container. The container is then sealed by welding or braz-
ing. Most such devices are double encapsulated; i.e., an inner
capsule contains the radioactive material and an outer container
protects the inner container. Double encapsulation increases .the
assurance of safe handling. 'The outer container may also be
brazed or welded, or simply screwed shut. All operations are
performed in hot cells to protect the workers.
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At some facilities, the bulk material purchased from the vendor
is subjected to physical and/or chemical processing to alter the
form of the material prior to encapsulation. For example, most
cobalt-60 sources contain cobalt in the form of metal foils or
microspheres. The cobalt is received from the vendor in the form
of cobalt metal, and the material is processed by heating the
metal to the melting point in a fluidizing furnace to form the
desired microspheres. Similarly, manufacturers of smoke
detectors generally obtain the bulk americium-241 in the form of
oxide powder. This powder is compacted to form wafers, sintered
in an induction furnace, ground to specifications, and hot-rolled
with gold foil to produce the encapsulated material for incorpor-
ation into the device.
2.2.2.2 Manufacture of Self-Illuminating Devices
Self-illuminating devices include watches, compasses, signs, and
aircraft instrumentation. Historically, radium-226 was used in
radio-luminescent products. However, the well-documented hazards
of working with radium and the advent of other materials with in-
herently superior characteristics have largely eliminated the use
of radium. Today, tritium and, to a much lesser extent, krypton-
85 and promethium-147 are used in the production of self-luminous
devices.
Two general types of self-illuminating devices are made: those
in which the radio-luminous material is incorporated into a paint
which is used to coat the dial and/or instrument hands; and those
in which a radioactive gas (tritium or krypton) is contained in a
phosphor-coated glass ampule.
Manufacturers of self-illuminating devices obtain the bulk radio-
nuclides in either gaseous or (rarely) liquid form from a vendor.
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In the case of devices incorporating self-luminous paint, the
manufacturing process involves the incorporation of the
radionuclide in the paint and the application of the paint to the
device. In the case of self-illuminating sources, the gaseous
radionuclide (tritium or krypton-85) is transferred to the glass
ampule and sealed. Both processes are carried out in areas with
high ventilation rates or in fume hoods to protect the workers.
2.2.3 Test and Research Reactors
The NRG licenses approximately 70 academic, research, and indus-
trial facilities to operate test and research reactors. Test and
research reactors are used as teaching devices, to study reactor
designs, to conduct research on the effects of radiation on ma-
terials, and to produce radioactive materials used by sealed
source and radiopharmaceutical manufacturers.
The design of such reactors and their sizes vary widely. Approx-
imately 15 research reactors are used primarily as teaching
devices and have very low power outputs (less than 15 watts).
The nuclear cores of these reactors have their uranium fuel dis-
persed and fixed in a plastic matrix. Given the design and use
of these teaching reactors, airborne releases cannot occur during
normal operations.
Research and test reactors used for experimental and production
purposes include both light-water pool and heavy-water tank-type
designs, ranging in power from 100 kilowatts to 10 megawatts.
All of these facilities use highly enriched uranium fuel, either
in metal or mixed carbide fuel elements.
In these reactors, experiments and/or production activities are
conducted by remotely inserting the target containing the
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material to be irradiated into the experimental ports or beam
holes that penetrate the reactor core. The target material is
subjected to the neutron flux of the reactor core for an appro-
priate period of time and then withdrawn via shielded transport
devices (called "rabbit systems") to a hot cell. The irradiated
material is examined or the product is recovered in the hot cell.
Product recovery may be as simple as dissolving a soluble salt in
water, or it may involve evaporation, precipitation, extraction,
distillation, and/or ion exchange.
Potential airborne releases from such facilities include the fis-
sion products in the core of the reactor, activation products
generated during the operation of the reactor, and releases from
the disassembly and recovery of target materials in the hot cell.
In general, the activation products, along with any gaseous fis-
sion products escaping the coolant, are released directly to the
atmosphere from the facility exhaust. Materials that become air-
borne during processing in the hot cell will be vented through
the hot cell's exhaust system. The effluent from the hot cell is
generally filtered through high efficiency particulate air (HEPA)
filters before release.
2.2.4 Non-Light-Water Reactor Fuel Fabricators
Only a few facilities produce the metal and mixed carbide fuel
used in test and research reactors.
The non-oxide fuel fabrication process begins with highly en-
riched uranium metal. The uranium metal may be mixed with an
alloying metal in an induction furnace. The fuel is then either
rolled, punched, drilled, or crushed and compacted, and machined
and shaped into the proper dimensions. Once the fuel is properly
formed, it is enclosed in aluminum or stainless steel. The
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enclosing process may involve injection casting, loading into a
can or mold, or simply covering the fuel with side plates and
rolling the metals together. Finished fuel elements are then in-
spected and cleaned prior to assembly into fuel bundles.
The production of mixed carbide fuel starts with highly enriched
uranium dioxide-thorium dioxide powder. (UO2~ThO2-)-. This powder
is mixed with graphite and heated to form uranium-thorium carbide
kernels. These kernels are formed into microspheres by heating
to a temperature in excess of the kernels' melting point. The
microspheres are then coated with carbon and silicon layers in a
fluidized bed furnace. Fuel rods are formed by injecting the
coated kernels and a matrix material into a hot mold. The fin-
ished rods are then inserted into a graphite block to form the
final fuel assembly.
2.2.5 Source Material Licensees ,
Two types of facilities are included in the category of "Source
Material Licensees" which is subject to the NESHAP: those invol-
ved in the extraction of metals from uranium- and thorium-bearing
ores, and those using depleted uranium metal or thorium in
various products.
Approximately 10 facilities are engaged in the recovery of metals
from source materials. In general, the products extracted from
the uranium- and thorium-bearing ores are refractory metals,
their oxides (columbium/niobium, zirconium, tantalum, and hafni-
um), or the rare earths (cerium, neodymium, dysprosium, etc.).
These extraction operations involve processes typical of metal
mining and beneficiation. Depending upon the specific facility
and the products under recovery, the processing may involve wet
chemical or solvent extraction, smelting, and high temperature
sintering.
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Facilities that manufacture products incorporating source mater-
ials include munitions producers using depleted uranium in armor-
piercing projectiles, manufacturers that make lanterns and gas
lights using thorium mantles, aerospace manufacturers using de-
pleted uranium for stabilizers and ballast, and welding rod manu-
facturers that use thorium in the metallic form. Such manufac-
turers generally receive the material in the physical form in
which it is used (e.g., depleted uranium in the form of metal
billets). The processing is confined to such metallurgical oper-
ations as casting, forging, machining, and polishing.
2.2.6 Waste Receivers/Shippers and Disposal Facilities
The radioactive wastes generated by facilities that use radio-
nuclides must be disposed of in an approved manner. In general,
wastes with high specific activities (such as uranium-contaminat-
ed scrap at non-oxide fuel fabrication facilities) will be re-
cycled and recovered. However, virtually every user of unsealed
radioactive materials will generate solid, low-level radioactive
wastes which require active disposal. Such wastes may be
incinerated on site or packaged and shipped off site to a
licensed low-level waste disposal facility.
Waste receivers and shippers (sometimes called "waste brokers")
are primarily collection and shipping agents for facilities
generating low-level wastes. Most such receiving/shipping
facilities simply collect the wastes in shipping containers
approved by the Department of Transportation from a number of
waste generating facilities, monitor the packages for con-
tamination, and hold the wastes at a warehouse until they arrange
a shipment to a licensed disposal site. The licenses of most
such receiving and shipping facilities do not allow the facility
to repack or even open the waste packages. However, several such
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facilities have been licensed to open, compact, and repackage
waste materials before shipment.
Currently, there are three low-level radioactive waste disposal
facilities which are accepting shipments for burial: the
Barnwell facility in South Carolina, the Beatty facility in
Nevada, and the Richland facility in Washington. Waste shipments
are checked for damage and contamination upon receipt and then
placed in excavated trenches. When a burial trench is filled
with waste it is backfilled with soil.
2.3 URANIUM FUEL CYCLE FACILITIES
The uranium fuel cycle includes uranium mills, uranium hexa-
fluoride conversion facilities, uranium enrichment facilities,
light-water reactor fuel fabricators, light-water power reactors,
and fuel reprocessing plants. With the exception of the uranium
enrichment facilities that are owned by the Federal government
and operated by contractors under the supervision of the Depart-
ment of Energy (DOE), these facilities are licensed by the
Nuclear Regulatory Commission (NRC) or the Agreement States.
2.3.1 Uranium Mills
Uranium mills extract uranium from ores which contain only 0.01
to 0.3 percent U3O8. Uranium mills, typically located near
uranium mines in the western United States, are usually in areas
of low population density. The product of the mills is shipped
to conversion plants, where it is converted to volatile uranium
hexafluoride (UFg) which is used as feed to uranium enrichment
plants.
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As of December 1988, of 27 uranium mills in the United States
licensed by the NRG or agreement states, four were operating,
eight were shut down, 14 were being decommissioned, and one had
been built but never operated. The eight shut down mills could
resume operations, but the 14 mills that are being decommissioned
will never operate again.
The operating mills have a capacity of 9,600 tons of ore per day.
The number of operating mills is down considerably from 1981,
when 21 mills were processing approximately 50,000 tons of ore
per day. This reduction reflects the decrease in the demand for
yellowcake. The mined ore is stored on pads prior to processing.
Crushing and grinding and a chemical leaching process separate
the uranium from the ore. The uranium product is dried and pack-
aged following recovery from the leach solution. The waste
product (mill tailings) is piped as a slurry to a surface impound-
ment area (tailings pile).
Radioactive materials released to the air during these operations
include natural uranium and thorium and their respective decay
products (e.g., radium, lead, radon). These radionuclides, with
the exception of radon, are 'released as particulates.
2.3.2 Uranium Conversion Facilities
The uranium conversion facility purifies and converts uranium
oxide (11303 or yellowcake) to volatile uranium hexafluoride
(UFg),the chemical form in which uranium enters the enrichment
plant.
There are currently two commercial uranium hexafluoride (UFg)
production facilities operating in the United States, the Allied
Chemical Corporation facility at Metropolis, Illinois and the
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Kerr-McGee Nuclear Corporation facility at Sequoyah, Oklahoma.
The Allied Corporation facility, a dry-process plant in operation
since 1968, has a capacity to produce about 12,600 mt of uranium
per year in the form of UF6.- The Kerr-McGee Nuclear Corporation
facility is a wet-process plant, in operation since 1970, with a
capacity of about 9,100 mt per year (AEC74, Do88).
Two industrial processes are used for uranium hexafluoride pro-
duction, the dry hydrofluor method and the wet solvent extrac-
tion method. Each method produces roughly equal quantities of
uranium hexafluoride; however, the radioactive effluents from the
two processes differ substantially. The hydrofluor method re-
leases radioactivity primarily in the gaseous and solid states,
while the solvent extraction method releases most of its radio-
active wastes dissolved in liquid effluents.
2.3.2.1 Dry Hydrofluor Process
The hydrofluor process consists of reduction, hydrofluorination,
and fluorination of the ore concentrates to produce crude uranium
hexafluoride. Fractional distillation is then used to obtain
purified UF6. Impurities are separated either as volatile com-
pounds or as a relatively concentrated and insoluble solid waste
that is dried and drummed for disposal.
2.3.2.2 Solvent Extraction Process
The solvent extraction process employs a wet chemical solvent
extraction step at the start of the process to prepare high
purity uranium for the subsequent reduction, hydrofluorination,
and fluorination steps. The wet solvent extraction method separ-
ates impurities by extracting the uranium from the organic
ซ
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solvent, leaving the impurities dissolved in a aqueous solution.
The raffinate is impounded in ponds at the plant site.
2.3.3 Fuel Fabrication Facilities
Light water reactor (LWR) fuels are fabricated from uranium which
has been enriched in U-235. At a gaseous diffusion plant natural
uranium in the form of UF6 is processed to increase the U-235
content from 0.7% up to 2% to 4% by weight. The enriched uranium
hexafluoride product is shipped to LWR fuel fabrication plants
where it is converted to solid uranium dioxide pellets and
inserted into zirconium alloy (Zircaloy) tubes. The tubes are
fabricated into fuel assemblies which are shipped to nuclear
power plants. There are seven licensed uranium fuel fabrication
facilities in the United States which fabricate commercial LWR
fuel. Of the seven, only five had active operating licenses as
of January 1, 1988. Of those five facilities, two use enriched
uranium hexafluoride to produce completed fuel assemblies and two
use uranium dioxide. The remaining facility converts UF6 to UO2
and recovers uranium from scrap materials generated in the var-
ious processes of the plant.
The processing technology used for uranium fuel fabrications con-
sists of three basic operations: (1) chemical conversion of UFg
to U02; (2) mechanical processing including pellet production
and fuel-element fabrication; and (3) recovery of uranium from
scrap and off-specification material. The most significant po-
tential environmental impacts result from converting UF6 to UO2
and from the chemical operations involved in scrap recovery.
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2.3.4 Nuclear Power Facilities
As of December 1986, there were 100 operable nuclear power
reactors in the United States, with a total generating capacity
of 85>177 MWe. With only one exception (a high temperature gas
cooled reactor), all of these nuclear power reactors are either
boiling water reactors (BWR) or pressurized water reactors (PWR);
Pressurized water reactors comprise approximately two-thirds of
the light-water generating capacity.
A light water-cooled nuclear power station generates electricity
using the same basic principles as a conventional fossil-fueled
(oil or coal) .power station except that the source of heat used
to produce steam is provided by nuclear fission instead of
combustion.
In a boiling water reactor, the coolant boils as it passes
through the reactor. The resulting steam is passed through a
turbine and a condenser. The condensed steam is then pumped back
into the reactor, The energy removed from the steam by the
turbine is transformed iato electricity by a generator.
The process is tlae same in a pressurized water reactor except
that the reactor coolant water is .pressurized to prevent boiling.
Energy is transferred through a neat exchanger (steam generator)
to a secondary system where the water does boil. Reactor coolant
water is kept at high pressures by maintaining a closed system
and electrically heating water in a tank called the pressurizer.
After passage through the steam generator, the water is returned
to the reactor. Secondary steam turns the turbine, is cooled in
the condenser, and is pumped back into the steam generator.
During the fission process, radioactive fission products are
produced and accumulate within the nuclear fuel. In addition,
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neutrons produced during fission interact within the fuel and
coolant to produce radioactive activation products. A reactor
may experience periodic fuel failure or defects which result in
the leakage of some of the fission and activation products out of
the fuel and into the coolant. Accordingly, a typical light
water reactor will experience build-up of radioactive fission and
activation products within the coolant. For both PWRs and BWRs
the radioactive contaminants which accumulate within the coolant
are the source of radioactive emissions from the facility.
2.4 DEPARTMENT OF DEFENSE FACILITIES
The Department of Defense (DOD) operates a number of facilities
that use unsealed sources of radioactive materials. In addition
to three research and test reactors and numerous medical facili-
ties, these include army bases that perform research and evalua-
tion of munitions using depleted uranium and naval shipyards that
service the Navy's nuclear-powered fleet.
The army bases that conduct research and development of muni-
tions using depleted uranium metal are licensed by the NRG. Ac-
tivities conducted at these facilities involve test firings and
evaluations of various experimental and stockpile depleted
uranium munitions such as armor piercing shells. At facilities
performing research and development, activities can include the
small-scale fabrication of depleted uranium projectiles. This
fabrication can include forging, shaping, and grinding of deplet-
ed uranium metal.
Nine naval shipyards construct, refuel, maintain, and overhaul
the submarines and ships of the Navy's nuclear-powered fleet:
Mare Island Naval Shipyard in Villejo, CA; General Dynamic's
Electric Boat Division, Groton, CT; Pearl Harbor Naval Shipyard,
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Pearl Harbor, HI; Portsmouth Naval Shipyard, Kittery, ME;
Ingallas Shipbuilding Division, Pascagoula, MI; U.S. Naval
Station and Naval Shipyard, Charleston, SC; Newport News Ship-
building and Drydock_Co.v, Newport News, VA; Norfolk Naval Ship-
yard, Portsmouth, VA; and Puget Sound Naval Shipyard, Bremerton,
WA.
In addition to the normal shipyard functions of construction,
maintenance and overhaul, these shipyards construct, test, re-
fuel, and maintain =the pressurized water reactors used to power
the nuclear fleet. The primary source of radioactive emissions
at naval shipyards is from the facilities that process and pack-
age radioactive wastes. These facilities handle solid low-level
radioactive wastes such as contaminated rags, paper, filters, ion
exchange resins, and scrap materials. Waste materials are sorted,
surveyed, and packaged for shipment.to disposal sites.
All effluent air systems at waste handling facilities are moni-
tored during, operation and-equipped with HEPA filters. Environ-
mental monitoring at these waste, handling facilities indicates
that the concentration of activity in the effluent air is actual-
ly lower than the background activity in the intake air (RI82).
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REFERENCES
AEC74 U.S. Atomic Energy Commission, Fuels and Materials
Directorate of Licensing, Environmental Survey of the
Uranium Fuel Cycle, April 1984. :
AM86 Amersham Corporation, "Products and Services for the Life
Sciences," Arlington Heights, IL, 1986.
BR84 Bremer, P.O., "Pharmaceutical FormPackaging," Chapter 4
in Safety and Efficiency of Radiopharmaceuticals, Martinus
Nijhoff Publishers, Boston, 1984.
CE79 Cehn, J.I. et al., A Study of Airborne Radioactive
Effluents From the Radiopharmaceutical Industry, Teknekron,
Inc., McLean, VA, 1979.
CO81 Cook, J.R., A Survey of Radioactive Effluent Releases From
Byproduct Material Facilities, NUREG-0819, U,S. Nuclear
Regulatory Commission, Washington, DC, 1981.
Do88 Dolezal, W., personal communication with D. Goldin,
SC&A, Inc. September 1988.
FDA86 U.S. Food and Drug Administration, "Abbreviated Summary of
Approved Radiopharmaceutical Drug Products," Washington, DC,
1986.
GA86 Gallagher, B., Ph.D., New England Nuclear, personal
communication, June 1986.
ICN ICN Biomedicals, Inc., "ICN Radiochemicals," Irvine, CA, no
date.
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KN84 Knapp, F.F., and Butler, T.A., editors, "Radionuclide
Generators: New Systems for Nuclear Medicine
Applications," ACS Symposium Series 241, American Chemical
Society, Washington, DC, 1984.
MA85 Mallinckrodt, Inc., "Product and Physical Data Radio-
pharmaceuticals," St. Louis, MO, 1985.
ME86 Medi-Physics, "Technical-Product Descriptions," Richmond,
CA, 1986.
ME78 Medical Economics Company, Litton Division, Physician's
Desk Reference for Radiology and Nuclear Medicine 1978-9,
Oradell, NJ, 1978.
ME83. Merck Company, Merck Index, 1983.
NEN New England Nuclear, "Radiopharmaceuticals and Nuclear
Medicine Sources," North Billerica, MA, no date.
NEN85 New England Nuclear, "Research Products 1985-6," North
, Billerica, MA, 1985.
NEN83 New England Nuclear, "Sources and Accessories .for Nuclear
Medicine," North Billerica,.MA, 1983.
NRC78 U.S. Nuclear Regulatory Commission, Radioactivity in
Consumer Products, NUREG/CP-0001, Washington, DC, 1978.
RA83 Rayudu, G., editor, Radiotracers for Medical Applications:
Volumes 1 and 2, CRC Press, Inc., Boca Raton, FL, 1983.
