520779006
SEPA Radiological Impact
Caused by Emissions
i
of Radionuclides
into Air in the United States
Preliminary Report
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CONTENTS
Page
Summary S-l
1 Introduction 1-1
2 Sources of Radioactive Emissions from Facilities
Licensed by the Nuclear Regulatory Commission 2.1-1
2.1 Uranium Fuel Cycle 2.1-1
2.1A Uranium Conversion Facilities 2.1A-1
2.IB Fuel Fabrication 2.1B-1
2.1C Light-Water Reactor Nuclear Power Plants 2.1C-1
2.2 High Temperature Gas Cooled Reactor (HTGR) 2.2-1
2.3 Radiopharmaceutical Industry 2.3-1
2.4 Test Reactors 2.4-1
2.5 Research Reactors 2.5-1
2.6 University Reactors 2.6-1
2.7 Shallow Land Burial of Low-Level Radioactive
Wastes 2.7-1
2.8 Plutonium Fuel Fabrication Facilities 2.8-1
2.9 Industrial Users and Other Categories 2.9-1
3 Sources of Emissions from Department of
Energy Facilities
3.1 Hanford Site 3.1-1
3.2 Savannah River Plant 3.2-1
3.3 Idaho National Engineering Laboratory 3.3-1
3.4 Los Alamos Scientific Laboratory 3.4-1
3.5 Lawrence Livermore Laboratory 3.5-1
3.6 Rocky Flats Plant 3.6-1
3.7. Mound Laboratory 3.7-1
3.8 Pantex Plant 3.8-1
3.9 Pinellas Plant 3.9-1
3.10 Sandia Laboratories 3.10-1
3.11 Nevada Test Site 3.11-1
3.12 Argonne National Laboratory 3.12-1
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CONTENTS
Page
3.13 Brookhaven National Laboratory 3.13-1
3.14 Oak Ridge Facilities 3.14-1
3.15 Portsmouth Gaseous Diffusion Plant 3.15-1
3.16 Paducah Gaseous Diffusion Plant 3.16-1
3.17 Ames Laboratory 3.17-1
3.18 Atomics International 3.18-1
3.19 Battelle Columbus Laboratory 3.19-1
3.20 Bettis Atomic Power Laboratory 3.20-1
3.21 Feed Materials Production Center 3.21-1
3.22 Knolls Atomic Power Laboratory 3.22-1
3.23 Shippingport Atomic Power Station 3.23-1
3.24 Reactive Metals, Inc., Company 3.24-1
3.25 Lawrence Berkeley Laboratory 3.25-1
3.26 Fermi National Accelerator Laboratory 3.26-1
3.27 Stanford Linear Accelerator Center 3.27-1
4 Sources of Emissions of Naturally Occurring
Radionuclides 4.0-1
4.1 Uranium Mining 4.1-1
4.2 Uranium Mills 4.2-1
4.3 Phosphate Industry 4.3-1
4.3A Ore Mining and Beneficiation 4.3-1
4.3B Ore Drying and Grinding 4.3-12
4.3C Phosphoric Acid Plant 4.3-17
4.3D Elemental Phosphorus Plant 4.3-26
4.4 Coal-fired Steam Electric Generating Plants 4.4-1
4.5 Metal and Nonmetal Mining and Milling 4.5-1
4.6 Radon from Water 4.6A-1
4.6A Geothermal Power Plant Sites 4.6A-1
4.6B Water Treatment Plants 4.6B-1
iv
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CONTENTS
Page
5 Other Source Categories 5.1A-1
5.1 Department of Defense Facilities 5.1A-1
5.1A U.S. Army Facilities 5.1A-1
5.IB U.S. Navy Facilities 5.1B-1
5.2 Particle Accelerator Facilities 5.2-1
APPENDICES
A Assessment Methodology A-l
B Health Risk Assessment Methodology B-l
C Source Term Calculations C-l
D Glossary of Terms and Abbreviations D-l
E List of Elements E-l
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SUMMARY
Radionuclides are emitted into the air from numerous sources
located throughout the United States, including such facilities as
nuclear power plants and other facilities pertaining to the nuclear
fuel cycle, national defense facilities, industrial plants, research
and development laboratories (including research reactors and
accelerators), medical facilities, certain mining and milling
operations, and fossil fuel combustion plants. As a result of the
operation of these facilities, radionuclides are released into the
atmosphere where they are dispersed into populated areas. Radiation
exposure to the public can then occur by breathing or swallowing
these materials. For some radionuclides, people can also be exposed
from direct radiation that is emitted from a cloud of the material
passing overhead or from'direct radiation emitted when the
radionuclides settle onto the ground.
Section 122 of the Clean Air Act Amendments of 1977, Public
Law 95-95, directed the Administrator of the Environmental
Protection Agency, to review all relevant information and determine
whether emissions of radioactive pollutants into ambient air will
cause or contribute to air pollution which may reasonably be
anticipated to endanger public health.
As a part of this review, the Agency has been assessing the
public health impact resulting from emissions of radionuclides into
air from a broad spectrum of major source categories. This report
presents the initial, preliminary results of these assessments. For
each facility or source category, the following are presented:
- the amount of radionuclides released into the
atmosphere;
- the radiation doses to individuals and
population groups;
- the lifetime risks to individuals;
- the number of fatal cancers in the exposed
population per year of facility operation.
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S-2
These preliminary results were obtained by using the
available data base on emission levels and the current calculational
methods for estimating dose and health risks. Simultaneously with
the development of this report, the Agency has been conducting more
in-depth assessments which are expected to be published within the
next six months. These assessments will be based on further
evaluations of emission levels, including the use of results from
recent field measurement studies and computer programs which will
incorporate more recently developed dose and health risk information
and methodology.
Source categories in this report have been divided into four
groups:
- facilities licensed by the Nuclear
Regulatory Commission or States under an
agreement with the Nuclear Regulatory
Commission;
- facilities operated and regulated by the
Department of Energy;
- facilities emitting naturally occurring
radionuclides;
- other facilities emitting minor amounts of
radionuclides.
Summaries of the emissions, dose rates, and risks associated
with representative model facilities of source categories within
these groups and with actual Department of Energy facilities are
shown in tables S-l through S-4. These data should be treated as
preliminary estimates and used carefully with the recognition that
they are highly uncertain.
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CHAPTER 1
INTRODUCTION
1.1 Content of Report
This report presents a preliminary analysis of the public
health impact caused by emissions of radionucTides into air from
major source categories in the United States. For each source
category, the following information is presented:
(1) A general description of the source category
and its size.
(2) A brief description of the processes or
activities which lead to the emission of
radioactive materials into the air.
(3) A summary of emissions data for the source
category.
(4) Estimates of the radiation doses and health
risks to both individuals and population
groups.
1.2 Source Categories
Radionuclides are emitted into the air from numerous sources
located throughout the United States including nuclear power plants
and other facilities in the nuclear fuel cycle, national defense
facilities, research and development laboratories (including
research reactors and accelerators), medical facilities, industrial
users, certain mining and milling operations, and fossil fuel
combustion plants.
For the purposes of the assessments presented in this report,
these sources have been divided into four major categories: (1)
facilities licensed by the Nuclear Regulatory Commission (NRC) or
States* under an agreement with NRC; (2) facilities operated and
regulated under the direction of the Department of Energy (DOE);
(3) facilities emitting naturally occurring radioactive materials;
and (4) other minor sources of emissions.
^Sources are licensed by the Nuclear Regulatory Commission (NRC)
or States which have entered into an agreement with the NRC whereby
certain regulatory authority is relinquished by the NRC and assumed
by the States pursuant to Section 274 of the Atomic Energy Act of
1954, as amended.
-------
1-2
DOE facilities were analyzed as single sources on a
site-by-site basis. The NRC and State licensed facilities, DOD
facilities, and the facilities emitting naturally occurring
radionuclides were grouped together into source categories on the
basis of similarity of activities or operations and analyzed on a
generic basis.
1.3 Emissions Data
An attempt was made to gather and document available
information on the radioactive emissions from the various activities
which use radionuclides and to use these data to estimate the public
health impact resulting from these emissions.
The availability of emissions data varied widely from very
detailed data for some source categories to little or no information
for other source categories. Where possible the analyses presented
are based on the use of reported emissions data obtained by
measurement. However, in the absence of measurement data, best
estimates based on calculated or extrapolated values were used. The
present status of information on emissions for the various source
categories and the origin of the data used in the health impact
analyses are discussed under the subsections entitled "Emissions of
Radionuclides."
1.4 Health Impact Assessment
The data on radioactive emissions has been used to estimate
the public health impact of these emissions. These assessments
include estimates of the following radiation exposures and health
risks:
(1) Dose-equivalent rates and working level
exposures to the most exposed individuals
(maximum individual);
(2) Collective dose-equivalent rates and working
level exposures to population groups;
(3) Lifetime risks to the maximum and average
individuals in the exposed population;
(4) The number of fatal cancers committed in the
exposed population per year of facility
operation.
The health risks estimated in this report are for fatal
cancers only. Our current practice is to assume that for whole body
-------
1-3
exposure, the number of genetic health effects and the number of
nonfatal cancers are each about the same as the number of fatal
cancers (EPA77).