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RES6 Reba, Richard C., M.D., Director of George Washington
University Nuclear Medicine Division, personal
communication, June 1986.
RI82 Rice, P.D., Sjoblom, G.L., Steele, J.M., and Harvey, B.F.,
Environmental Monitoring and Disposal of Radioactive Wastes
from U. S. Naval Nuclear-Powered Ships and Their Support
Facilities, Report NT-82-1, Naval Nuclear Propulsion
Program, Department of the Navy, Washington, DC, 1982.
SQ85 Squibb Diagnostics, "Technical Product Descriptions," New
Brunswick, NJ, 1985.
SU84 Sutter, S.L. et al., Emergency Preparedness Source Term
Development for the Office of Nuclear Material Safety and
Safeguards-Licensed Facilities, NUREG/CR-3796, prepared by
Pacific Northwest Laboratory for the U.S. Nuclear .
Regulatory Commission, Washington, DC, 1984.
TU80 Tubis, M. and Wolf, W., Radiopharmacy, John Wiley and Sons,
New York, 1980.
US85 U.S. Pharmacopeia and the National Formulary, Washington,-
DC, 1985.
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CHAPTER 3. DETERMINATION OF EMISSIONS
3.1 INTRODUCTION
This chapter provides background on the methods approved by the
Agency for use by the owners or operators of NRC-licensed and
non-DOE Federal facilities in determining the quantities of
radioactive materials emitted to the atmosphere.
Assessment of - the doses resulting from the release of radioactive
materials into the air begins with a determination of the source
term. Since dose is partly a function of the chemical and physi-
cal forms of the radionuclides in the effluent, a fully defined
source term includes the emission rate and the physical charac-
teristics of each chemical species in the effluent.
However, to demonstrate compliance with the standard, the source
term need not be determined in such detail. In lieu of requiring
the complex analytical procedures necessary to determine the
exact composition of the source term, the approved procedures for
determining compliance (see Chapter 4) incorporate dosimetrically
conservative assumptions (i.e., assumptions that maximize the
dose) about the physical characteristics and chemical forms of
the radionuclides in the effluent. These conservative assump-
tions allow the source term to be defined simply by the emission
rate for each radionuclide.
As discussed in Section 3.4, emissions may be calculated based on
monitoring data obtained using EPA-approved sampling and analyti-
cal methods. However, since many of the facilities covered by
the NESHAP do not have monitoring data on the quantities of radio-
nuclides released to the atmosphere, the Agency has derived air-
borne emission factors and effluent control adjustment factors
applicable to the facilities covered by the NESHAP. An airborne
3-1
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emission factor is defined as the quantity of the material re- ,
leased to the air per unit time divided by the quantity of the
material handled in an unsealed form over the same period of
time. The owner or operator of a facility covered by the NESHAP
may apply these EPA-approved emission factors and effluent
control adjustment factors to the quantities of radioactive
materials handled annually in unsealed form to determine the
facility's emissions.
The remainder of this chapter is organized as follows: Section
3.2 describes the basis for the EPA-approved emission factors and
presents these factors and their limitations; Section 3.3 pre-
sents the EPA-approved effluent control adjustment factors and
the basis for the values assigned to each'type of control; and
Section 3.4 presents the basis for the EPA-approved sampling and
analytical methods that may be used to calculate emissions.
3.2 EMISSION FACTORS
The fraction of material released to the atmosphere depends on
the physical and chemical form of the radionuclide and the han-
dling that the material undergoes. By analyzing the currently
available emissions data for a range of radionuclides in various
physical forms and undergoing differing processing/handling com-
binations, the Agency has derived generic "upper-limit" emission
factors which, within the restriction given, can be used to esti-
mate emissions conservatively. The approved emission factors
apply to uncontrolled releases. Thus, when controls are used,
credit can be taken for the control efficiency by applying the
effluent control adjustment factors presented in Section 3.3.
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Section 3.2.1 describes the approaches taken to obtain data and
to derive the emission factors. Section 3.2.2 presents the ap-
proved emission factors and the limitations on their use.
3.2.1 Derivation of Emission Factors
The Agency conducted an extensive search for data suitable for
deriving emission factors. Such data include reported release
fractions, measured emissions, measured air concentrations, and
measured intakes of radionuclides by workers. The approaches
taken to obtain data are discussed in Section 3.2.1.1. The deri-
vation of emission factors from each category of data is discus-
sed in Section 3.2.1.2.
3.2.1.1 Procedures for Obtaining Data
Approaches used to obtain relevant data for defining airborne
emission factors include:
A keyword search of the following bibliographic
data bases: '
Pollution Abstracts (includes articles from
the journal Health Physics);
International Pharmaceuticals (contains bib-
liographic references to articles on pharma-
ceutical chemistry and pharmacy practices);
The National Technical Information Service
(NTIS) Data Base (contains references to
government-sponsored reports);
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Health Planning and Administration (contains
references on managerial practices in the
medical field); and
Environmental Bibliography (contains general
environmental references).
The search of the primary bibliographic data bases used
combinations of the keywords "airborne emissions,"
"radioactive waste," "radiopharmaceutical," "hospital,"
and "medical."
The Public Document Room File Classification
System of the NRG was searched using relevant key-
words. This system is an automated mechanism for
searching NRC-generated documents, meeting records,
correspondence, etc.
Representatives of a number of hospitals were con-
tacted to collect information on the quantities of
radionuclides emitted to the air. The hospitals
that were contacted are: Washington University
Medical Center; Georgetown University Hospital;
George Washington University Hospital; Howard Uni-
versity Hospital; National Institutes of Health;
Washington Hospital Center; Veterans Administra-
tion Hospital in Washington; Bethesda Naval Hospi-
tal; Children's Hospital; Holy Cross Hospital;
Montgomery County Hospital; and Suburban Hospital.
Representatives of the NRC's Material Licensing
Branch were contacted in order to obtain relevant
information from NRG personnel who license facili-
ties subject to the NESHAP. In addition, three
3-4
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NRG inspectors (in NRC Regions I, II, and III)
were contacted in an attempt to gain air emission
data obtained during inspections.
Representatives of the State Health Departments in
New York, Pennsylvania, Virginia, and Maryland and
the Health Department in the District of Columbia
were contacted in order to collect air emission
monitoring data and radionuclide release estimates
for hospitals. In. addition, a representative of
the Conference of Radiation Control Program
Directors was contacted in an attempt to collect
relevant information. The membership of the Con-
ference comprises all directors of radiation con-
trol programs in the 50 States, U.S. Territories,
and some large municipal agencies.
Members of the Council on Radiopharmaceuticals
. within the Society of Nuclear Medicine were con-
tacted to obtain technical information concerning
radiopharmaceuticals (e.g., physical and chemical
forms and extent of use) and to identify possible
sources of information on airborne releases from
medical facilities.
A printout of a data base maintained by the NRC's
Office for Analysis and Evaluation of Operational
Data was obtained and reviewed. This data base is
a clearinghouse for written incident reports made
to the Commission. In particular, reports from
hospitals involving accidental releases of radioac-
tive material from 1980 through 1985 were reviewed.
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In total, over 70 useful books, and hundreds of reports, publica-
tions, and pieces of correspondence pertaining to the handling
and airborne release of radionuclides were obtained through this
search. After a preliminary screening of these materials, refer-
ences that included emission factors, emissions data, air concen-
trations, or accidental intakes were reviewed in depth to deter-
mine if they included enough information to be useful in deriving
emission factors.
3.2.1.2 Analysis of Data
Analysis of information relating to the derivation of emission
factors varied depending on whether the data were reported as
release fractions, airborne emissions, airborne concentrations,
or worker intakes. The following subsections present the data
obtained and describe the methods and assumptions used to derive
emission factors from each of these types of data. Table 3-1,
presented at the end of this section, summarizes the emission
factors derived from the literature review. The entries in the
table are keyed to the references discussed below.
Emission Factors Based on Reported Release Fractions
To be useful, information on the physical form of the material
and the handling or processing associated with the reported,
release fraction is needed. Unfortunately, since the NRC's reg-
ulations are based on Maximum Permissible Concentrations, such
direct measurements are rarely made. Only two references
were obtained which provided release fractions based on measure-
ments (CE79, EI83).
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In A Study of Airborne Radioactive Effluents from the Radiopharm-
aceutical Industry (CE79). the airborne emissions from two unnamed
hospitals were monitored: one for xenon-133 and the other for
iodine-131.
The emissions of xenon-133 from a patient administration proce-
dure was simulated by loading xenon in a spirometer bell jar.
The spirometer was subsequently vented through a wall duct which
connected to the main hospital exhaust system. A grab sample was
taken from within the exhaust system and analyzed for its xenon
content. No emission controls were present in the exhaust system
between the point of release and the point where the sample was
taken. The experiment was repeated four times. Based on the
flow rate of the exhaust system, an emission factor of 4E-1 was
calculated.
At the second hospital, emissions of iodine-131, resulting from
two separate patient applications of sodium iodide in a labora-
tory, were measured. The patient administrations took place over
a 41-minute period. Samples were taken over a period of 81 min-
utes to span the administration period and a prudent additional
amount of time. No effluent controls were present between the
point of release and the point where the sample was taken. Based
on the volume of air sampled, the emission factor for iodine-131
was 1E-5.
"The Fraction of Material Released as Airborne Activity During
Typical Radioiodinations" (EI83) describes the results of more
than 150 release fractions calculated for iodine-125 used in re-
search at the Washington University Medical Center. The calcu-
lated release fractions are based on the measured iodine trapped
on activated charcoal samplers, the fumehood and air sampler flow
rates, and the activity used in the experiments. The experiments
involved variations of the chloramine-T, lactoperoxidase, and
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iodine monochloride radipiodination techniques. They reflect
activities in all of the facility's research laboratories. The
mean emission factor is 8E-4. The reference includes a statisti-
cal analysis of the data which indicates that the average emis-
sion factor from a series of 10 radioiodination procedures would
not exceed 1E-3 at the 99 percent confidence level.
Emission Factors Derived from Emissions Data
In order to derive emission factors from emissions data, it is
necessary to know the quantity of material that is being proces-
sed, the form of the material and the processing or handling that
it receives, and the types and efficiencies of any effluent con-
trols used to reduce emissions.
The amount of emissions data available was limited because the
NRG does not apply uniform licensing criteria concerning emis-
sions monitoring and reporting to the facilities subject to the
NESHAP. As a result, most of the facilities subject to the
NESHAP are not required to perform effluent monitoring, and most
of the facilities that are required to monitor emissions-are not
required to report these data. Even when emissions data are re-
ported, the additional information needed to derive emission
factors are infrequently provided.
The NRG conducted a voluntary survey of its licensees in 1980 to
determine the quantities of radionuclides handled and releases of
these materials to air and to water. The resulting data base,
published as NUREG-0819 (CO81), includes responses from 50 per-
cent of the NRC-licensed facilities handling unsealed radioactive
materials.
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Unfortunately, the survey did not request information on how
emissions were determined, the physical forms of the radionu-
clides used and released, or the effluent controls in use. The
lack of information on physical forms and effluent controls
renders the data base unsuitable for developing the pre-effluent
control emission factors of interest. Moreover, examination of
the individual data points suggests that many of the reported
airborne emissions are only rough estimates rather than measured
values. The crudeness of many estimates is evidenced in cases
where releases exceeding the total quantity of material handled
are reported. Thus, the postcontrol emission factors that can be
calculated from the data are highly suspect. Because of these
inherent limitations, the data in NUREG-0819 were not used to de-
termine emission factors.
Only two references were obtained that had sufficient information
to derive emission factors (AL82, BR87). The'first was the "Ap-
plication for Renewal .of Source Materials License: SUB-526,
Docket 40-3392" for Allied Chemical UF6 Conversion Plant (AL82).
The Allied Chemical plant converts yellowcake to uranium hexa-
fluoride. The handling and processing of materials in this
facility is typical of facilities covered by the NESHAP. The
process includes roasting and sizing the yellowcake feed received
in 55-gallon drums from uranium mills, contacting the 0303 feed
with cracked ammonia to reduce it to UC-2, reacting the UO2 with
hydrogen fluoride to form UF4, and reacting the UF4 with fluorine
to form uranium hexafluoride (UF6). The UF6 product is recovered
in cold traps and distilled to yield the pure UFg product.
The application for renewal of the Allied Chemical Plant's source
material license contains detailed airborne emissions data for
the years 1979 through 1981. These emissions are given by indi-
vidual 'release points. The application also contains'detailed
information on the effluent controls used on each release point
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and the process served by each release point. This information
and the knowledge that the plant was operating at capacity during
this period made it possible to calculate emission factors for
the various steps in the uranium conversion process.
Roasting and Sizing
The beneficiation of the crude yellowcake involves roasting and
sizing the yellowcake to provide a uniform feed. The 3-year
average emissions from this handling are 50 kg/yr uranium. Since
12,700 metric tons of U as UF6 are processed annually, this is
equivalent to a post-treatment emission factor of 4E-06, assuming
no loss of uranium in the process.
All of the release points for this process step are equipped with
two baghouse filters in series. Assuming an operational, effi-
ciency of 99.5 percent for these filters (see Section 3.3), a
pretreatment emission factor of 8E-04 is derived.
Reduction
The uniform yellowcake feed is contacted with cracked ammonia to
yield uranium oxide (UO2). The 3-year average emissions from re-
ducing the feed are 17.9 kg/yr U. Since no controls are used on
the release points for the reducing step of the process, the
emission factor calculated for this step is less than 2E-06.
Hvdrofluorination
in the hydrofluorination step the UO2 is contacted with hydrogen
fluoride to yield UF4, or greensalt. Only release points that
handle the effluent from the hydrofluorination process with no
effluent controls are considered. The 3-year average releases
from the hydrofluorination blowers are 46.5 kg/yr U, equivalent
to an emission factor of 4E-06.
3-10
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Fluorination
In the fluorination process, greencake feed (UF4) is reacted with
fluorine to produce the volatile uranium hexafluoride (UFg) prod-
uct. The 3-year average emissions from the fluorination process
are 78 kg/yr U, equivalent to a post-treatment emission factor of
6E-06. All release points from this processing step are treated
to remove fluorine, hydrogen fluoride, and UFg. Treatment in-
cludes spray towers, packed towers, and metal filters. Assuming
that these systems have an efficiency of 99.5 percent, a pre-
treatment emission factor of 1E-03 is calculated. Note that at
this step, the feed has been converted to the hexafluoride form,
a volatile solid.
In making the calculations, it was necessary to assume the quan-
tity of the throughput and to assign efficiencies to the effluent
controls. Since the plant was known to be operating at full ca-
pacity during the period, the annual throughput was assumed to be
12,700 metric tons of U as UFg. The assigned removal efficiency
of 99.5 percent for the effluent controls is based on information
on fabric filters and sintered metal filters presented in Chapter
2 of Control Technology for Radioactive Emissions to the Atmos-
phere at U.S. Department of Energy Facilities (MO84).
The second source of detailed emissions data was a personal com-
munication from a large manufacturer of sealed sources (BR87).
The manufacturer provided posteffluent-.control emissions data
(Ci/yr) and throughputs for 1985, 1986, and the first 6 months of
1987 for a facility handling batches of cobalt-60 'pellets con-
taining several kilocuries of activity. The facility manufac-
tures sources used in' radiation therapy. The process involves
transferring quantities of the cobalt pellets to the individual
sources in a hot cell. The pellets have a high level of surface
contamination, and dusting is evident during the transfer process.
3-11
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Posteffluent control emission factors, calculated by dividing the
measured releases by the quantity throughput, were calculated to
be 1E-13, 1E-11, and 1E-8 for each of the 3 years, respectively.
Assuming an efficiency of 99.5 percent for the single-stage HEPA
filter (see Section 3.3), pretreatment emission factors of 2E-11,
2E-9, and 2E-6 are calculated.
While the increase in these emission factors over the period
could not be definitively explained, it appears that the emis-
sions data for the later years include a significant contribution
from residual contamination in the hot cell.
Emission Factors Derived from Measured Concentration Data
*' ' '
In order to derive emission factors from measured concentration
data, it is necessary to know the activity of the radionuclides
present, the volumetric flow in the vent or area where the
samples were taken, and the efficiency of any effluent controls
between the point where the release occurs and the point where the
samples are obtained. For noncontinuous releases (such as patient
administrations), it is also necessary to know the duration of the
process or procedure. The physical form of the material and the
handling it undergoes must be known to characterize the circum-
stances associated with the release.
While concentration data are abundant in,the literature, only
five references provided enough of the additional data to derive
emission factors (BNH80, BR78, EA80, LU80, WA87).
Air sample reports were prepared for three administrations of
iodine-131 under normal conditions to patients at the Bethesda
Naval Hospital (BNH80). The hospital's radiation safety staff
3-12
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prepared the reports to help demonstrate compliance with the
NRC's maximum permissible concentration (MFC) for iodine-131.
The iodine-131 was administered orally as sodium iodide in a
sodium chloride solution. Two diagnostic doses and one therapeu-
tic dose were given. Following each administration, the amounts
of iodine-131 in samples of the room air were measured, and each
value was divided by the volume of the air sample and corrected
for the efficiency of the sampling filter to yield the air
concentration.
To calculate the quantity of iodine-131 released to the air in
each case, these reported air concentrations were multiplied by
an air flow rate of 0.3 m3/sec and a time duration appropriate
for such patient administrations. The quantity released was
divided by the quantity administered to derive emission factors
of 3E-7, 2E-5, and 5E-5 for the three cases.
The air flow rate of 0.3 m3/sec is based on a review of flow
rates in hoods venting directly to the atmosphere (NCRP89). The
time duration for each case was assumed to be 10 minutes, a typi-
cal period when preparing and administering doses of iodine
(BR78).
The second useful reference described an investigation of occupa-
tional exposure resulting from handling millicurie quantities of
iodine-131 (BR78). These experiments measuring airborne radio-
activity released from liquid and capsule forms of iodine-131
were conducted at the Monongahela Valley Hospital in North
Charleroi, Pennsylvania.
A vial containing liquid (3-6 milliliters and 100-145 millicuries)
iodine-131 was uncapped for 10 minutes to simulate typical
3-13
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conditions for treatment preparation and administration. The ex-
periment was conducted four times. The experiment was repeated
six times with the vial left open for 1 hour to investigate the
effect of time on the quantity of material released. In each ex-
periment, the air concentrations of iodine-131 were measured by a
sampler placed 4 feet from the vial. After correcting the meas-
ured activity for the efficiency of the sampler (78 percent), the
air concentrations were calculated by dividing the measured activ-
ity by the total volume of air that passed through the air sampler.
Three similar experiments were conducted using iodine-131 in cap-
sule form (the exact chemical form of the iodine was not reported
in the study, but iodine-131 in capsules is typically in the form
of sodium iodide). In each case, a vial containing 5 capsules of
iodine-131 (approximately 20 millicuries each) was opened and ex-
posed to the air for 1 hour. The resulting airborne concentra-
tions were determined as described above.
The reported concentrations were multiplied by an assumed air
flow rate of 0.3 m3/sec (based on NCRP89) and the length of time
the vial was left open to determine the total quantities of
iodine-131 released to the air. The quantity released was divided
by the initial quantity present to derive the emission factors.
For the vials containing liquid and left open for 10 minutes, the
iodine emission factors ranged from 3E-7 to 9E-5, with an average
of 2E-5. For the vials with liquid left open for 1 hour, the
emission factors were roughly a factor of two higher, ranging
from 6E-6 to 2E-4, with an average of 4E-5. No increase in air-
borne activity was found in one of the experiments with the
capsules. From the air concentrations for the other two experi-
ments, the derived emission factor was 9E-7.