Assessment Methodology
The dose assessments for the DOE facilities sites were
carried out on a site-by-site basis using data directly from DOE
reports (Chapters 3). DOD facilities were assessed using site
specific and generic assessments from DOD reports (Chapter 5).
The dose assessments for the NRC and State licensed
facilities, and for sources emitting naturally occurring radioactive
materials (Chapters 2 and 4) were carried out on a generic basis
using model facilities located on generic sites. These generic dose
assessments were carried out using the AIRDOS-II (Mo77) computer
code with some minor modifications. This methodology is described
in detail in Appendix A. Cancer risks have been estimated using the
risk per rem and risk per working level-year* level conversion
factors shown in table B-l of Appendix B. The risk/rem conversion
factors were developed from information in the BEIR report
(BEIR72). The risk/WL-y conversion factors were developed primarily
from information on uranium miner exposures (EPA79 Section 4.0).
Dose and health risk methodology has evolved substantially in
recent years. Some of these changes have not yet been incorporated
in the methodology used in carrying out the assessments presented in
this report. However, because of the preliminary nature of this
report and the limited time available for its preparation, the use
of the existing (off the shelf) methodology and DOE and DOD data
were believed to be statisfactory for this effort.
Future health impact assessments of radioactive emissions
under the Clean Air Act will use more recently developed
methodology. The Oak Ridge National Laboratory (ORNL) under
contract to EPA is developing computer programs incorporating this
revised methodology. The existing calculational models are being
revised based on the latest information on transport, uptake and
metabolic behavior of the various radionuclides. In particular,
*A person exposed to one working level continuously for one year
is considered to have received an integrated exposure equal to one
person-working level year.
-------
1-4
these programs will use ORNL's INREM-II and S factor codes. The
cancer risk methodology will use "life-table" data which considers
age-specific characteristics of somatic health effects. The
life-table analysis will consider relative and or absolute risk for
each radionuclide and latency periods and risks plateaus for each
type of health effect where possible. Information from the BEIR-III
report (BEIR79) will be incorporated into this methodology following
publication of this report.
Definition of Terms Used in Health Impact Assessment
Maximum Individual
Dose equivalent rates and working level exposures are
presented for what is titled the "maximum individual." For DOE and
DOD sites the maximum individual represents either a hypothetical
individual at the site boundary or the nearest actual individual or
group of individuals. For NRC and State licensed facilities and
sources of natural radioactive materials, the maximum individual
represents those individuals living closest to the source of
emissions. For the generic assessments, these individuals were
considered to be located approximately 500 meters from the point of
release in the predominant wind direction. For area sources this
location was nominally 500 meters from the edge of the source. In
the cases of elevated releases where the location of maximum
exposure was at a distance beyond 500 meters, the dose to the
maximum individual was calculated at the location of maximum
exposure.
The dose rates presented for the maximum individual are 50-
year committed dose equivalents. This is the dose equivalent that
will be accumulated over a 50-year period following an intake.
The working level exposures presented for the maximum
individual are the radon-222 decay product levels to which an
individual would be exposed assuming 70 percent equilibrium (i.e.,
100 p Ci/L radon-222 =0.7 WL).
Average Individual
Dose rates and working level exposures for an average
individual within 80 km of a source were obtained by dividing the
collective dose rates and working level exposures for the region by
the population of the region.
-------
1-5
Population
The term population refers to the population living within a
radius of 80 kilometers of a source unless otherwise noted in the
text. For a few source categories, exposures are presented for the
population of the United States or the World and these cases are
specifically identified in the appropriate tables.
Collective dose equivalent rates and working level exposures
are expressed in units of person-rem/year and person-working levels
and are the sum of the dose equivalent rates or working level
exposures to all the individuals in the exposed population due to
the releases from a source. Further details of these calculations
are contained in Appendix A.
Individual Lifetime Risks and Number of Fatal Cancers
The individual lifetime risks are the fatal cancer risks to
individuals which would result from a lifetime of exposure (70
years) to the doses and working levels estimated for those
individuals. The lifetime risk to the maximum individual was
obtained by multiplying the dose equivalent rates and working level
exposures by 70 to obtain the lifetime exposure and then multiplying
this value by risk/rem or risk/WL-year factors as described Appen-
dix B,
-------
1-6
REFERENCES
BEIR72 Advisory Committee on the Biological Effects of Ionizing Ra-
diation, 1972, The Effects of Population Exposures to Low Levels
of Ionizing Radiation, National Academy of Sciences, Washington,
D.C.
EPA77 Environmental Protection Agency, 1977, Radiological Quality
of the Environment in the United States, EPA 520/1-77-009, Office
of Radiation Programs, Washington, D.C.