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The article "Monitoring of Airborne Contamination During the Han-
dling of Technetium-99m and Radioiodine" describes testing at the
Western Infirmary, Glasgow (EA80). The facility provides radio-
pharmaceutical and radiochemicals for hospitals and handles large
activities of technetium-99m, iodine-125, and iodine-131. The
assembly of technetium-99m radiopharmaceutical generators, tech-
netium-99m generator elution, and the division of iodine-125 and
iodine-131 stocks were monitored to determine airborne radioac-
tivity.
Airborne radioactivity resulting from the daily generation of
technetium-99m by Mo-99 generators in a fume cupboard was moni-
tored inside the cupboard. The sampler was placed within 100
millimeters of the work site and operated throughout the elution
procedure. Airborne releases from technetium-99m dose kit as-
sembly (i.e., the incorporation of technetium-99m into a range of
radiopharmaceuticals) were also monitored. Mean airborne concen-
trations, representing 10 measurements of the elution procedure
and 20 measurements of the dose assembly procedure, and the dura-
tion of each procedure were reported.
Airborne releases from the division of stocks of-iodine-125 solu-
tions and the preparation of therapeutic doses from stocks of
iodine-131 solutions were also monitored. While both are in the
x~ '
form of sodium iodide, the iodine-125 solution is acidic while
the iodine-131 solution is formulated to be basic. Division of
the iodine-125 stock involves opening a small screw-capped vial
containing the solution of sodium iodide and transferring'ali-
quot s with a pipette to other vials. During the preparation of
the iodine-131 doses, the sodium iodide solution is transferred
from rubber-capped vials with a syringe to smaller rubber-capped
vials. Mean air concentrations, representing,10 measurements of
each of these activities, and the duration of each activity were
reported. \
3-15
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In order to estimate the quantity of technetium-99m, iodine-125,
and iodine-131 released to the air during each procedure, the mean
concentrations were multiplied by an assumed air flow of 0.3
m3/sec (NCRP89) and the duration of the procedure. The quanti-
ties released were divided by the initial activities present to
derive the following emission factors : 1E-5 for technetium-99m,
generator elution; 7E-6 for the Tc-99m dose assembly procedure;
2E-3 for the division of the iodine-125 stock; and 2E-5 for the
preparation of iodine-131 doses.
The next reference describes a comparison study of radioiodine
volatility from formulations of sodium iodide oral solutions used
to treat thyroid cancer (LU80). This study was performed at the
Tripler Army Medical Center in Honolulu, Hawaii. One of the
formulations contained an oxidant (sodium bisulfite) intended to
decrease the volatility of the iodine in solution.
Vials containing the iodide solution (80 to 176 millicuries) were
uncapped and vented in a hood for 5 to 7 minutes where flowing
air removed any airborne iodine to an exhaust vent on the roof.
Air samples were collected from the rooftop vent using a
constant-flow air sampler, and the resulting airborne concentra-
tions of iodine-131 were reported.
From the airborne concentrations reported for the rooftop vent,
an assumed air flow rate of 0.3 m3/sec (NCRP89), and the maximum
duration of 7 minutes for venting the vials, the total quantity
of iodine-131 released to the air was calculated for each of nine
ventings. Dividing each of the quantities released by the init-
ial quantity yielded the following emission factors. For the
five ventings involving iodine-131 solution not containing sodium
bisulfite, the emission factors ranged from 3E-4 to 1E-3 with a
mean value of about 9E-4. The emission factors for the four
3-16
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ventings involving the reformulated sodium iodide solution were
about a factor of 70 lower, ranging from 2E-6 to 2E-5, with a
mean of 1E-5.
The final reference actually describes a study conducted to
determine the feasibility of developing data-based guidelines for
deciding which specific radiation protection measures should be
used for operations involving unsealed radioactive materials
(WA87). Although the study focuses on protection of workers, one
of the purposes was to develop generic "dispersibility coeffi-
cients." The authors' definition ,of dispersibility coefficients
is identical to the definition of emission factors used in this
report. In the absence of site-specific data, the authors recom-
mend dispersibility coefficients of 1 for gases, 1E-3 for powders,
1E-4 for water-based liquids, and 1E-6 for solids subjected to
sawing, grinding, sanding, or polishing.
The study also presents emission factors derived by the authors
from data collected during site visits to nine facilities. Of
these, two reflect pre-effluent control releases. For the
forging and filing processes at a facility manufacturing nuclear
fuel elements, an emission factor of 1E-7 is reported. For the
incinerator area, where uranium contaminated wastes are burned,
the reported emission factor is 5E-7.
.Emission Factors Derived from Data on Worker Intakes
Measured accidental intakes of radionuclides by workers represent
the last category of data: used to derive emission factors. While
accidental intakes can occur through skin contamination,
ingestion, or inhalation, only intakes by inhalation were consid-
ered. When such data are reported in terms of the inhaled frac-
tion of the activity handled (IFTAH), they can be used to derive
the emission factors of interest.
3-17
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In deriving the emission factors from IFTAH data, it is necessary
to know the fraction of material released that is inhaled by the
worker. Since this quantity is not directly measured, it had to
be estimated theoretically. An upper limit to this fraction,.
0.01 (derived in Appendix B), was used in all of the emission
factor estimates based on measured intakes.
Because of the manner in which the fraction 0.01 was derived,
only IFTAH data on accidental intakes were used to derive
emission factors. Two references present IFTAH data for
accidents (DO66, FR68).
Donth and Maushart present information involving accidental in-
takes from 14 incidents during a 4-year period at the
Karlsruhe Nuclear Research Center in Karlsruhe, Germany (DO66).
Franke et al., via a literature survey, collected data on 20
accidents from which IFTAH data were derived (FR68). These 20
accidents include the 9 incidents reported by Donth and Maushart,
where enough information was available to determine IFTAH.
The 20 accidents reported by Franke et al. were reviewed to deter-
mine the activity that led to the accidental intake and to ensure
that no process containment or effluent controls were present
which would have reduced the potential exposure of the worker.
Fifteen of the reported accidents, ranging from a broken container
to explosions in gloveboxes, were deemed suitable for deriving
emission factors.
Six emission factors for radionuclides in liquid form were de-
rived from the IFTAH data: three for liquids at ambient tempera-
tures and three for liquids at elevated temperatures. For
liquids at ambient temperatures, the emission factors range from
2E-6 to 2E-5. The emission factors for liquids at elevated temp-
eratures range from 3E-4 to 1E-3. Seven emission factors were
3-18
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derived for radionuclides in powder forms. For powders at ambi-
ent temperatures, the six derived emission factors range from 1E-6
to 7E-5. The emission factor for the one accident involving a
powder at elevated temperatures is 1E-3. Finally, two emission
factors were derived for solids: 3E-6 for a solid at ambient
temperatures, and 2E-4 for a solid which was being brazed at
900ฐC.
3-19
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Table 3-1. Summary of Reported and Derived Emission Factors
Radio-
nuclide Description of Process or Activity
Emission
Factor Reference
U-nat
Xe-133
1-131
1-131
1-131
1-131
1-131
1-131
1-131
1-131
1-131
1-125
1-125
MFP
Sr-90
Tc-99m
Tc-99m
Zr/Nb-95
MATERIALS IN GASEOUS FORM
Production and recovery of UF6
4 simulated patient dose administrations
MATERIALS IN LIQUID FORM AT AMBIENT TEMPERATURES
Patient dose administration
Patient dose administration
Patient dose administration
Patient dose administration
4 simulated patient dose administrations with
vials left open 10 minutes
6 experiments with vials left open for 1 hour
10 dose preparations using a syringe to, transfer
solution from a rubber-capped vial to smaller
rubber-capped vials ( solution formulated to
reduce volatility)
5 simulated dose preparations using solutions
formulated to reduce volatility
4 simulated dose preparations using solutions
not buffered to reduce volatility
150 measurements in clinical research
laboratories during radioiodination procedures
10 dose preparations pipetting solution from an
open vial to smaller open vials (solution not
formulated to reduce volatility)
Accidental rupture of container in a laboratory
Accidental spilling in a laboratory
10 elutions from Mo-99/Tc-99m generators
20 preparations of Tc-99m Dose Kits
Accidental spilling in a laboratory
1E-3
4E-1
1E-5
3E-7
2E-5
5E-5
2E-5
4E-5
7E-6
1E-5
9E-4
8E-4
2E-3
5E-5
2E-6
1E-5
7E-6
4E-5
AL82
CE79
CE79
BNH80
BNH80
BNH80
BR78
BR78
EA80
LU80
LU80
EI83
EA80
FR68
FR68
EA80
EA80
FR68
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Table 3-1. Summary of Reported and Derived Emission Factors (continued)
Radio-
nuclide Description of Process or Activity
Emission
Factor Reference
MFP
MFP
Ru-106
Cs-137
Pa-233
Pu-238
Pu-239
Sr-90
Sr-90
U-nat
Eu-152
U-nat
Co-60
Pu-239
U-nat
U-nat
Cs-137
MATERIALS IN LIQUID FORM AT ELEVATED TEMPERATURES
Accidental splashing during distillation in
a laboratory
Accidental splashing during distillation in
a laboratory
Evaporation of sample in a laboratory
MATERIALS IN POWDER FORM AT AMBIENT TEMPERATURES
Accidental rupture of container in a laboratory
Accidental rupture of container in a laboratory
Glovebox explosion
Glovebox explosion
Accidental rupture of window in a laboratory
Broken container in storeroom
Conversion of UO2 to UF4
MATERIALS IN POWDER FORM AT ELEVATED TEMPERATURES
Accidental spill while heating in a hood
Roasting and sizing of yellowcake
MATERIALS IN SOLID FORM AT AMBIENT TEMPERATURES
Manufacture of sealed sources in a hot cell
Explosion of container during cryogenic
operation in laboratory
Forging and filing of uranium billets
MATERIALS IN SOLID FORM AT.ELEVATED TEMPERATURES
Incineration of uranium-contaminated scrap
Brazing of materials at 900ฐ C in a laboratory
7E-4
1E-3
3E-4
FR68
FR68
FR68
2E-5
2E-5
1E-5
1E-6
4E-5
7E-5
4E-6
1E-3
8E-4
2E-9
3E-6
1E-7
-5E-7
2E-4
FR68
FR68
FR68
FR68
FR68
FR68
AL82
FR68
AL82
BR87
FR68
WA87
WA87
FR68
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3.2.2 Approved Emission Factors
The approved emission factors presented in Table 3-2 were deter-
mined from the information given in Section 3.2.1. They were
developed separately for gases, liquids and powders, and solids
and capsules.
These emission factors may not be used for estimating emissions
from uranium mill tailings piles. Instead, use the methods given
in NRC87.
3.2.2.1 Gases
The radionuclides used in gaseous form at the facilities subject
to the NESHAP include argon, carbon-14, krypton, tritium, and
xenon. Additional radionuclides such as sulfur, bromine, and
uranium may also exist in gaseous forms. Krypton and xenon are
used primarily at medical facilities, while carbon-14 and tritium
are widely used in research facilities. Gaseous tritium is also
used to manufacture self-illuminating signs and instruments.
The approved emission factor for all radionuclides in gaseous
form is 1.0. It is recognized that this emission factor will
overestimate the actual emissions in many cases. Clearly, an
industrial facility using gaseous tritium to produce self-illu-
minating signs or a radiopharmaceutical manufacturer producing
xenon for patient administration would not have any product to
sell if all of the gaseous tritium or xenon were released to the
air. While only two emission factors were derived in Section
3.2.1 for gases, the emission factor of 4E-1, measured during a
simulated patient administration of xenon, indicates that the
fractional release 'can indeed approach unity. Thus, based on the
intrinsic properties of gases, the manner in which they are used
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Table 3-2. EPA-Approved Emission Factors
Physical Form-Processing
Emission Factor
Gases -All processes
1.0
Liquids or Powders-Processes involving
temperatures in excess of 100ฐ.C
or intentional dispersion
1.0
Any radionuclide, regardless of physical
form that is intentionally released
into the environment
1.0
Liquids or Powders-Any process not involving
either temperatures in excess of ,
100ฐ C or intentional dispersion
1E-3
Solids-Processes involving temperatures
in excess of 100ฐ C
1.0
**
Solids -Any process not involving temperatures
in excess of 100ฐ C
1E-6
Mo-99 in Mo-99/Tc-99m generators
1E-6
**
Any material with a boiling point of 100ฐ C or less shall be
considered a gas for the purpose of determining the appropriate
emission factor.
Capsules containing radionuclides in liquid or powder form can
be considered to be solids.
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in medical and research applications, and the limited amount of
data available, an emission factor of 1.0 for gases has been
selected.
3.2.2.2 Liquids and Powders
Almost all of the radionuclides used by the facilities subject to
the NESHAP can be obtained in the form of liquids or powders.
Radionuclides used for medical purposes are almost always in the
liquid form for ease of administration. Similarly, radionuclides
used in research are most frequently in the form of liquids or. ,
powders to facilitate handling. Facilities that manufacture
radiopharmaceuticals handle materials in liquid form/ as do the
facilities that produce self-illuminating watches and other
products using tritiated paints. Producers of smoke detectors
and other products using radioactive materials in a solid or
encapsulated form may receive the material in powdered form and
convert it to a solid.
The emission factors derived in Section 3.2.1 for liquids range
from 3E-7 to 2E-3; those derived for powders range from 1E-6 to
1E-3. These emission factors represent more than 250 measured
data points. They reflect such diverse activities as radio-
pharmaceutical dose preparation, patient dose preparation and
administration, beneficiation of powdered ore concentrates,
chemical processing and conversion, routine research activities,
and laboratory accidents. Because of the similarity in the mag-
nitudes of the emission factors derived for liquids and powders,
a single emission factor of 1E-3 is approved for both forms.
This value is roughly equivalent to the maximum emission factor
derived for either liquids or powders.
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The background information on the emission factors given in Table
3-1 was reviewed to determine the limits of their applicability.
Based upon this review, it was determined that some activities to
which these radionuclides are subjected could result in emission
-factors greater than.lE-3. The following paragraphs explain the
basis of this conclusion and present restrictions on the
application of the approved emission factor for liquids and
powders.
The dispersibility of a material to the air is a complex phenome-
non, depending on a number of physical and chemical parameters
associated with the material and its processing. Two significant
parameters associated with high releases are the volatility of
the material and the temperature of the process. Although none
of the derived emission factors for liquids and powders exceed
the magnitude of the approved value, several of the values in
Table 3-1 are roughly comparable. In particular, several of the
derived emission factors for liquid solutions containing iodine
in various chemical forms and some for materials at elevated
temperatures are close to or roughly equivalent to 1E-3.
It is well known that iodine in the elemental form is volatile.
Although solutions that contain iodine used in research and
medicine are generally buffered to reduce the formation of ele-
mental iodine, some of the iodine may be converted to elemental
iodine and released to the air. The extent to which this has oc-
curred in the processes reviewed to derive the emission factors
given in Table 3-1 is not completely known. However, the activi--'
ties reviewed include a variety of medical and research uses of
radioiodines, and the data reflect fractional releases from both
buffered and nonbuffered iodine solutions. Thus, the approved
emission factor of 1E-3 is believed to be adequate for iodine
solutions not subject to elevated temperatures. Lacking data for
other volatile radionuclide solutions, the EPA has determined
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that any material with a boiling point of 10QQ c or less shall be
considered to be a gas (i.e., emission factor = 1) for the pur-
pose of determining emissions.
The data for radionuclides processed at elevated temperatures are
also insufficient to demonstrate the adequacy of the 1E-3 emission
factor for all activities involving elevated temperatures. For
this reason, the EPA is assigning an emission factor of 1 to all
radionuclides that are heated to temperatures of 100ฐ C or high-
er. Although this may be conservative, the Agency wants to be
certain that there is almost no probability that the approved
emission factors will be exceeded in practice. .
Similarly, the data on which the approved 1E-3 emission factor is
based do not cover activities in which radioactive material is
deliberately dispersed in the environment. In the absence of in-
formation relating to such activities, the Agency is assigning an
emission factor of 1 to any activity involving intentional dis-
persion of radionuclides in the environment.
There is also a special case in which an emission factor smaller
than IE-3 is approved for liquids and powders. This is for the
use of molybdenum-99 generators to produce technetium-99m. As
discussed in Chapter 2, these generators are widely used for med-
ical applications. TheNRC's regulations (10 CFR Section
35.14(b)(4)(iii)) stipulate that the amount of molybdenum-99 pre-
sent in the technetium-99m elute that is administered to a
patient cannot exceed 0.001 millicurie/millicurie technetium-99m.
Since the elution of technetium-99m from the generator is essen-
tially a closed system, the only molybdenum-99 which is available
for release to the air is that which is contained in the elute.
Since the physical decay characteristics of molybdenum-99 are
such that only 87 percent of the molybdenum-99 in the generator
can be eluted as technetium-99m, the fraction of molybdenum-99
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contained in the eluted solution cannot exceed 9E-4. Applying
the approved emission factor for liquids to this maximum fraction
in the elute, the result is an airborne emission factor for moly-
bdenum-99 of 9E-7. In practice, this value will be much lower,
since most generators are designed to meet the more stringent
purity requirements of the U.S. Pharmacopoeia, i.e., 0.15 micro-
curie of molybdenum- 9 9 per millicurie technetium-99m. Given
these considerations, the Agency will allow the emission factor
for solids (1E-6) to be applied to molybdenum-99 in molybdenum-
99/technetium-99m generators.
3.2.2.3 Solids
Relatively few of the facilities covered by the NESHAP use radio-
active materials in solid or capsule forms. The most prevalent
use of solid materials is in the manufacture of sealed radiation
sources. A few radiopharmaceuticals, notably iodine-131, are
available as capsules. If a liquid or powder is contained in a
sealed capsule and is not exposed to a temperature exceeding 100ฐ
C, it may be considered to be a solid.
The approved emission factor for solids and capsules is 1E-6.
This value is consistent: with the derived emission factors. It
is valid for all radionuclide/handling processing combinations .
except those involving volatile solids or solids subjected to
elevated temperatures. For solids with a boiling point of 100ฐ C
or less, the material shall be considered a gas, and an emission
factor, of 1 shall: be used. For solid materials subjected to
high-temperature processing, that is heated to 100ฐ C or higher,
an emission factor of 1 shall be used.
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3.3 ADJUSTMENT FACTORS FOR EFFLUENT AIR CONTROL DEVICES
3.3.1 Use of Effluent Air Control Devices
The use of effluent controls can significantly reduce the quan-
tities of radionuclides released to the environment. For example,
based on the emission factor presented in Table 3-2 for liquids,
air emissions for a hospital using 100 millicuries of iodine-131
in solution would be 100 microcuries in the absence -of effluent
controls. If, however, the material were vented through an
exhaust system equipped with charcoal filters, much of the iodine
would be removed before its release to the atmosphere, and the
air releases would be less than 100 microcuries. Similarly, the
releases from a facility handling 1 millicurie of americium oxide
powder would be estimated to be 1 microcurie in the absence of
effluent controls. However, if the material were handled within
a glovebox equipped with a HEPA filter, the actual quantity of
material released to the environment would be a small fraction of
the material reaching the filter. Therefore, in estimating the
airborne emissions, adjustment factors based on the types of
emission controls used can be applied to the uncontrolled emis-
sion factors presented in Table 3-2.
This section describes the emission controls used at facilities
covered by the NESHAP and characterizes their effectiveness. Ap-
proved adjustment factors and the rationale for their selection
are provided for each type of control. These approved adjustment
factors, which are summarized in Table 3-3, can be applied to the
approved uncontrolled emission factors given in Table 3-2 to de-
rive the emission rates needed to determine compliance.
Two classes of controls are used to control airborne releases at
the facilities covered by the NESHAP. The first class comprises
air flow or ventilation systems. The objective of these systems
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Table 3-3. Approved Adjustment Factors for Effluent Controls
Control
Types of
Airborne
Radionuclides
Controlled
Approved
Adjustment
Factor
Comments and
Conditions
HEPA Filters
All particulate 0.01/
forms (e.g., stage
technetium)
Fabric Filters
Particulates
Sintered Metal Filters Particulates
Activated Carbon Iodine Gas
Filters
Douglas Bags:
Held one week or Xenon
longer for decay
Released within
one week
Xenon Traps
Xenon
Xenon
0.1
0.1
0.5/wk
0.1
Not applicable for
gaseous radionuclides;
periodic testing would
be prudent to ensure
high removal efficiency
Monitoring would be
prudent to guard
against tears in filter
Insufficient data to
make recommendation
Efficiency is time
dependent monitoring
is necessary to ensure
effectiveness
Based on xenon half-
life of 5.3 days;
Provides no reduction
to public exposures
Efficiency is time
dependent -- monitoring
is necessary to ensure
effectiveness
3-29
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Table 3-3. Approved Adjustment Factors for Effluent Controls (continued)
Control
Venturi Scrubbers
Types of
Airborne
Radionuc 1 ide s
Controlled
Particulates
Gases
Approved
Adjustment
Factor
0.1
1
Comments and
Conditions
Although Venturis may
remove gases ,
variablility in gaseous
removal efficiency
dictates adjustment
factor for particulates
only
Packed-Bed Scrubbers
Electrostatic
Precipitators
Soluble Gases 0.1
Particulates 0.05
Not applicable for
particulates or
insoluble gases
Not applicable for
gaseous radionuclides
3-30
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is to redirect the radionuclides away from,the workers and to the
outside atmosphere where dilution occurs. These controls, pri-
marily unfiltered fume hoods and vent stacks, do not reduce the
quantity of radionuclides released to the environment. Use of
these types of controls does not affect a facility's emissions,
and no adjustment factor is approved. The second class of con-
trols physically trap and hold radionuclides, thus preventing
their escape to the outside environment. Controls that fall into
this class include filters, bags, and traps. Use of these types
of controls reduce a facility's airborne releases.
3.3.1.1 HEPA Filters
HEPA filters contain continuous sheets of fiberglass paper pleat-
ed back and forth over corrugated separators housed in either a
wooden or steel casing. Individual HEPA filters are designed for
air flows of approximately 1,000 cfm but are frequently arranged
in filter banks to handle larger air flows. HEPA filters are
ideal for capturing particulate contaminants in effluent streams
with moderate flow rates and dust loadings. They provide no pro-
tection against radionuclides emitted in gaseous forms (e.g.,
krypton and xenon).
The integrity and efficiency of HEPA filters can deteriorate if
they are exposed to excessive moisture, heat, corrosive fumes, or
vibration. These conditions might be expected in some industrial
facilities, where harsh chemical processing operations take place
and a wide variety of contaminants is emitted.
Although HEPA filters are rated at a minimum efficiency of 99.97
percent for 0.3 micrometer particulates, empirical data from the
Savannah River Laboratory indicate that the in-service efficiency
for a single-stage HEPA filter is approximately 99.5 percent
3-31
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(SRP79). Moreover, to achieve this efficiency, a HEPA filter
must be installed properly, maintained frequently, and replaced
periodically. For this reason, a conservative adjustment factor
of 0.01 is approved for single-stage HEPA filters. For each ad-
ditional stage of HEPA filters mounted in series, an additional
factor of 100 can be allowed. Thus, adjustment factors of 0.0001
and 0.000001 are approved for two-stage and three-stage HEPA
systems, respectively.
3.3.1.2 Baghouse Filters
A baghouse filter (sometimes referred to as a fabric filter) is
another high-efficiency particulate control system. The baghouse
filter system consists of filter bags of felt (or a similar
material) arranged in tubular fashion in an enclosed housing.
The effluent stream is blown through the filter bags, which trap
the particulates primarily on the collected material which builds
up on the bags. As the buildup of material on the bags increases,
resistance to flow increases. Thus, baghouse filters must be
equipped with a shaking, vibrating, or reverse-flow device to
remove a portion of the collected dust and recondition the bags.
Like HEPA filters, baghouses may be adversely affected by moist-
ure, high temperatures, or corrosive materials in the effluent.
Unlike HEPA filters, they are designed for high air flows and
heavy dust loadings. The rated efficiency of a baghouse filter
can be as high as 99.9+ percent. According to the data presented
in MO84, efficiencies of 99.5 percent are typical for particu-
lates in the respirable range of 10 microns or less. However,
during actual operations, the efficiency will be lower due to
ruptures in the bags and releases during bag reconditioning.
Therefore, an adjustment factor of 0.1 is approved for baghouse
filters.
3-32
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3.3.1.3 Sintered Metal Filters
Sintered metal filters are designed for high-efficiency particu-
late removal when the temperature of the effluent air or corro-
sive contaminants in the effluent make HEPA and baghouse filters
unsuitable. A sintered metal filter is constructed by fusing, a
mass of metal particles together under heat and pressure. Rated
efficiencies are equivalent to HEPA and baghouse filters; i.e.,
greater than 99 percent. However, sintered metal filters are not
commonly used at the facilities covered by the NESHAP. Moreover,
since data on their operational efficiencies could not be ob-
tained, an approved adjustment factor is not provided for these
devices.
3.3.1.4 Activated Carbon Filters
Adsorption on activated carbon is used primarily to control gase-
ous emissions of radioactive iodine. Carbon filters are usually
housed in fume hoods. The carbon is usually treated with
triethylene diamine (TEDA) to guard .against the retention of noble
gases and to enhance the retention of organic radioiodine. Car-
bon filters are rarely used alone; they are usually preceded by
particulate filters to remove materials that could load and clog
the carbon.
The efficiency of a freshly mounted carbon filter, frequently ex-
ceeds 99 percent (MO84). According to Cehn et al. (CE79), how-
ever, the removal efficiency may drop to as low as 90 percent
prior to cartridge replacement. Factors that affect the effic-
iency include the type of charcoal used, the age of the filter,
the iodine concentration in the discharge air, the air flow rate,
the relative humidity, the temperature, the type of impregnants
used, and the extent of iodine loading on the filter. In general,
3-33
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replacement of spent carbon filters is required about once per
year, although this frequency may vary depending on the chemical
purity of the process air and other factors mentioned above.
Based on a conservative estimate of 90 percent for the removal
efficiency, the approved adjustment factor for activated carbon
filters is 0.10.
3.3.1.5 Douglas Bags
The Douglas bag is an inexpensive control device used to limit
xenon releases during patient administrations. The Douglas bag
is a single-use container designed to capture xenon exhaled from
the patient. Following patient administration, xenon within a
Douglas bag is either bled immediately (or within a short time)
into a fume hood or held in a lead tank for radioactive decay
before release.
The efficiency of a Douglas bag depends on the length of reten-
tion of the xenon before it is released to the atmosphere. The
half-life of xenon-133, the principal isotope of xenon used in
medical facilities, is 5.3 days. Therefore, if the xenon is held
for about 5 days prior to release, almost half of the xenon
decays.
Because retention times vary between hospitals and even between
separate departments within the same hospital, it is not possible
to approve a single value for the adjustment factor. If xenon is
released from a Douglas bag in less than 1 week after the time of,
administration, no adjustment factor may be applied to the xenon
emission factor. For each week of retention, the xenon emission
factor may be adjusted by a factor of 0.5. For example, if xenon
were retained for a period of 3 weeks, the adjustment factor
would be 0.5 X 0.5 X 0.5 = 0.125.
3-34
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3.3.1.6 The Xenon Trap
The xenon trap is a relatively_new but increasingly popular
method of controlling xenon releases from hospitals. The system
consists of a pipe filled with an activated carbon cartridge and
an air pump which draws the contaminated air through the carbon
filter. With xenon traps, the air may either be exhausted, to the
outside or recycled back to the clinical area where the adminis-
tration takes place. ' , '-
Cehn et al. (CE79) state that a xenon trap removes 98 percent of
the xenon in an emission stream; Early et al. (EA85) cite a re-
moval efficiency of 95 percent. Problems do occur, however, due
to saturation of the carbon. Because it is quite difficult to
determine saturation, frequent monitoring is necessary to ensure
adequate performance. In addition, the effectiveness of the
carbon cartridge may be reduced due to water-vapor retention.
For protection against water vapor in a patient's exhalation, a
silicagel cartridge is frequently placed in front of the carbon
filter. The silica-gel cartridge requires weekly checks to ensure
the effectiveness of the system.
Therefore, a conservative estimate of 90 percent is assigned to
the efficiency of xenon traps to account for common problems in
their performance. Accordingly, the approved adjustment factor
for a xenon trap is 0.10.
3.3.1.7 Venturi Scrubbers
Venturi scrubbers are used primarily to control particulate
material, although they may be moderately effective ;in control-
ling gases that are soluble or reactive with the scrubber solu-
tion (usually water). In a venturi scrubber, effluent air is
3-35
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forced at high speed through a pipe restriction in which water is
radially introduced. The drop in pressure of the effluent air
caused by the restriction atomizes the water, and the water drop-
lets effectively intercept particulates (CA84).
The particle removal efficiency of venturi scrubbers is reported
to be approximately 95 percent. Their gas removal efficiency is
quite variable (depending on the type of reaction and the time of
contact between the effluent air and the scrubber solution).
Therefore, the approved adjustment factor is conservatively given
as 0.1 for particulates, and no adjustment factor is provided for
gases.
3.3.1.8 Packed-Bed Scrubbers
A packed-bed scrubber is a vessel filled with randomly oriented
packing material such as ceramic rings, spirals, or saddles. A
scrubbing liquid is fed to the top of this packed bed while the
effluent air flows through the bed either concurrently, counter-
currently, or crosscurrently with the scrubber solution. As the
liquid flows through the bed, it wets the packing material, which
traps gaseous emissions.
Packed-bed scrubbers can be highly efficient in removing gaseous
material (sometimes more than 99 percent efficient) but are gen-
erally not effective in removing particulates (BO80). The fac-
tors affecting the efficiency of packed bed scrubbers include the
depth of the packing material, the contact time of the effluent
with the material, the solubility of the gaseous material, the
drop in pressure of the effluent air as it passes through the
packing material, the effluent air flow rate, and the type of
scrubbing liquid used.
3-36
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A conservative estimate of the efficiency of a packed bed scrub-
ber in removing soluble gases is 90 percent. Therefore, the ap-
proved adjustment factor for soluble gases is 0.1. Since these
systems are not designed remove small particulates, an adjustment
factor for particulates is not given.
3.3.1.9 Electrostatic Precipitators
Electrostatic precipitators (ESPs) are costly but very effective
devices for particulate removal. ESPs employ high voltages to
impart a negative electrical charge to particles in the effluent.
The particles are attracted to and collected on positively
charged plates. ESPs usually provide higher removal efficiencies
for particles smaller than 1 micron in diameter than do venturi
scrubbers. Their particle removal efficiencies may exceed 97
percent (SK); however, the ESPs' performance is sensitive to the
electrical resistivity and stickiness of the particles as well as
to the particle size (BO80).
Considering these factors, the approved particulate adjustment
factor for electrostatic precipitators is 0.05 >.
3.4 APPROVED SAMPLING AND ANALYTICAL METHODS
Facilities that wish to determine their emissions empirically may
do so. However, the methods used to obtain and analyze samples
must conform to the following requirements:
1. Effluent flow rate measurements (velocity and
volumetric flow) from point sources shall be made
3-37
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using: Reference Method 2 of Appendix A to 40 CFR
Part 60 for stacks or large vents; or Reference
Method 2A of Appendix A to 40 CFR Part 60 for pipes
and small vents.
2. Effluent sampling points shall be selected using
the criteria of Reference Method 1 of Appendix A to
40 CFR Part 60.
3. Representative samples of the effluent stream shall
be withdrawn in accordance with the guidance pre-
sented in ANSI-N13.1, "Guide to Sampling Airborne
, Materials at Nuclear Facilities."
4. Radionuclides shall be collected and measured using
procedures based on a principle of measurement
equivalent to those described in Method 114 of
Appendix B to 40 CFR Part 61.
The methods and principles of measurement given in Appendixes A
and D of 40 CFR Part 60 are standard sampling and radioanalytical
techniques that have been reviewed for suitability and are in
common use.
If an owner or operator of an existing facility determines that
it is impractical to sample an effluent stream in accordance with
requirements 2 and 3, alternative procedures may be used pro-
vided: the reason the required methods are impractical is fully
documented; the alternative procedures are fully documented; the
alternative procedures will not cause emissions to be significantly
understated; and the emissions do not cause doses in excess of 10
percent of the standard. The owner or operator of a facility
which employs analytical methods not described in Appendix D must
submit the methods to the Agency for review and approval prior to
using them to determine emissions.
3-38
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REFERENCES
AL82
BA83
Allied Chemical UFg Conversion Plant, "Application for Re-
newal of Source Materials License: SUB-526, Docket 40-
3392," Metropolis, IL, July 1982.
Barrish, E.L., et al., "Radioactive Waste Management at
Biomedical and Academic Institutions," Proceedings of an
International Conference on Radioactive Waste Management
Held by the IAEA In Seattle, WA, May 12-20, 1982, 1983.
BNH80 Bethesda Naval Hospital, "Three Air Sample Reports Follow-
ing Normal lodinations at Bethesda Naval Hospital in Sep-
tember and October 1980," unpublished, provided by Capt.
Charles Galley.
B080
BR87
BR78
Bonner, T., et al., Engineering Handbook for Hazardous
Waste Incineration. Monsanto Research Corporation, pre-
pared for the U.S. Environmental Protection Agency,
Washington, DC, 1980.
Brodsky, A., personal communication with a large manufac-
turer of sealed sources, 1987.
Browning, E.J., Banerjee, K., and Reisinger, W.E., "Air-
borne Concentrations of 1-131 in a Nuclear Medicine Labor-
atory," Journal of Nuclear Medicine. Vol. 19, pp. 1078-
1081, 1978.
BU76 Burchsted, C.A., et al., Nuclear Air Cleaning Handbook;
Design, Construction, and Testing of High-Efficiency Air
Cleaning Systems for Nuclear Application. ERDA 76-21,
Distribution Categories UC-11, 70, Oak Ridge National Lab-
oratory, Oak Ridge, TN, 1976.
3-39
-------
CA84 California Air Resources Board and the South Coast Air
Quality Management District, Southern California Hazardous
Waste Incineration: A Feasibility Study, Prepared for
the Southern California Hazardous Waste Management
Project, 1984.
CE79 Cehn, J.I., A Study of Airborne Radioactive Effluents From
the Pharmaceutical Industry, Final Report, Prepared by
Teknekron, Inc., for the U.S. EPA Eastern Environmental
Research Facility, Montgomery, AL, 1979.
CO81 Cook, J., A Survey of Radioactive Effluent Releases From
Byproduct Material Facilities, NUREG-0819, Office of
Nuclear Material Safety and Safeguards, U.S. Nuclear Regu-
latory Commission, Washington, DC, 1981.
DO66 Donth, H.H., and Maushart, R., "Empirical Radiotoxicity
Hazards and General Licensing Exemptions," Health
Physics, Vol. 12, No. 1, pp. 106-108, 1966.
EA80 Eadie, A.S., Horton, P.W., and Hilditch, T.E., "Monitoring
of Airborne Contamination During the Handling of Tech-
netium-99m and Radioiodine," Phys. Med. Biol., Vol. 25,
No. 6, pp. 1079-1087, 1980.
EA85 Early, P.J., Principles and Practice of Nuclear Medicine,
The C.B. Mosby Company, St. Louis, MO, 1985.
EI83 Eichling, J., "The Fraction of Material Released as Air-
borne Activity During Typical Radioiodinations," Proceed-
ings of the 9th Biennial Conference of Campus Radiation
Safety Officers, University of Missouri at Columbia, June
6-8, 1983.
3-40
-------
EG83a EG&G Idaho, Inc., Improved Low-Level Radioactive Waste
Management Practices for Hospitals and Research Institu-
tions, Report No. DE83-015465, 1983.
EG83b EG&G Idaho, Inc., Massachusetts Low-Level Radioactive
Waste Management Survey, Report No. DOE/LLW-19T, 1983.
EPA84 U.S. Environmental Protection Agency, Background Informa-
tion Document (Integrated Risk Assessment), Final Rules
for Radionuclides. Volume II, Report No. EPA 520/1-84-022-
2, Washington, DC, "l-984.
FR68
FR75
Franks, T., Hermann, G., and Hunzinger, W., "A Quantita-
tive Estimation of the Hazards Involved in Work with Ra-
dionuclides," in Radiation Protection - Proceedings of the
First International Congress of Radiation Protection,
Rome, 1966. Vol. 2., edited by W.S. Snyder, et al.,
Pergamon Press, London, pp. 1401-1406, 1968.
Frost, D. and Jammet, H., Manual on Radiation Protection
in Hospitals and General Practice. Volume 2, Unsealed
Sources, World Health Organization, 1975.
Gregory, W.D. and Maille, H.D., "Incineration of Animal
Radioactive WasteA Comparative Cost Analysis," Health
Physics, 29 , pp. 389-392, September 1975.
ICF86 ICF, "Hazardous Waste Incineration Feasibility Study for
Vandenberg Air Force Base," prepared for Headquarters
Space Division, Department of the Air Force, 1986.
GR75
IL86
Illinois Department of Nuclear Safety, Annual Report on
the Survey of Low-Level Radioactive Waste Generators in
Illinois for 1984. 1986.
3-41
-------
LU80 Luckett, L.W., and Stotler, R.E., "Radioiodine Volatiliza-
tion from Reformulated Sodium Iodide 1-131 Solution-,"
Journal of Nuclear Medicine, 21:477-479, 1980.
MO84 Moore, E.B., Control Technology for Radioactive Emissions
to the Atmosphere at U.S. Department of Energy Facilities,
Report No. PNL-4621, Prepared for U.S. EPA Eastern Environ-
mental Radiation Facility by Pacific Northwest Laboratory,
Richland, WA, 1984.
NCRP89 The National Council on Radiation Protection and Measure-
ments, Screening Techniques for Determining Compliance
with Environmental Standards: Releases of Radionuclides
to the Atmosphere, NCRP Commentary No. 3, Revision of
January 1989, Bethesda, MD, January 1989.
NRC87 "Methods for Estimating Radioactive and Toxic Airborne
Source Terms for Uranium Milling Operations," Regulatory-
Guide 3.59, U.S. Nuclear Regulatory Commission, March 1987.
NY84
OZ84
New York State, Low-Level Radioactive Waste Management
Study, Volume Two, Main Report, 1984.
SK
Oztunali, O. and Roles, G., De Minimus Waste Impacts Anal-
ysis Methodology, NUREG/CR-3585, U.S. Nuclear Regulatory
Commission, Washington, DC, 1984.
Skoski, L., et al., Airborne Radioactive Emission Control
Technology, Volume III. Prepared by Dames and Moore for
the Office of Radiation Programs, U.S. Environmental
Protection Agency, Washington, DC, no date.
3-42
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SRP79 Lee','M. W. , and Stoddard, D.H. , Memorandum to J. T.
Buckner, "Statistical Analysis of HEPA Filtration
Systems," DPST-79-359, Savannah River Laboratory, Tech-
nical Division, duPont Company, Savannah, GA, May 1, 1979
HHS83 U.S. Department of Health and Human Services, "Workshop
Manual for Radionuclide Handling and Radiopharmaceutical
Quality Assurance," Food and Drug Administration, Public
Health Service, Rockville, MD, 1983.
WA87 Watson, C.E., and Fisher, D.R., Feasibility Study of a
Data-Based System of Decisions Regarding Occupational
Radiation Protection Measures, NUREG/CR-4856, prepared by
Pacific Northwest Laboratory for the U.S. Nuclear Regula-
tory Commission, Washington, DC, February 1987.
3-43
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CHAPTER 4. COMPLIANCE PROCEDURES AND EXEMPTION CRITERIA
4.1 INTRODUCTION
The NESHAP limits the quantities of radioactive materials that
can be released to the air annually to such quantities that will
not result in any member of the public receiving a dose greater
than the standard. Since the doses resulting from exposure to
radioactive materials cannot be measured directly, they must be
calculated based on the specific radionuclides involved, the mag-
nitude of the exposure for each exposure pathway, and pathway-
specific dose factors derived from dosimetric and metabolic
models.
The significant exposure pathways for radionuclides released to
the air are: air immersion, ground-surface contamination, inhal-
ation, and ingestion. While the magnitude of the exposure from
each of these pathways can, theoretically, be determined by envi-
ronmental measurements, such measurements are difficult and
costly. Moreover, as noted in Chapter 1, at the concentrations
consistent with the limits of the standard, it may not be possi-
ble to distinguish the fraction that is due to the emission from
the part due to background radioactivity.
The difficulties and costs associated with environmental measure-
ments led to the development of mathematical models, such as the
computer code AIRDOS-EPA, for estimating exposures. The actual
exposure caused by the release of radioactive materials into the
air is an extremely complex function. It depends on the kind and
quantity of radionuclides released; the physical configuration of
the facility releasing the materials; the dispersion, transport
and build-up of the radionuclides in the soil and foodstuffs; and
the proximity to the facility of individuals and farms producing
foodstuffs.
4-1
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To account for these factors, the more sophisticated models re-
quire extensive data inputs. For example, the AIRDOS-EPA code
provides for the user to supply: radionuclide-specific release
rates by lung clearance class; radionuclide-specific scavenging
coefficients and deposition velocities; release point-specific
data concerning release height, volumetric flow rate, and heat
content of the effluent air; site-specific meteorological data
including lid height, joint frequency distributions of wind by
stability class, and annual precipitation rates; and site-
specific demographic data including population distribution and
the fraction of food produced within the assessment area.
In general, the more specific the input data, the more precise
will be the estimate of exposure and dose. However, it is not
necessary to determine precisely the doses that result from the
emissions of a given facility to determine whether that facility
is in compliance with the limit established by the NESHAP. Com-
pliance will be unambiguously demonstrated when estimated upper-
bound doses are below the limit of the standard. Furthermore,
valid upper-bound dose estimates may be calculated using dosimet-
rically conservative assumptions in lieu of site-specific data
for any of the parameters affecting dose. .
The validity of upper-bound estimates has been widely recognized.
In its 1984 report on radiological assessment, the National
Council on Radiation Protection and Measurements (NCRP) recom-
mended that " the 'best' model for a given [dose] assessment
will be the model which is easiest to use and which produces re-
sults within an acceptable degree of,accuracy" (NCRP84). In its
Commentary No. 3, "Screening Techniques for Determining Compli-
ance with Environmental Standards: Releases of Radionuclides to
the Atmosphere," the NCRP used this approach to develop a series
of three "Screening Levels" for estimating upperbound doses
(NCRP89). At each successive level, the dose estimate becomes
4-2
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more precise as additional user-supplied site-specific data and
additional computations replace dosimetrically conservative para-
metric values that were assumed at the preceding level.
The EPA has approved the NCRP's Commentary No. 3 as a means of
determining compliance with the NESHAP and has extended the ap-
proach to develop three additional approved compliance proce-
dures. The use of conservative assumptions in the derivation of
the compliance procedures ensures that the level of protection
afforded the public by the limit of the standard is fully
achieved. By providing these compliance procedures, the Agency
will minimize the burden on licensees of determining compliance
with the standard. In addition, the Agency has adopted exemp-
tion criteria to reduce the burden of the reporting and
construction approval requirements of the NESHAP.
Section 4.2 discusses the basis for the models used in developing
the approved procedures for estimating the dose from a given re-
lease. In Section 4.3, each of the procedures is briefly des-
cribed in terms of the user inputs required and the dispersion
models and other assumptions used in the assessment. Section 4.4
presents the Agency's criteria for determining exemption from the
reporting requirements of the NESHAP.
4.2 MODELS USED IN THE COMPLIANCE PROCEDURES
4.2.1 Atmospheric Dispersion Models
The dose received from a given release of radioactive materials
into the air depends on the exposure received through each of the
?* " .. - .
significant environmental pathways. Calculating the magnitudes
of these exposures first entails determining the concentra-
tion of the radionuclides in the air at the point where the most
4-3
-------
exposed individual is located, and, if different, the concentra-
tion at the location where foodstuffs consumed by that individual
are raised.
The air concentration at any point in the environment is a com-
plex function of the quantity of the radioactive material re-
leased, the configuration of the facility from which the material
is released, the distance from the point of the release to the
locations of interest, the prevailing meteorological conditions,
and various depletion processes which remove the radioactive ma-
terial from the effluent plume as it moves from the point of .,
release to the location of the receptor.
The simplest approximation of the air concentration for each of
the exposure pathways is to assume the concentration is the same
as the concentration at the point of release. Since it cannot be
greater than the concentration at the point of release, the maxi-
mum concentration in the environment can be calculated by the
following equation:
C = Q/V
(1)
where,
C = concentration (Ci/m3),
Q = release rate of the radionuclide (Ci/s), and
V = the volumetric flow in the stacl^ or vent from which the
material is released (m^/s).
The release rate (Q) is determined using the approved procedures
presented in Chapter 3. Where the volumetric flow (V) in the
4-4
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stack or vent is not known, the value of 0.3 m3/s may be used as
a default. The NCRP recommends this default, value, based on a
conservative estimate of typical flows in fume hoods (NCRP89)..
Equation 1 does not account for any dispersion of the material in
the atmosphere. This assumption is unrealistic even when the
receptor is close to the point of the release, as the wind does
not always blow in the same direction. To account for this fact,
equation 1 is modified to:
= fQ/V
(2)
where,
f - the frequency of wind toward the direction of the receptor
(dimensionless) .
The frequency of .wind is computed for each of 16 compass points.
For all of the compliance procedures except one, the value of f
is assumed to be 0.25, the value recommended by the NCRP. This
value is conservative, as actual meteorological data sets rarely
show a value greater than 15 percent (f = 0.15) in the predomin-
ate wind direction.
While equations 1 and 2 are appropriate for use in "upper-bound"
dose estimates, they do not account for dilution of the concen-
tration in the effluent plume as it travels from the .point of re-
lease to the location of the receptor. Diffusion and atmospheric
turbulence are the primary processes acting to reduce the concen-
trations in the plume. Secondary removal processes include
gravitational fallout and wet and dry deposition.
The degree of dilution resulting from atmospheric turbulence and
diffusion depends upon the stability of the atmosphere, the joint
4-5
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frequency distribution of wind speed and direction, and the dis-
tance from the point of release to the location of the receptors.
Additional factors that influence dilution include the height at
which the release occurs, the rise of the effluent plume due to
the momentum and/or thermal buoyancy of the gases in the effluent,
and the relationship between the height of the release and the
heights of the building from which the release occurs and sur-
rounding structures.
The importance of each of these factors varies depending upon the
physical configuration of the facility from which the release oc-
curs and the locations of the receptors. Thus, several disper-
sion equations (given in NCRP86) are used in the compliance pro-
cedures to calculate air concentrations. For situations where
the release point is isolated from the perturbed air flow caused
by buildings, the dispersion is calculated based on Gifford's
formulation of the ground-level centerline Gaussian plume
equation (GI68):
fQ
c =
exp
-1/2
(3)
where,
u = mean wind speed (m/sec),
H = height of effluent release (m) , and
y and z = the horizontal and vertical diffusion parameters (m) .
The extent of horizontal and vertical diffusion is a function of
the atmospheric stability and the distance (x) from the point of
release to the point of the receptor. For the annual average air
4-6
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concentrations of interest, neutral atmospheric stability is ap-
propriate (GI75 ). Briggs (BR74) gives the horizontal and ver-
tical diffusion parameters as:
= (0.08k)
= (0.06x)
+ O.OOOlx)
Yd + O.OOlSx)
(4)
(5)
In developing the compliance procedures, the sector-average for-
mulation of the Gaussian model given in equation 6 is used for
those situation's where the point of release (H) is greater than;
2.5 times the height of the building (Hfe) from which the release
occurs:
fQP
C ='
(6)
u
where,
P is the Gaussian diffusion factor given by Fields and Miller
(FI80) as:
2.032
P =
exp
-1/2
(7)
Where H is greater than zero, the concentration is not simply an
inverse function of the distance from the point of release to the
location of the receptors ; over some distance (dependent upon the
height of release), the concentrations will be lower than the max-
imum ground- level concentration at the point where the plume
4-7
-------
touches down. In developing the compliance procedures, the value
of P at the point of maximum ground-level concentration is used
over all such distances to ensure that the maximum exposure is
evaluated.
In cases where buildings perturb the air flow, the Gaussian model
may not apply. Three situations are considered in the develop-
ment of the compliance procedures: the receptor is located in
the same building from which the release occurs, the receptor is
located within the near wake region (wake recirculation zone),
and the receptor is located beyond the near wake region.
When the receptor is located in the same building from which the
release occurs, the concentration at the point of the receptor is
calculated according to the recommendations of Wilson and Britter
(WI82) as follows:
C = Bf
(8)
where,
= the mean wind speed (m/sec) in undisturbed air at the level
of the building ' s roof , and
B0 = an empirically derived constant that accounts for potential
zones of stagnation along vertical surfaces due to building
wakes .
In the procedures, B0 is set equal to 30, which is the -largest
value given by Wilson and Britter. Such a value will occur when
both the point of the release and the receptor are located on the
lower third of the building.
4-8
-------
For the situations where building wakes are a factor, but the re-
ceptor is not living in the building from which the release oc-
curs, the appropriate dispersion model depends on whether the re-
ceptor is located in the wake recirculation zone (near "wake).
The extent of the wake recirculation zone is a function of the
cross-sectional area of the building from which the release oc-
curs. It is defined as including all distances 'x, where x is
less than or equal to the square root of the projected building
area. For this situation, the dispersion equation of Miller and
Yildran is used (MI84):
fQ
C =
(9)
where,
hb = the smaller of the building height or width, and
K = a constant with the value 1m.
Where the receptor is located outside the wake recirculation
zone, the Gaussian model (equation 6) is modified by replacing
the diffusion parameter P with a diffusion parameter B, which ac-
counts for the reduced concentrations due to trapping and recirc-
ulation of the effluent within the near wake, zone. Values of B
for various building sizes and distances are presented in Figure
4 of NCRP's Commentary No. 3 (NCRP89).
Where plume rise due to momentum or buoyancy are considered they
are calculated according to the equations provided by Briggs, and
are defined in the User's Guide for COMPLY (EPA89b).
4-9
-------
4.2.2 Models Used to Estimate Exposures
Once the ground- level air concentrations are calculated at the
locations of interest, it is necessary to estimate the magnitude
of the exposure received via each of the significant pathways.
For each of the procedures except the "Compliance Model," of the
COMPLY code, the doses resulting from exposures by each of these
pathways are calculated in the manner described in NCRP's Commen-
tary No. 3 (NCRP89).
For the "Compliance Model" (Level 4 of the COMPLY code) the meth-
ods are generally the same, but there are some differences. The
contribution of daughter radionuclides to the dose from external
exposure is handled internally by the computer program, rather
than being built into the dose factors. Build-up of the daugh-
ters in the food chain is also calculated by the program instead
of being compensated for in the dose factors. The differential
equations in the ensuing discussion are solved using the methods
given by Skrable (Sk74).
The air concentration for estimating the exposure from immersion
and inhalation is handled as follows. The concentration of the
parent, N]_, is calculated from:
dNA1/dt = -
where NAi is the concentration of parent atoms, (= Caj_r/Xi), and
Cair is the concentration of the parent calculated from Caj_r =
Q(C/Q).
The daughter concentration is ,
dNAi/dt = NAi_i -
4-10
-------
The initial condition is Ni = 0 at t = 0 . The equations are
solved for t = T, where T is the transport time from the source
to the receptor (= x/u) , where x is .the distance and u is the
average wind speed. '
The concentration of decay products contributing to the dose from
the exposure to contaminated ground is estimated as follows:
dNQl/dt = vNAi - (Ai + Ahl)NGi
= >NAi
Ahl)NGi;
where v is the deposition velocity and ^i is the environmental
removal constant. All the concentrations are initially set to
zero, and the equations are solved for t = 100 years. '
Grow-in of decay products in the food chain is accounted for as
follows. The concentration of the leaves of plants is given by,
MdNL1/dt = vAfrNA1 - M(Ai + AW)NL;L.
Since Y = m/A, this becomes,
dNL1/dt = (vfr/Y)NA1 - (AX +
The equations for the daughters are ,
dNL1/dt = (vfr/Y)NAi
where M is the plant mass growing on area A, fr is the retention
factor, AW the weathering constant, and Y the biomass per unit
area at harvest. The initial concentrations are all zero. These
differential equations are solved for the N's at t = te, the
length of the growing season. The N's are converted to activity
by the equation Cj_ = AiNi-
4-11
-------
The concentration in the soil available for uptake by the roots
of the plants is given by,
= AvNAi -
MdNRi/dt = AvNAi
where M is the mass of soil in the root zone, A the surface area
of soil on which the radioactivity deposits, v the deposition
velocity, and ^hl the removal constant for harvesting and
leaching. Because P (the areal density of the root zone) is
equal to M/A, the equations become,
+ Xhl)NRl, and
dNR1/dt = v/P)NAl -
dNRi/dt = (v/P)NAi
The concentration in 'the plant is simply
Ci = Bvi(AiNRi)
and the total concentration is the sum of the leaf and soil
uptake concentrations. The equations are solved for t = 100
years for build-up, and the N's .are converted to activity by
After harvest, slaughter, or milking the parent and daughter
nuclides behave according to
dNi/dt = -
4-12
-------
The initial values are the values at the end of the growing
season. The times are the delay times for the vegetables, meat,
and milk. Note that the concentrations are assumed to be
constant from the time the animal eats the forage until it is
milked or slaughtered.
These concentrations are used with the pathway equations in the
NCRP's Commentary No. 3 (NCRP89) to calculate intake and dose.
Tritium and carbon-14 at Level 4 are treated slightly differently
than in the NCRP's Commentary No. 3. The NCRP's approach assumes
that the specific activity of carbon-14 and tritium are the same
in the food as in the atmosphere. Level 4 uses a similar
approach, described in Baker (Ba76). Instead of assuming^ that
the specific activity of tritium is the same in the food product
as in the atmosphere, the Baker method accounts for some dilution
by non-tritiated water. In addition, Level 4 uses the equations
given by Baker directly, rather than using transfer factors.
Wet deposition is treated as follows. According to the
International Atomic Energy Agency (IA82), the washout factor,
W(l/m2), is given by
W = Ncp-j_/2TtXiUj_,
where N is the number of sectors, c is a factor equal to
yr/mm-sec for particulates and 1.2x10-5 yr/mm-sec for iodines,
is the rainfall rate in sector i, m/yr, Xi is the distance from
the source to the receptor, and Ui is the annual average wind
speed.
The "wet deposition velocity" , vw is defined as
Vw = AW/C,
4-13
-------
where Aw is the flux of the material deposited on the ground
(Ci/m2-sec) and C is the air concentration (Ci/m3). The flux is
given by
Aw = QW,
where Q is the annual average release rate.
Combining these equations,
Vw = AW/C = (Q/C)Ncpi/(2TiXiUi) .
The total deposition is then the sum of the wet and dry deposi-
tion velocities. The precipitation rate is taken to be 1 meter
per year.
The only other difference between Level 4 and the other compli-
ance procedures is the value of the parameters. The values used
in Level 4 are those listed in Table 7-2 in Chapter 7 of Volume 1
of the Environmental Impact Statement prepared in support of the
NESHAP (EPA89).
4.2.3 Derivation of Dose Conversion Factors
After calculating the magnitude of, the exposure for each environ-
mental pathway, the resulting, doses to the organs of the body are
calculated by applying appropriate radionuclide-specific dose
factors for each pathway. For the compliance procedures, these
dose factors were derived according to the dosimetric and meta-
bolic models recommended by the International Council of Radiation
Protection (ICRP) in Publication 30. The weighting factors
recommended by the ICRP in Publication 26 are applied to the
organ dose factors to derive the effective whole-body 'dose equiv-
4-14
-------
alent. These 'dose factors are tabulated for each pathway in
either Limiting Values of Radionuclide Intake and Air Concentra-
tion and Dose Conversion Factors for Inhalation, Submersion, and
Ingestion (EPA88) or External Dose Rate Conversion Factors for
Calculation of Dose to the Public (DOE88).
4.3 APPROVED COMPLIANCE PROCEDURES
The compliance procedures developed for use by facilities covered
by the NESHAP are described fully in EPA89a, and EPA89b. These
procedures are based on the air dispersion, terrestrial
transport, and dosimetric and metabolic models described in
Section 4.2, and,, in the case of the first two procedures, the
emission factors developed, by the Agency to estimate maximum
radionuclide emissions. The procedures minimize both the amount
of site-specific data that the user must provide and the computa-
tions that,the user must make to determine compliance. In fact,
all but one of 'the procedures can be performed using only paper
and pencil and a liand calculator. The one procedure, the
"Compliance Model"' of the computer code COMPLY, that does require
a personal computer to perform, is extremely easy to run. The
other procedures can also be performed using the COMPLY computer
code.
Each of the compliance procedures is described briefly in the
following paragraphs.
4.3.1 Procedure 1; Quantity of Material Handled
The first compliance procedure is based on.the quantities of
radionuclides handled at the facility. The Agency has derived a
4-15
-------
Table of Annual Possession Quantities for Environmental Compliance
(see Table 4-1) which gives the quantity of each radionuclide
that can be handled annually without causing any member of the .
public to receive a dose in excess of the limits of the standard.
In deriving the quantities in the table, the Agency assumed the
most exposed individual resides 10 meters from the point of re-
lease and obtains food grown at a location 100 meters from the
point of release. The most conservative model for treating the
dispersion of radionuclides in the air was used to estimate con-
centrations at 10 and 100 meters from the release point, which
was assumed to be at ground level. The quantity of each radio-
nuclide that could be released without exceeding the limit of
the standard was then calculated using the pathway-specific
effective whole-body equivalent dose factors for each radionuclide.
The annual possession quantity for each physical form was then
calculated by applying the appropriate emission factors. Because
of the assumptions made regarding the dispersion of the effluent,
the procedure may be used only by facilities where no one resides
within 10 meters of the release point and no food production occurs
within 100 meters.
To apply this procedure, users at facilities meeting the restric-
tion need only to:
a. determine the quantity of each physical form of each
radionuclide handled at the facility;
b. compute for each physical form of each radionuclide
the ratio of the quantity handled to the annual pos-
session quantity given in Table 1 of Section 61.106;
and
c.
sum the computed ratios,
4-16
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance
Radionuclide
Ac-225
Ac-227
Ac-228
Ag-106
Ag-106m
Ag-108m
Ag-llOm
Ag-111
Al-26
Ara-241
Am-242
Ara-242m
Ara-243
Ara-244
Am-245
Am-246
Ar-37
Ar-41
As-72
As-73
As-74
-As-76
As-77
At- 211
Au-193
Au-194
Au-195
Au-198
Au-199
Ba-131
'Ba-133
Ba-133m
Ba-135m
Ba-139
Ba-140
Ba-141
Ba-142
Be-7
Be- 10
Bi-206
Gaseous
Form*
9.6E-05
1.6E-07
3.4E-03
1.6E+00
2.6E-03
6.5E-06
9.4E-05
6.7E-02
4.0E-06
2.3E-06
' 1.8E-02
2.5E-06
2.3E-06
.4.6E-02
7.0E+00
9.8E-01
1.4E+06
1.4E+00
2.9E-02
6.0E-02
4.3E-03
8.8E-02
7.9E-01
l.OE-02
4.2E-01
3.5E-02
3.3E-03
4.6E-02
1.5E-01
l.OE-02
4.9E-05
9.3E-02
5.8E-01
4.7E+00
2.1E-03
1.3E+00
1.1E+00
.2.3E-02
3.0E-03
3.1E-03
Liquid/
Powder
Forms
9. 6E-02
1.6E-04
3.4E+00
1.6E+03
2.6E+00
6.5E-03
9.4E-02
6.7E+01
4.0E-03
2.3E-03
. 1.8E+01
2.5E-03
2.3E-03
4.6E+01
7.0E+03
9.8E+02
- -
_
2.9E+01
6.0E+01
4.3E+00
8.8E+01
7.9E+02
l.OE+01
4.2E+02
3.5E+01
3.3E+00
4.6E+01
1.5E+02
l.OE+01
4.9E-02
9.3E+01
. ' 5. 8E+02
4.7E+03
2.1E+00
1.3E+03
1.1E+03
2.3E+01
3.0E+00
3.1E+00
Solid
Form*
9.6E+01
1.6E-01
3.4E+03
1.6E+06
2.6E+03
6.5E+00
9.4E+01
6.7E+04
4.0E+00
2.3E+00
1.8E+04
2.5E+00
2.3E+00
4.6E+04
7.0E+06
9.8E+05
_
_
2.9E+04
6.0E+04
4.3E+03
8.8E+04
7.9E+05
l.OE+04
4.2E+05
3.5E+04
3.3E+03
4.6E+04
1.5E+05
l.OE+04
4.9E+01
9.3E+04
5.8E+05
4.7E+06
2.1E+03
1.3E+06
1.1E+06
2.3E+04
3.0E+03
3.1E+03
See footnotes at the end of the table.
4-17
-------
Table 4-1
Annual Possession Quantities for Environmental
Compliance (continued)
Radionuc lide
Bi-207
Bi-210
Bi-212
Bi-213
Bi-214
Bk-249
Bk-250
Br-77
Br-80
Br-80m
Br-82
Br-83
Br-84
C-ll
C-14
Ca-41
Ca-45
Ca-47
Cd-109
Cd-113
Cd-113m
Cd-115
Cd-115m
Cd-117
Cd-117m
Ce-139
Ce-141
Ce-143
Ce-144
Cf-248
Cf-249
Cf-250
Cf-251
Cf-252
Cf-253
Cf-254
Cl-36
Cl-38
Cm-242
Cm-243
Gaseous
Form*
8.4E-06
4;2E-03
4.7E-02
6.0E-02
1.4E-01
7.0E-04
l.OE-01
7.5E-02
1.2E+01
1.5E+00
1.6E-02
9.9E+00
5.6E-01
1.3E+00
2.9E-01
2.7E-02
5.8E-02
1.1E-02
5.0E-03
3.3E-04
4.4E-04
5.4E-02
l.OE-02
5.6E-02
1.3E-01
2.6E-03
1.8E-02
l.OE-01
1.7E-03
2.0E-05
1.7E-06
4.0E-06
1.7E-06
6.4E-06
3.3E-04
3.6E-06
1.9E-04
6.5E-01
6.0E-05
3.3E-06
Liquid/
Powder
Forms
8.4E-03
4.2E+00
4.7E+01
6.0E+01
1.4E+02
7.0E-01
l.OE+02
7.5E+01
1.2E+04
1.5E+03
1.6E+01
9.9E+03
5.6E+02
1.3E+03
2.9E+02
2.7E+01
5.8E+01
1.1E+01
5.0E+00
3.3E-01
4.4E-01
5.4E+01
l.OE+01
5.6E+01
1.3E+02
2.6E+00
1.8E+01
l.OE+02
1.7E+00
2.0E-02
1.7E-03
4.0E-03
1.7E-03
6.4E-03
3.3E-01
3.6E-03
1.9E-01
6.5E+02
6.0E-02
3.3E-03
Solid
Form*
8.4E+00
4. 2E+03
4.7E+04
6.0E+04
1.4E+05
7.0E+02
l.OE+05
7.5E+04
1.2E+07
1.5E+06
1.6E+04
9.9E+06
5.6E+05
1.3E+06
2.9E+05
2.7E+04
5.8E+04
1.1E+04
5.0E+03
3.3E+02
4.4E+02
5.4E+04
l.OE+04
5.6E+04
1.3E+05
2.6E+03
1.8E+04
l.OE+05
1.7E+03
2.0E+01
1.7E+00
4.0E+00
1 . 7E+00
6.4E+00
3.3E+02
3.6E+00
1.9E+02
6.5E+05
6.0E+01
3.3E+00
See footnotes at the end of the table,
4-18
-------
.Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Cm-244
Cra-245
Cm-246
Cm-247
. Cm-248
Cm-249
Cra-250
Co-56
" Co-57
Co- 5 8
Co-58m
Co-60
Co-60m
Co-61
Cr-49
Cr-51
Cs-129
Cs-131
Cs-132
Cs-134
Cs-134m
Cs-135
Cs-136
Cs-137
Cs-138
Cu-61
Cu-64
Cu-67
Dy-157
Dy-165
Dy-166
Er-169
Er-171
Es-253
Es-254
Es-254m
Eu-152
Eu-152m
Eu-154
Eu-155
Gaseous
Form*
4.2E-06
2.3E-06
2.3E-06
2.3E-06
6.4E-07
4.6E+00
1.1E-07
2.4E-04
1.6E-03
9.0E-04
1.7E-01
1.6E-05
4.0E+00
3.8E+00
9.0E-01
6.3E-02
1.5E-01
2.8E-01
1.3E-02
5.2E-05
3.2E-01
2.4E-02
2.1E-03
2.3E-05
4.4E-01
4.0E-01
5.2E-01
1.5E-01
4.4E-01
5.6E+00
8.1E-02
4.0E-01
3.6E-01
2.6E-04
2.3E-05
1.8E-03
1.6E-05
3.5E-01
2.0E-05
5.2E-04
: Liquid/
, Powder
Forms
4.2E-03
2.3E-03
2.3E-03
2.3E-03
6;4E-04
4.6E+03
1.1E-04
2.4E-01
1.6E+00
9.0E-01
1.7E+02
1.6E-02
4.0E+03
3.8E+03
9.0E+02
6.3E+01
1.5E+02
2.8E+02
, 1.3E+01
5.2E-02
3.2E+02
2.4E+01
2.1E+00
2.3E-02
4.4E+02
4.0E+02
5.2E+02
1.5E+02
4.4E+02
5.6E+03
8.1E+01
, 4.0E+02
3.6E+02
2.6E-01
2.3E-02
1.8E+00
1.6E-02
3.5E+02
2.0E-02
5.2E-01
Solid
Form*
4.2E+00
2.3E+00
2.3E+00
2.3E+00
6.4E-01
4.6E+06
1.1E-01
2.4E+02
1.6E+03
9.0E+02
1.7E+05
1.6E+01
4.0E+06
3.8E+06
9. OE+05
6.3E+04
1.5E+05
2.8E+05
1.3E+04
5.2E+01
3.2E+05
2.4E+04
2.1E+03
2.3E+01
4.4E+05
4. OE+05
5.2E+05
1.5E+05
4.4E+05
5 . 6E+06
8.1E+04
4. OE+05
3.6E+05
2.6E+02
2.3E+01
1.8E+03
1.6E+01
3.5E+05
2.0E+01
5.2E+02
See footnotes at the end of the table.
4-19
-------
Table 4-1
Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Eu-156
F-18
Fe-52
Fe-55
Fe-59
Fm-254
Fm-255
Fr-223
Ga-66
Ga-67
Ga-68
Ga-72
Gd-152
Gd-153
Gd-159
Ge-68
Ge-71
Ge-77
H-3
Hf-181
Hg-193m
Hg-197
Hg-197m
Hg-203
Ho-166
Ho-166m
1-123
1-124
1-125
1-126
1-128
1-129
1-130
1-131
1-132
1-133
1-134
1-135
In-Ill
In-113m
Gaseous
Form*
3.2E-03
5.6E-01
4.9E-02
1.4E-01
i:3E-03
1.8E-02
4.0E-03
1.4E-01
5.6E-02
1.1E-01
7.6E-01
3.6E-02
4.4E-06
2.0E-03
6.8E-01
2.3E-04
2.6E+00
l.OE-01
1.5E+01
2.5E-03
9.5E-02
2.4E-01
2.5E-01
5.2E-03
2.8E-01
6.0E-06
4.9E-01
9.3E-03
6.2E-03
3.7E-03
9.3E+00
2.6E-04
4.6E-02
6.7E-03
2.0E-01
6.7E-02
3.2E-01
1.2E-01
4.9E-02
2.1E+00
Liquid/
Powder
Forms
3.2E+00
5.6E+02
4.9E+01
1.4E+02
1.3E+00
1.8E+01
4.0E+00
1.4E+02
5.6E+01
1.1E+02
7.6E+02
3.6E+01
4.4E-03
2.0E+00
6.8E+02
2.3E-01
2.6E+03
l.OE+02
1.5E+04
2.5E+00
9.5E+01
2.4E+02
2.5E+02
5.2E+00
2.8E+02
6.0E-03
4.9E+02
9.3E+00
6.2E+00
3.7E+00
9.3E+03
2.6E-01
4.6E+01
1 6.7E+00
2.0E+02
6.7E+01
3.2E+02
1.2E+02
4.9E+01
2.1E+03
Solid
Form*
3.2E+03
5.6E+05
4.9E+04
1.4E+05
1.3E+03
1.8E+04
4.0E+03
1.4E+05
5.6E+04
1.1E+05
7.6E+05
3.6E+04
4.4E+00
2.0E+03
6.8E+05
2.3E+02
2.6E+06
l.OE+05
1.5E+07
2.5E+03
9.5E+04
2.4E+05
2-.5E+05
5.2E+03
2.8E+05
6.0E+00
4.9E+05
9.3E+03
6.2E+03
3.7E+03
9.3E+06
2.6E+02
4.6E+04
6.7E+03
2.0E+05
6.7E+04
3.2E+05
1.2E+05
4.9E+04
2.IE+06
See footnotes at the end of the table.
4-20
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
In-114m
In-115
In-115m
In-116m
In-117
In-117m
Ir-190
Ir-192
Ir-194
Ir-194m
K-40
K-42
K-43
K-44
Kr-79
Kr-81
Kr-83m
Kr-85
Kr-85m
Kr-87
Kr-88
La-140
La-141
La-142
Lu-177
Lu-177m
Mg-28
Mn-52
Mn-52m
Mn-53
Mn-54
Mn-56
Mo- 9 3
Mo-99**
Mo-101
Na-22
Na-24
Nb-90
Nb-93m
Nb-94
Gaseous
Form*
4.9E-03
2.7E-04
1.4E+00
3.5E-01
1.3E+00
7.6E-02
3.5E-03
9.7E-04
2.5E-01
1.5E-04
6.8E-05
2.9E-01
6.0E-02
4.9E-01
7. OE+00
1.8E+02
2.0E+04
8.4E+02
1.1E+01
2. OE+00
4.2E-01
1.6E-02
1.1E+00
2.3E-01
1.4E-01
3.5E-04
2.1E-02
3.5E-03
5. 2E-01
5.7E-02
2.5E-04
2.5E-01
1.5E-03
5.7E-02
8.4E-01
3.2E-05
2.6E-02
2.5E-02
1.2E-02
6.0E-06
Liquid/
Powder
Forms
4.9E+00
2.7E-01
1.4E+03
3.5E+02
1.3E+03
7.6E+01
3.5E+00
9.7E-01
2.5E+02
1.5E-01
6.8E-02
2.9E+02
6.0E+01
4.9E+02
-
_
-
-
-
-
1.6E+01
1.1E+03
2.3E+02
1.4E+02
3.5E-01
2.1E+01
3.5E+00
5.2E+02
5.7E+01
2.5E-01
2.5E+02
1.5E+00 .
5.7E+01
8.4E+02
3.2E-02
2.6E+01
2.5E+01
1.2E+01
6.0E-03
Solid
Form*
4.9E+03
2.7E+02
1.4E+06
3.5E+05
1.3E+06
7.6E+04
3.5E+03
9.7E+02
2.5E+05
1.5E+02
6.8E+01
2.9E+05
6.0E+04
4.9E+05
'mm
ป
.
-
_
1.6E+04
1.1E+06
2.3E+05
1.4E+05
3.5E+02
2.1E+04
3.5E+03
5.2E+05
5.7E+04
2.5E+02
2.5E+05
1.5E+03
5.7E+04
8.4E+05
3.2E+01
2.6E+04
2.5E+04
1.2E+04
6. OE+00
See footnotes at the end of the table.
4-21
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Nb-95
Nb-95m
Nb-96
Nb-97
Nd-147
Nd-149
Ni-56
Ni-57
Ni-59
Ni-63
Ni-65
Np-235
Np-237
Np-238
Np-239
Np-240
Np-240m
Os-185
Os-191m
Os-191
Os-193
P-32
P-33
Pa-230
Pa-231
Pa-233
Pa-234
Pb-203
Pb-205
Pb-209
Pb-210
Pb-211
Pb-212
Pb-214
Pd-103
Pd-107
Pd-109
Pm-143
Pm-144
Pm-145
Gaseous
Form*
2.3E-03
2.0E-02
2.5E-02
l.OE+00
3.0E-02
1.1E+00
2.0E-03
2.1E-02
2.2E-02
1.4E-01
7.0E-01
3.0E-02
1.8E-06
1.9E-02
l.OE-01
6.5E-01
4.7E+00
9.2E-04
9.0E-01
3.8E-02
2.9E-01
1.7E-02
1.2E-01
6.3E-04
8.3E-07
9.3E-03
9.3E-02
8.3E-02
1.2E-02
1.1E+01
5.5E-05
1.2E-01
6.0E-03
1.2E-01
2.1E-01
8.2E-02
9.4E-01
7.6E-04 '
1.1E-04
5.2E-04
Liquid/
Powder
Forms
2.3E+00
2.0E+01
2.5E+01
. l.OE+03
3.0E+01
1.1E+03
2.0E+00
2.1E+01
2.2E+01
1.4E+02
7.0E+02
3.0E+01
1.8E-03
1.9E+01
'l.OE+02
6.5E+02
4.7E+03
9.2E-01
9.0E+02
3.8E+01
2.9E+02
1.7E+01
1.2E+02
6.3E-01
8.3E-04
9.3E+00
9.3E+01
8.3E+01
1.2E+01
1.1E+04
5.5E-02
1.2E+02
6.0E+00
1.2E+02
2,1E+02
8.2E+01
9.4E+02
7.6E-01
1.1E-01
5.2E-01
Solid .
Form*
2.3E+03
2.0E+04
2.5E+04
l.OE+06
3.0E+04
1.1E+06 .
2.QE+03
2.1E+04
2.2E+04
1.4E+05
7.0E+05
3.0E+04
1.8E+00
1.9E+04
l.OE+05
6.5E+05
4.7E+06
9.2E+02
9.0E+05
3..8E+04
2.9E+05
1.7E+04
1.2E+05
6.3E+02
8.3E-01
9.3E+03
9.3E+04
8.3E+04
1.2E+04
1.1E+07
5.5E+01
1.2E+05
6.0E+03
1.2E+05
2.1E+05
8.2E+04
9.4E+05
7.6E+02
1.1E+02
5.2E+02
See footnotes at the end of the table.
4-22
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Pm-146
Pm-147
Pra-148
Pm-148m
Pm-149
Pm-151
Po-210
Pr-142
Pr-143
Pr-144
Pt-191
Pt-193
Pt-193m
,Pt-195m
Pt-197
Pt-197m
Pu-236
Pu-237
, Pu-238.
Pu-239
Pu-240
Pu-241
Pu-242
Pu-243
Pu-244
Pu-245
Pu-246
Ra-223
.Ra-224
Ra-225
Ra-226
Ra-228
Rb-81
Rb-83
Rb-84
Rb-86
Rb-87
Rb-88
Rb-89
Re-184
Gaseous
Form*
4.4E-05
2.6E-02
1.7E-02
7.6E-04
2.8E-01
1.2E-01
9.3E-05
2.8E-01
l.OE-01
.1.5E+01
6. 4E-02
2.1E-02
4.8E-01
. 1.4E-01
1.1E+00
3.6E+00
7.0E-06
2.3E-02
2.7E-06 ,
2.5E-06
2.5E-06
1.3E-04
2. 5E-06
.'. 3.8E+00
2.4E-06
2.1E-01
4.8E-03
1.3E-04
3.2E-04
1.3E-04
5.5E-06
1.3E-05
4.2E-01
1.4E-03
2.0E-03
1.7E-02
l.OE-02
1.7E+00
6.4E-01
1.8E-03
Liquid/
Powder
Forms
4.4E-02
2.6E+01
1.7E+01
7.6E-01
2.8E+02
1.2E+02
9.3E-02
2.8E-F02
l.OE+02
1.5E+04
6.4E+01
2.1E+01
4. 8E+02
1.4E+02
1..1E+03
3.6E+03
7.0E-03
2.3E+01
2.7E-03
2.5E-03
2.5E-03
1.3E-01
2.5E-03
3.8E+03
2.4E-03
2.1E+02
4.8E+00
1.3E-01
3.2E-01
1.3E-01
5.5E-03
. 1.3E-02
4.2E+02
1.4E+00
2.0E+00
1.7E+01
l.OE+01
1.7E+03
6.4E+02
1.8E+00
Solid
Form*
4.4E+01
2.6E+04
1.7E+04
7.6E+02
2.8E+05
1.2E+05
9.3E+01
2.8E+05
l.OE+05
1.5E+07
6.4E+04
2.1E+04
4.8E+05
1.4E+05
1.1E+06
3.6E+06
7.0E+00
2.3E+04
2.7E+00
2.5E+00
2.:5E+00
1.3E+02
2.5E+00
3.8E+06
2.4E+00
2.1E4-05
4.8E+03
1.3E+02
3.2E+02
1.3E+02
5.5E+00
1.3E+01
4.2E+05
1.4E+03
2.0E+03
1.7E+04
l.OE+04
1.7E+06
6.4E+05
1.8E+03
See footnotes at the end of the table.
4-23
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Re- 18 4m
Re-186
Re-187
Re-188
Rh-103m
Rh-105
Ru-97
Ru-103
Ru-105
Ru-106
S-35
Sb-117
Sb-122
Sb-124
Sb-125
Sb-126
Sb-126m
Sb-127
Sb-129
Sc-44
Sc-46
Sc-47
Sc-48
Sc-49
Se-73
Se-75
Se-79
Si-31
Si-32
Sm-147
Sm-151
Sm-153
Sn-113
Sn-117m
Sn-119m
Sn-123
Sn-125
Sn-126
Sr-82
Sr-85
Gaseous
Form*
3.6E-04
' 1.9E-01
9.3E+00
3.7E-01
'1.7E+02
3.4E-01
8.3E-02
3.1E-03
2.9E-01
5.9E-04
7.5E-02
2.0E+00
3.9E-02
6.0E-04
1.4E-04
1.8E-03
7.6E-01
2.0E-02
1.8E-01
1.4E-01
4.0E-04
1.1E-01
1.1E-02
l.OE+01
1.6E-01
1.1E-03
6.9E-03
4.7E+00
7.2E-04
1.4E-05
3.5E-02
2.4E-01
1.9E-03
2.3E-02
2.8E-02
1.8E-02
7.2E-03
4.7E-06
1.9E-03
1.9E-03
Liquid/
Powder
Forms
3.6E-01
1.9E+02
9.3E+03
3.7E+02
1.7E+05
3.4E+02
8.3E+01
3.1E+00
2.9E+02
5.9E-01
7.5E+01
2. OE+03
3.9E+01
6.0E-01
1.4E-01
1.8E+00
7.6E+02
2.0E+01
1.8E+02
1. 4E+02
4.0E-01
1.1E+02
1.1E+01
l.OE+04 ,
1.6E+02
1.1E+00
6.9E+00
4.7E+03
7.2E-01
1.4E-02
3.5E+01
2.4E+02
1.9E+00
2.3E+01
2.8E+01
1.8E+01
7.2E+00
4.7E-03
1.9E+00
1.9E+00
Solid
Form*
3.6E+02
1.9E+05
9.3E+06
3.7E+05
1.7E+08
3.4E+05
8.3E+04
3.1E+03
2.9E+05
5.9E+02
7.5E+04
2.0E+06
3.9E+04
6.0E+02
1.4E+02
1.8E+03
7.6E+05
2.0E+04
1.8E+05
1.4E+05
4.0E+02
1.1E+05
1.1E+04
l.OE+07
1.6E+05
1.1E+03
6.9E+03
4.7E+06
7.2E+02
1.4E+01
3.5E+04
2.4E+05
1.9E+03
2.3E+04
2.8E+04
1.8E+04
7.2E+03
4 . 7E+00
1.9E+03
1.9E+03
See footnotes at the end of the table.
4-24
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Sr-85m
Sr-87m
Sr-89
Sr-90
Sr-91
Sr-92
Ta-182
Tb-157
Tb-160
Tc-95
Tc-95m
Tc-96
Tc-96m
Tc-97
Tc-97m
Tc-98
Tc-99
Tc-99m
Tc-101
Te-121
Te-121m
Te-123
Te-123m
Te-125m
Te-127
Te-127m
Te-129
Te-129m
Te-131
Te-131m
Te-132
Te-133
Te-133m
Te-134
Th-226
Th-227
Th-228
Th-229
Th-230
Th-231
Gaseous
Form*
1.5E+00
1.2E+00
2.1E-02
5.2E-04
1.2E-01
2.5E-01
4.4E-04
2.2E-03
8.4E-04
, 9.0E-02
1.4E-03
5.6E-03
7.0E-01
1.5E-03
7.2E-02
6.4E-06
9.0E-03
1.4E+00
3.8E+00
6.0E-03
5.3E-04
1.2E-03
2.7E-03
1.5E-02
2.9E+00
7.3E-03
6.5E+00
6.1E-03
9.4E-01
1.8E-02
6.2E-03
1.2E+00
2.9E-01
4.4E-01
3.0E-02
6.4E-05
2.9E-06
4.9E-07
3.2E-06
8.4E-01
Liquid/
Powder
Forms
1.5E+03
1.2E+03
2.1E+01
5.2E-01
1.2E+02
2.5E+02
4.4E-01 .
2.2E+00
8.4E-01
9.0E+01
1.4E+00
5.6E+00
7.0E+02
1.5E+00
7.2E+01
6.4E-03
9.0E+00
1.4E+03
3.8E+03
6.0E+00
5.3E-01
1.2E+00
2.7E+00
1.5E+01
2.9E+03
7.3E+00
6.5E+03
6.1E+00
9.4E+02
1.8E+01
6.2E+00
1.2E+03
2.9E+02
4.4E+02
3.0E+01
6.4E-02
2.9E-03
4.9E-04
3.2E-03
8.4E+02
Solid
Form*
1.5E+06
1.2E+06
2.1E+04
5.2E+02
1.2E+05
2.5E+05
4. 4E+02
2.2E+03
8.4E+02
9.0E+04
1.4E+03
5.6E+03
7.0E+05
1.5E+03
7.2E+04
6.4E+00
9.0E+03
1.4E+06
3.8E+06
6.0E+03
5.3E+02
1.2E+03
2.7E+03
1.5E+04
2.9E+06
7.3E+03
6.5E+06
6.1E+03
9.4E+05
1.8E+04
6.2E+03
1.2E+06
2.9E+05
4.4E+05
3.0E+04
6.4E+01
2.9E+00
4.9E-01
3.2E+00
8.4E+05
See footnotes at the end of the table.
4-25
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Th-232
Th-234
Ti-44
Ti-45
Tl-200
Tl-201
Tl-202
Tl-204
Tm-170
Tm-171
U-230
U-231
U-232
U-233
U-234
U-235
U-236
U-237
U-238
U-239
U-240
V-48
V-49
W-181
W-185
W-187
W-188
Xe-122
Xe-123
Xe-125
Xe-127
Xe-129m
Xe-131m
Xe-133
Xe-133m
Xe-135
Xe-135m
Xe-138
Y-86
Y-87
Gaseous
Form*
6.0E-07
2.0E-02
5.2E-06
4.0E-01
4.4E-02
1.8E-01
l.OE-02
2.5E-02
2.4E-02
5.9E-02
5.0E-05
1.4E-01
1.3E-06
7.6E-06
7.6E-06
7.0E-06
8.4E-06
4.7E-02
8.6E-06
8.3E+00
1.8E-01
1.4E-03
1.3E+00
1.1E-02
1.6E-01
1.1E-01
l.OE-02
7.6E-02
1.6E+00
6.0E-01
7.0E+00
7.6E+01
2.2E+02
5.2E+01
6.0E+01
7.6E+00
4.2E+00
9.9E-01
2.8E-02
2.3E-02
Liquid/
Powder
Forms
6.0E-04
2.0E+01
5.2E-03
4.0E+02
4.4E+01
1.8E+02
l.OE+01
2.5E+01
2.4E+01 .
5.9E4-01
5.0E-02
1.4E+02
1. 3E-03
7.6E-03
7.6E-03
7.0E-03
8.4E-03
4.7E+01
8.6E-03
8.3E+03
1.8E+02
1.4E+00
1.3E+03
1.1E+01
1.6E+02
1.1E+02
l.OE+01
7.6E+01
1.6E+03
-
2.8E+01
2.3E+01
Solid
Form*
6.0E-01
2.0E+04
5.2E+00
4.0E+05
4 . 4E+04
1.8E+05
l.OE+04
2.5E+04
2.4E+04
5.9E+04
5.0E+01
1.4E+05
1.3E+00
7.6E+00
7. 6E+00
7.0E+00
8 . 4E+00
4.7E+04
8.6E+00
8.3E+06
1.8E+05
1.4E+03
1.3E+06
1.1E+04
1.6E+05
1.1E+05
l.OE+04
7.6E+04
1.6E+06
-
2.8E+04
2.3E+04
See footnotes at the end of the table.
4-26
-------
Table 4-1 Annual Possession Quantities for Environmental
Compliance (continued)
Radionuclide
Gaseous
Form*
Liquid/
Powder
Fprras
Solid
Form*
Y-88
Y-90
Y-90m
Y-91
Y-91m
Y-92
Y-93
Yb-169
Yb-175
Zn-62
Zn-65
Zn-69
Zn-69m
Zr-86
Zr-88
Zr-89
Zr-93
Zr-95
Zr-97
2.5E-04
1.1E-01
4.3E-01
1.8E-02
1.6E+00
7.0E-01
3.8E-01
5.5E-03
2.1E-01
8.6E-02
4.4E-04
2.7E+01
2.0E-01
2.4E-02
2.7E-04
1.6E-02
2.8E-03
6.4E-04
4.6E-02
2.5E-01
1.1E+02
4.3E+02
1.8E+01
1.6E+03
7.0E+02
3.8E+02
5.5E+00
2.1E+02
8.6E+01
4.4E-01
2.7E+04
2.0E+02
2.4E+01
2.7E-01
1.6E+01
2.8E+00
6.4E-01
4.6E+01
2.5E+02
1.1E+05
4.3E+05
1.8E+04
1.6E+06
7.0E+05
3.8E+05
5.5E+03
2.1E+05
8.6E+04
4.4E+02
2.7E+07
2.0E+05
2.4E+04
2.7E+02
1.6E+04
2.8E+03
6.4E+02
4.6E+04
*Radionuclides boiling at 100ฐ C or less must be considered a
gas. Capsules containing radionuclides in liquid or powder form
can be considered .to be solids.
**Mo-99 contained in a generator to produce Technetium-99 can be
assumed to be a solid.
4-27
-------
This procedure will demonstrate compliance with the standard if the
sum of the ratios computed for each radionuclide handled is less
than or equal to unity.
4.3.2 Procedure 2; Concentration Limits
The Agency has also derived a Table of Air Concentration Levels
for Environmental Compliance (see Table 4-2) specifying maximum
concentrations of radionuclides in the effluent air. For each
radionuclide, the value given in the table is the maximum ground-
level air concentration that would not result in a dose exceeding
the standard. This method is extremely conservative in that it
assumes that no dispersion occurs between the point of release
and the point at which the most exposed individual resides, and
that the most exposed individual grows all of his or her own food
at that location. The table of air concentration cannot be used if
the receptor is within 3 stack diameters of the point of release.
This procedure also requires very little data input by the user.
The concentration in the effluent can be either measured or cal-
culated using the EPA-approved emission factor and the volumetric
flow in the vent or stack. Compliance with the standard will be
demonstrated if the concentration in the stack is less than or
equal to four times the value given in the table. The factor of
four accounts for the fact that the wind does not always blow in
the same direction; wind blowing toward the most exposed individual
25 percent of the time is a conservative upper-bound value. If
more than one radionuclide is released, or if releases occur at
more than one release point, the user simply calculates the ratio
of the actual concentration to the concentration given in Table 4-2
for each radionuclide and/or release point and sums the results.
If the sum of the ratios is four or less, compliance with the
standard has been demonstrated.
4-28
-------
Table 4-2. Concentration Levels for Environmental Compliance
Concentration Concentration
Radionuclide (Ci/m3) Radionuclide (Ci/m3)
Ac-225
Ac-227
Ac-228
Ag-106
Ag-106m
Ag-108m .
Ag-llOm
Ag-111
Al-26
Am-241
Am-242
Am-242m:
Am-243 .
Am-244
Am-245
Am-246 ,
Ar-37
Ar-41
As-72
As-73
As-74
As-76
As- 77
At- 2 11
Au-193
Au-194
Au-195
Au-198
Au-199
Ba-131
Ba-133
Ba-133m
Ba-135m
Ba-139
Ba-140
Ba-141
Ba-142
Be-7
Be-lQ
Bi-206
9.1E-14
1.6E-16
3.7E-12
1.9E-09
1.2E-12 >
7.1E-15
9.1E-14
2.5E-12
4.8E-15
1.9E-15
1.5E-11
2.0E-15
1.8E-15
4.0E-11
8.3E-09
1.2E-Q9
1.6E-03
1.7E-09
2.4E-11
1.1E-11
2.2E-12
5.0E-11
1.6E-10
1.1E-11
3.8E-10 .
3.2E-11
3.1E-12
2.1E-11
4.8E-11
7.1E-12 .
5.9E-14
5.9E-11
1.8E-10
5.6E-09
1.3E-12
1.4E-09
1.3E-09
2.3E-11
1.6E-12
2.3E-12
Bi-207
Bi-210
Bi-212
Bi-213
Bi-214
Bk-249
Bk-250
Br-77
Br-80
Br-80m
Br-82
Br-83
Br-84
C-ll
C-14
Ca-41
Ca-45
Ca-47
Cd-109
Cd-113
Cd-113m
Cd-115
Cd-115m
Cd-117
Cd-117m
Ce-139
Ce-141
Ce-143
Ce-144
Cf-248 .
Cf-249
Cf-250
Cf-251
Cf-252
Cf-253
Cf-254 -..
Cl-36
Cl-38
Cm-242
Cm-243 ..--
l.OE-14
2.9E-13
5.6E-11
7.1E-11
1.4E-10
-5.6E-13.
9.1E-11
4.2E-11
1.4E-08
1.8E-09
1.2E-11
1.2E-08
6.7E-10
1.5E-09
l.OE-11
4.2E-13
1.3E-12
2.4E-12
5.9E-13
9. IE- 15
1.7E-14
1.6E-11
8.3E-13
6.7E-11
1.6E-10
2.6E-12
6.3E-12
3.0E-11
6.2E-13
1.8E-14 ,
1.4E-15
3.2E-15
1.4E-15
5.6E-15
3.1E-13
3.0E-15
2.7E-15
7.7E-10
5.3E-14
2.6E-15
4-29
-------
Table 4-2. Concentration Levels for Environmental Compliance (continued)
Concentration Concentration
Radionuclide (Ci/m3) Radionuclide (Ci/m3)
Cm-244
Cm-245
Cm-246
Cm-247
Cm-248
Cm-249
Cm-250
Co-56
Co-57
Co-58
Co-58m
Co-60
Co-60m
Co-61
Cr-49
Cr-51
Cs-129
Cs-131
Cs-132
Cs-134
Cs-134m
Cs-135
Cs-136
Cs-137
Cs-138
Cu-61
Cu-64
Cu-67
Dy-157
Dy-165
Dy-166
Er-169
Er-171
Es-253
Es-254
Es-254m
Eu-152
Eu-152m
Eu-154
Eu-155
3.3E-15
1.8E-15
1.9E-15
1.9E-15
5.0E-16
3.7E-09
9. IE- 17
1.8E-13
1.3E-12
6.7E-13
1.2E-10
1.7E-14
4.3E-09
4.5E-09
1.1E-09
3.1E-11
1.4E-10
3.3E-11
4.8E-12
2.7E-14
1.7E-10
4.0E-13
5.3E-13
1.9E-14
5.3E-10
4.8E-10
5.3E-10
5.0E-11
5.0E-10
6.7E-09
1.1E-11
2.9E-11
4.0E-10
2.4E-13
2.0E-14
1.8E-12
2.0E-14
3.6E-10
2.3E-14
5.9E-13
Eu-156
F-18
Fe-52
Fe-55
Fe-59
Fm-254
Fra-255
Fr-223
Ga-66
Ga-67
Ga-68
Ga-72
Gd-152
Gd-153
Gd-159
Ge-68
Ge-71
Ge-77
H-3
Hf-181
Hg-193m
Hg-197
Hg-197m
Hg-203
Ho- 166
Ho-166m
1-123
1-124
1-125
1-126
1-128
1-129
1-130 '
1-131
1-132
1-133
1-134
1-135
In-Ill
In-113m
4-30
1.9E-12
6.7E-10
5.6E-11
9.1E-12 ,
6.7E-13
2.0E-11
4.3E-12
3.3E-11
6.2E-11
7. IE- 11
9.1E-10
3.8E-11
5.0E-15
2.1E-12
2.9E-10
2.0E-13
2.4E-10
l.OE-10
1.5E-09 ,
1.9E-12
l.OE-10
8.3E-11
1.1E-10
l.OE-12
7.1E-11
7.1E-15
4.3E-10
6.2E-13
1.2E-13
1.1E-13
1.1E-08
9.1E-15
4.5E-11
2.1E-13
2.3E-10
2.0E-11
3.8E-10
1.2E-10
3.6E-11
2.5E-09
^^-^.^
-------
Table 4-2. Concentration Levels for Environmental Compliance (continued)
Concentration Concentration
Radionuclide (Ci/m3) Radionuclide (Ci/m3)
In- 114m
In-115
In-115m
Iri- 116m
, In-117
In-117m
Ir-19Q
Ir-192
Ir-194
Ir-194m
K-40
K-42
K-43
K-44
Kr-79
Kr-81
Kr-83m
Kr-85
Kr-85m
Kr-87
Kr-88
La- 140
La- 141
La- 142
Lu-177
Lu-177m
Mg-28
Mn-52
Mn-52m
Mn-53
Mn-54
Mn-56
Mo-93
Mo-99
Mo- 101
Na-22
Na-24
Nb-90
Nb-93m
Nb-94
9.1E-13
7. IE- 14 :
1.6E-Q9
4.2E-10
1.6E-09
9.1E-11
2.6E-12
9.1E-13
1.1E-10 ,
1.7E-13
2.7E-14
2.6E-10
6.2E-11
5.9E-10
8.3E-09
2,1E-07
2.3E-05
l.OE-06
1.3E-08
2.4E-09
5.0E-1Q
1.2E-11
7.7E-10 '
2.7E-10
2.4E-11
3.6E-13
1.5E-11
2.8E-12
6.2E-10
1.5E-11
2.8E-13
2.9E-10
1.1E-12
1.4E-11
l.OE-09
2.6E-14
2.6E-11
2.6E-11 .
l.OE-11
7. IE- 15
Nb-95
Nb-95m
Nb-96
Nb-97
Nd-147
Nd-149
Ni-56
Ni-57
Ni-59
Ni-63
. Ni-65
Np-235
Np-237
Np-238
Np-239
Np-240
Np-240m
,0s-185
Os- 19 1m
Os-191
Os-193
P-32
P-33
Pa-230
Pa-231
Pa-233
Pa-234
Pb-203
Pb-205
Pb-209
Pb-210
Pb-211
Pb-212
Pb-214
Pd-103
Pd-107
Pd-109
Pm-143
Pm-144
Pm-145
2.2E-12
1.4E-11 .
2.4E-11
1.2E-09
7.7E-12
7.1E-10
1..7E-12
1.8E-11
1.5E-11
1.4E-11
8.3E-10
2.5E-11
1.2E-15
1.4E-11
3.8E-11
7.7E-10
5.6E-09
l.OE-12
2.9E-10
1.1E-11
9.1E-11
3.3E-13
2.4E-12
3.2E-13
5.9E-16
4.8E-12
1.1E-10
6.2E-11
5.6E-12
1.3E-08
2.8E-15
1.4E-10
6.3E-12
1.2E-10
3.8E-11
3.1E-11
4.8E-10 .
9.1E-13
1.3E-13
6.2E-13
4-31
-------
Table 4-2. Concentration Levels for Environmental Compliance (continued)
Concentration Concentration
Radionuclide (Ci/m3) Radionuclide (Ci/m3)
Pm-146
Pm-147
Pm-148
Pm-148m
Pm-149
Em- 151
Po-210
Pr-142
Pr-143
Pr-144
Pt-191
Pt-193
Pt-193m
Pt-195m
Pt-197
Pt-197m
Pu-236
Pu-237
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Pu-243
Pu-244
Pu-245
Pu-246
Ra-223
Ra-224
Ra-225
Ra-226
Ra-228
Rb-81
Rb-83
Rb-84
Rb-86
Rb-87
Rb-88
Rb-89
Re- 184
5.3E-14
1.1E-11
5.0E-12
6.7E-13
4.2E-11
7.1E-11
7.1E-15
1.1E-10
7.1E-12
1.8E-08
4.3E-11
1.8E-11
4.8E-11
3.2E-11
4.0E-10
2.6E-09
5.9E-15
1.9E-11
2.1E-15
2.0E-15
2.0E-15
l.OE-13
2.0E-15
4.2E-09
2.0E-15
2. IE- 10
2.2E-12
4.2E-14
1.5E-13
5.0E-14
3.3E-15
5.9E-15
5.0E-10
3.4E-13
3.6E-13
5.6E-13
1.6E-13
2.1E-09
7 . 1E-10
1.5E-12
Re- 184m
Re- 186
Re- 187
Re- 188
Rh-103m
Rh-105
Ru-97
Ru-103
Ru-105
Ru-106
S-35
Sb-117
Sb-122
Sb-124
Sb-125
Sb-126
Sb-126m
Sb-127
Sb-I29~
Sc-44
Sc-46
Sc-47
Sc-48
Sc-49
Se-73
Se-75
Se-79
Si-31
Si-32
Sin- 147
Sm-151
Sm-153
Sn-113
Sn-117m
Sn-119ra
Sn-123
Sn-125
Sn-126
Sr-82
Sr-85
3.7E-13
1.8E-11
2.6E-10
1.7E-10
2.1E-07
1 . 3E-10
6.7E-11
2.6E-12
2.8E-10
3.4E-13
1.3E-12
2.4E-Q9
1.4E-11
5.3E-13
1.6E-13
1.4E-12
9.1E-10
7.1E-12
7.7E-11
1.7E-10
4.2E-13 '
3.8E-11
9.1E-12 -
1.2E-08
1.7E-10
1.7E-13
1.1E-13
5.6E-09
3.4E-14
1.4E-14
2.1E-11
5.9E-11
1.4E-12
5.6E-12
5.3E-12
1.1E-12
1.7E-12 '
5.3E-15
6.2E-13 ' :
1.8E-12
4-32
-------
Table 4-2."' Concentration Levels for Environmental Compliance (continued)
.;.,' "-""-" Concentration '" '"'( t"'^'." Concentration,
Radionuclide (Ci/m3) 'Radionuclide (Ci/m3)
Sr-85m /
Sr-87m
Sr-89 ; - - '
Sr-90
Sr-91 " '
Sr-92
Ta-182
Tb-157 .
Tb-160 '
Tc-95
Tc-95m
Tc-9'6
Tc-9.6m
Tc-97
Tc-97m
Tc-98
Tc-99
Tc-99m
Tc-101
Te-121
Te-121m
Te-123
Te-123m
Te-125m
Te-127
Te-127m
Te-129
Te-129m
Te-131
Te-131m
Te-132
Te-133
Te-133m
Te-134
Th-226
Th-227
Th-228
Th-229
Th-230
Th-231
1.6E-09,,..
1.4E-09"-'.
1.8E-12 :;:':;:
1.9E-14" "
9.1E-11
2.9E-10 ,'
4.5E-13 ;
.' 2-.5E-12 ' .'.
' 7.7E-13" ' '"'
l.OE-10
1.4E-12
5.6E-12
6.7E-10
7.1E-13
7.1E-12
6.7E-15
1.4E-13
1.7E-09
4.5E-09
l.OE-12
1.2E-13
1.4E-13
2.0E-13
3.6E-13
l.OE-09
1.5E-13
7.7E-09
1.4E-13
9.1E-11
l.OE-12
7.1E-13
9.1E-10
2.2E-10
5.3E-10
3.4E-11
3.8E-14
3. IE- 15
5.3E-16
3.4E-15
2.9E-10
Th-232,..';:
Th-234 . v. '
Ti-44 -
Ti-45," ..:::-
Tl-200
Tl-201...
Tl-202. :
Ti-204 , ;,
Tm-170 "
Tm-171 :
U-230
U-231
U-232
U-233
U-234
U-235
U-236
U-237
U-238
U-239
U-240
V-48
V-49
W-181
W-185
W-187
W-188
Xe-122
Xe-123
Xe-125
Xe-127
Xe-129m
Xe-131m
Xe-133
Xe-133m
Xe-135
Xe-135m
Xe-138
Y-86
Y-87
6.2E-16"'.-','. -
2. 2E- 12 -'.-..
6.2E-15 .
4.8E-10.'',
4.5E-11
l.OE-10
5.0E-12.
1.2E-12 ,
: 3.3E-12 .."'
2.6E-11
1.5E-14
4.2E-11
1.3E-15
7.1E-15
7.7E-15
7.1E-15
7.7E-15
l^OE-11
8.3E-15
4.3E-09
1.3E-10
l.OE-12
1.6E-10
6.7E-12
2.6E-12
7.7E-11
5.3E-13
9.1E-11
1.6E-09
1.1E-11
8.3E-09
9.1E-08
2.6E-07
6.2E-08
7 . lErQg.
9.1E-09
5.0E-09
1.2E-09
3.0E-11
1.7E-11
4-33
-------
Table 4-2. Concentration Levels for Environmental Compliance (continued)
Concentration Concentration
Radionuclide (Ci/m3) Radionuclide (Ci/m3)
Y-88
Y-90
Y-90m
Y-91
Y-91m
2.7E-13
1.3E-11
1.9E-10
2.1E-12
1.3E-09
Zn-65
Zn-69
Zn-69m
Zr-86
Zr-88
9.1E-14 ,
3.2E-08
1.7E-10
2.4E-11
3.1E-13
Y-92
Y-93
Yb-169
Yb-175
Zn-62
8.3E-10
2.9E-10
3.7E-12
4.3E-11
9.1E-11
Zr-89
Zr-93
Zr-95
Zr-97
1.3E-11
2.6E-12
6.7E-13
3.8E-11
4-34
-------
This procedure also requires very little data input by the user.
The concentration in the effluent can be either measured or cal-
culated using the EPA-approved emission factor and the volumetric
flow in the vent or stack. Compliance with the standard will be
demonstrated if the concentration in the stack is less than.or
equal to four times the value given in the table. The factor of
four accounts for the fact that the wind does not always blow in
the same direction; wind blowing toward the most exposed individual
25 percent of the time is a conservative upper-bound value. If
more than one radionuclide is released, or if releases occur at
more than one release point, the user simply calculates the ratio
of the actual concentration to the concentration given in Table 4-2
for each radionuclide and/or release point and sums the results.
If the sum of the ratios is four or less, compliance with the
standard has been demonstrated.
4.3.3 Procedure 3: NCRP Screening Procedures
The third procedure is to calculate doses using the screening
procedures developed by the NCRP. The "Screening Levels'.' pre-
sented in Commentary No. 3 all begin with the quantities of
radionuclides released to the air from the facility. These
quantities may be determined using any EPA-approved method.
The three procedures follow a tiered approach, in which the con-
servative assumptions regarding the dispersion of the effluent
and the location of the farm producing foodstuffs are relaxed as
the user supplies additional site-specific information. At the
first level, the user simply provides the release rates for each
nuclide and the volumetric flow of the release point. The con-
centration of radionuclides at the point, of release is used to
determine the exposure for all pathways. The effective whole-body
dose equivalent is computed by applying the radionuclide-specific
4-35
-------
screening factors supplied by the NCRP to the estimated cbnoentfa-
tions and summing the results. These screening factors incorpdrate
the terrestrial transport models and dosimetric mbdels descfribed''in
Section 4.2 and include doses from all pathways. If" compliance is
demonstrated at this level, the user stops. - ? r " : .--:v:;.' ' ;=.';5C
. -. .-. .- :.. .."o --- '""-'' l'^:- ^^-r '/.-<'.".; o.j K&.iHi'J :,.
If compliance is not shown, the user moves to the second1level. 'At
this level, the user, guided by a decision diagram, selects the ap-
propriate air dispersion model for the facility and calculates^the
concentration of radionuclides in air' at 'the distance where' a - :
receptor lives. The mannerinwhich the calculation is done -.'"'..::
ensures that the assessment is made at the point of:maximum con-
centration. The same' screening factors-used in the first level-are
then applied to this^concentration to compute the resulting effec-
tive whole-body dose equivalents. Again, 'ifcompliance' is"hb;E"-7:;;"
demonstrated, the user goes on to the third level.
At the third level, the user:".ident!ifeies" the ^'location of^tihe near-"
est farms producing foodstuffs and calculates the concentrations
of radionuclides in air for those:locations using:the appropriate'
dispersion model. Screening factors that consider only the:'ln-":";-'
gestion pathway are then applied to this concentration to deter-^'^'
mine the dose from the ingestion pathway, screening factors'-that
include only air immersion] inhalation,-and ground-surface con-r;:
tamination are applied to the concentration in air at the loca-
tion where the most exposed individual resides. The doses from
all pathways are then summed to determine compliance. :l ";..,:&
4.3.4 Procedure 4: Compliance Model of the COMPLY Computer Code
The Agency has developed the COMPLY computer code to assist the
regulated community in determining compliance with the limits of
the NESHAP. The COMPLY code is a computer implementation of the
first three compliance procedures and includes an additional
4-36
-------
level .-- .th^/'Cpmp^^ is
an extension of the NjqRp;1.;^ procedures;,,. differs .frpm'tha third ,':,^
level of the NCRP's Screening Procedures in several ways/ These
differences,, are, dpcumen^d ,fully in..EPA8.9>b.
The air dispersion models, account for plume rise, and, allow the
user to input site-specific meteorological data in the .form of a
wind rose .showing^ actual... frequencies. of wind, in each. jpf the.. 16
directions and the, ,.actual mean, wi,nd speed for each, direction.
Doses resulting from., the^ ingestion. pathway, are more,realistically
calculated by allowing the user to specify the location of the
ne a^S f a5ra.producing,, vegetables and ^he^near^st farm producing
beef v?d-:?*Mk.: -Wl13*^1/ more, reali.stic,; environmental, parameters
(AIRDOS-EPA default values) are used.
is^easvo JPs rW.?^ ^V^ding,, the, user with pn-screen
messages,, and^prompts . f or., the needed data ...^.The^user,. may begin at
the first level and proceed to, higher,, levels as,. needed to demon-
strate compliance. For each assessment mode, the code compares
t*16.:c.OI^Pytฎd doses with^ the :limits, of the. ;standard to, determine^ .
whether complianc.e, ha_s. -been dempnsttrat:ed^,.. The. code also, supplies
t'ie,, u,?e? -With-^a hard copy;report- showing ..the- input, values supplied
and a summary of the computed .doses. The code, .written in
FORTRAN, can be run on any IBM-PC or PC-compatible computer run-
ning MSr-DOS version-2.0.;pr^later, ,and.,,havฑng at lea-st,512 ,..,..
kil9t>Ytfs.:pf memory and either.one floppy disk drive and-,a hard,,
disk- or,twp floppy disk drives. ,. ...... ... .. \ ,r.,.-....
4.4 EXEMPTION CRITERIA
Facilities covered by the NESHAP are subject to the reporting and
approval requirements of Part 61, Subpart I, Sections 61.104(a)
and 61.106(a). The Agency has determined that these requirements
could represent an unnecessary burden on small users. Therefore,
4-37
-------
the Agency has adopted the following exemption criteria for
facilities covered by the these requirements:
o An existing facility will be exempt if the effective
dose equivalent that is caused by all emissions from
the facility is less than 1 mrem per year.
o Any new construction or modification of an existing
facility will not need to file an application for
approval if one of the following conditions is met:
1. The effective dose equivalent that is caused by all
emissions from the facility is less than 1 mrem per
year.
2. The effective dose equivalent that is caused by all
emissions from the new construction or modification is
less than 0.1 mrem per year.
Exemption may be determined using the Table of Annual Possession
Quantities for Environmental Compliance, the Table of Air
Concentration for Environmental Compliance, Screening Levels 1-3
of NCRP Commentary No. 3, or the COMPLY Code.
Exemption may not be determined using other EPA-approved proce-
dures. The EPA believes it is reasonable to require facilities
using other methods to submit their input data and results -for
review.
4-38
-------
.REFERENCES
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4-41
-------
-------
APPENDIX A: ADDITIONAL REFERENCES
The following references, in addition to those listed, in the
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A-l
-------
Achievable," NUREG-0267, U.S. Nuclear Regulatory Commission,
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and W. D. Norwood, Exce'rpta Medica Foundation, Amsterdam, 1968,
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Hospitals and General Practice, Volume 2, "Unsealed Sources,"
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A-4
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A-12
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APPENDIX B: DERIVATION OF EMISSION FACTORS FROM IFTAH DATA
Several of the emission factors presented in Section 3.2 are de-
rived from measurements of the inhaled fraction of the activity
handled (IFTAH). The IFTAH is defined as the ratio of the amount
of activity inhaled by a worker to the amount of activity handled
during the exposure. In order to derive an emission factor from
an IFTAH, it is also necessary to know the fraction of the activ-
ity made airborne from the process which is inhaled by the
worker (F). Once a value for F is obtained, the emission factor
can be derived from:
Emission Factor = IFTAH / F .
Since measured values for F are not available, it is necessary to
estimate them theoretically. Consider a worker handling material
on an open bench in a room with no ventilation and with relative-
ly "still" air moving at a speed of 15 feet per minute (HE64,
SU84). Assume that a release occurs at a point on the bench that
is at a distance r from the worker's breathing zone, and that an
amount q of respirable material is released to the air. Further
assume that, on the average, the ejected material is released
uniformly in all directions into the upper hemisphere of radius
r, and that the air stream that carries the material is suffi-
ciently violent to propel the initial volume of material to dis-
tance r instantaneously. This sudden ejection of the material to
distance r tends to maximize the quantity of material inhaled
before natural convection dissipates the concentration. The re-
entry of material into the breathing zone with time is neglected
in this analysis, since even minimal ventilation will rapidly
carry the airborne material into the ventilation system.
B-l
-------
The concentration inside the hemispherical puff of (2/3)n;r3 at
t = 0 is then be taken to be
C0 = q/(2nr3/3). (!
Assume that at distance r, the "boundary convection area" (the
surface of equal concentration), the velocity vector has magni-
tude v, and half the time points away from the source, and half
the time toward the source of material. (Convection currents
move in random directions.) This allows us to calculate an ef-
fective dilution rate following the initial release of material,
which is not likely to extend beyond about 1 meter before it is
dissipated, regardless of the initial particle and air velocity
(HE64).
The differential fractional dilution per unit time of the con-
centration within the boundary convection area then becomes the
fractional amount of replacement air entering the boundary con-
vection area through one-half the area, which is
- dC/Cdt = = 0.5 v 2Ttr2/(2ur3/3)
= 3 v/2r
(2)
By integrating Eq. 2 from t = 0 to a variable time t, and from Co
to C(t), we obtain the concentration as a function of time:
C(t) = q exp(-At)/(2itr3/3)
(3)
B-2
-------
If we assume a uniform breathing rate, b cm3 air per minute, then
the total amount I of activity inhaled is given by
t=ฐฐ
I = r b C(t) dt = (3bq/2nr3) f exp(-3v/2r) t dt
r
f
t=0 t=0
= bq/nr2 v
(4)
The fractional intake of the amount g released to the air in the
puff is then given by
F. = I/g = b/Tir2 v .
(5)
Letting b = 20,000 cm3/minute (20 breaths per minute times 1,000
cm3 of air per inspiration) and v = 15 feet/minute =7.64 cm/sec,
we obtain a conservative relationship for the value of F as a
function of distance between the source and the breathing zone
F = 333/7.64itr2
= 13.9/r2.
(6)
For r = 1 meter (100 cm), F = 0.0014; for r = 0.66 meters (66 cm,
about 2 feet), F = 0.0032; and for r =30.5 cm (1 foot), F =
0.015. It is not likely that the puff will be maintained in
significant concentrations beyond a distance of approximately one
meter, and Hemeon assumes on the basis of industrial experience
that "arm's length work" takes place at 2 feet and "close work"
at 1 foot (HE64).
B-3
-------
At the breathing rate assumed, inspirations and expirations
each take approximately 1.5 seconds. Air moving at 7.6 cm/sec
will move approximately 12 cm in 1.5 seconds. The radius of the
1-liter volume of air inhaled is
r = Onv/4)1/3 = 6.2
cm.
(7)
Further, the mean time for removal of a particle from the breath-
ing zone consistent with this model is given by
= I/A = 2r/3v =8.7 sec at 100 cm,
(8)
and the half-time for decrease of the concentration at 100 cm is
Tl/2 = 0.693/A = 6 sec.
(9)
This is only two breathing cycles at 100 cm. Thus, if a person
is exhaling rather than inhaling when the release occurs (and
thus also diluting further the concentration by creating add-
itional air movement), then the fractions inhaled will be smaller
than the F values calculated above. On the other hand, if the
worker is beginning a cycle of inhalation when the release occurs,
then higher F values may be expected.
Considering all the factors in the above analysis, we estimate
the following value for the upper limit of the fraction of the
activity released to the air in respirable form that is inhaled
by the worker
(10)
B-4
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The use of a maximum rather than a minimum value for F would
appear contrary to our objective of deriving an upper-limit esti-
mate for the emission factor. However, we are not interested in
estimating the fractional release to air from a single observa-
tion; rather, our objective is to obtain an upper bound for a
distribution of fractional releases. Thus, the following line of
reasoning is applicable here.
The upper bound (e.g., the 95th percentile) of the underlying
probability distribution of worker intakes "would result from
incidents that produce releases of large fractions of the mater-
ial in process, combined with the large fractions inhaled of that
amount released (which could, for example, result from processes
involving proximity of the worker to the material combined with
more frequent and vigorous agitation of the material). Thus, the
"upper bound" fractional release of material to air (for the
distribution of events) will be estimated using an "upper bound"
estimate of fractional intake~'of material made airborne and the
"upper bound" IFTAH values.
A mathematical statement may help to clarify the foregoing.
Let p'(x) = the probability density function of frac-
tional releases x, and
p"(y//x) = the conditional probability density func-
tion of fractional intakes y of that
released, given the fractional amount
released is x.
In general, p"(y//x) can be a function of x as well as y, since
the more violent disturbances (such as explosions) that cause a
release of activity to the air could also disturb the air cur-
rents that transport the material to a worker's breathing zone.
B-5
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However, the influence of the value of x on air currents in the
breathing zone can be assumed to be negligible for the smaller
releases.
Now, the probability density of fractional intake i, P(i), of an
amount of radioactive material placed into process is given by
i) = p'(x) p"
(11)
where
i = x y
(12)
The probability, P1, of an intake fraction greater than or equal
to i = I is then given by
1 - f P(i
) di
(13)
i=0
1 -if
P(xy) ((di/dx)dx+ (di/dy)dy), (14)
i=0
since x and y are assumed to be independent random variables.
Thus, substituting new limits of the variables of integration,
Eg. 13 becomes
P'(i>.I) =
x = I/y y = I/x
1 - f P(xy) y dx - f P(xy) x dy
x=0 y=0
(15)
B-6
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It is noted that x, y, and P(i) can never be negative. Therefore,
higher intake values, i, are more probable when both y and x are
maximized together, so that the ranges of integration of both
terms are minimized, and thus Pf is maximized.
B-7
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