EPA79 Environmental Protection Agency, 1979, Indoor Radiation Ex-
posure Due to Radium-226 in Florida Phosphate Lands,
EPA-520/4-78-0013, Office of Radiation Programs, Washington, D.C.
Mo77 Moore R. E., 1977, The AIRDOS-II Computer Code for Estimating
Radiation Dose to Man from Airborne Radionuclides in Areas Sur-
rounding Nuclear Facilities, ORNL-5245, Environmental Sciences
Division Publication No. 974, Oak Ridge National Laboratory,
Oak Ridge, Tennessee 37830.
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CHAPTER 2
SOURCES OF EMISSIONS FROM FACILITIES LICENSED BY THE
NUCLEAR REGULATORY COMMISSION
2.1. Uranium Fuel Cycle
The Environmental Radiation Protection Standards for Nuclear
Power Operations published by EPA defined the uranium fuel cyle as
"... the operations of milling of uranium ore, chemical conversion
of uranium, isotopic enrichment of uranium, fabrication of uranium
fuel, generation of electricity by a light-water-cooled nuclear
power plant using uranium fuel, and reprocessing of spent uranium
fuel, to the extent that these directly support the production of
electric power for public use utilizing nuclear energy, but excludes
mining operations, operations at waste disposal sites,
transportation of any radioactive material in support of these
operations, and the reuse of recovered non-uranium special nuclear
and by-product materials from the cycle" (EPA77). These standards
(40 CFR 190) pertain to EPA responsibilities for the protection of
the environment.
The uranium fuel cycle facilities which are operated as a
part of the commercial nuclear fuel cycle are regulated by the
Nuclear Regulatory Commission except for uranium mills located in
the Agreement States of New Mexico, Colorado, Washington, and
Texas. Enrichment facilities are operated and controlled by the
Department of Energy.
Uranium mines and mills are discussed in sections 4.1 and 4.2
of this report. There are no commercial reprocessing plants
operating in the United States today.
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2.1-2
REFERENCES
EPA77 Environmental Protection Agency, 1977, Radiation Protection
Programs, Environmental Radiation Protection Standards for Nuclear
Power Operations, Federal Register, Vol. 42, No. 9, Thursday,
January 13, 1977.
40CFR190 Code of Federal Regulations, Part 40, Protection of the
Environment, Government Printing Office, Washington, D.C.
-------
2.1A-1
2.1A Uranium Conversion Facilities
2.1A.I General Description
The conversion facility purifies and converts uranium oxide
(U^Og) to uranium hexafluoride (UF5), the chemical form in
which uranium enters the enrichmentjjlants. There are two existing
commercial UF5 production facilities' (Kerr-McGee Corporation at
Sequoyah, Oklahoma, and Allied Chemical Corporation at Metropolis,
Illinois) with a combined capacity of about 17,000 MT of uranium per
year in the form of UF6 (table 2.1A-1). NRC (NRC76) has estimated
that five new facilities will be needed by the year 2000 to meet
fuel cycle requirements if neither uranium nor plutonium is recycled.
2.1A.2 Process Description
Two industrial processes are used for uranium hexafluoride
production (EPA73), the dry hydrofluor method and the solvent
extraction method. Each method is used to produce roughly equal
quantities of uranium hexafluoride feed for enrichment; however, the
radioactive effluents from the two processes differ substantially.
The hydrofluor method releases radioactivity primarily in the
gaseous and solid states, while the solvent extraction method
releases most of its radioactive wastes dissolved in liquid
effluents.
Dry Hydroflour Process
The hydrofluor process consists of reduction,
hydrofluorination and fluorination of the ore concentrates to
produce crude uranium hexafluoride, followed by fractional
distillation to obtain a pure product. The dry hydrofluor process
separates impurities either as volatile compounds or as solid
constituents of ash.
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 separates
impurities by extracting the uranium into organic solvent leaving
the impurities dissolved in an aqueous solution.
-------
2.1A-2
Table 2.1A-1. Location and capacity of uranium conversion
facilities in the United States (AEC74)
Capacity
State and Company Location uranium/year
(metric tons)
Illinois
Allied Chemical Corp. Metropolis 12,600
Oklahoma
Kerr-McGee Nuclear Corp. Sequoyah 4,500
2.1A.3 Emissions of Radionuclides
Because no irradiated material is handled by conversion
facilities, all radionuclides present also occur in nature. These
nuclides are radium, thorium, uranium, and their respective decay
products. Uranium is the major source of radioactivity in the
gaseous effluents. Possible chemical species of uranium effluents
include U30s, UO?, UF4, UFe, (NH^U^O/, and Wz^Z- In tf?e wet
solvent extraction method, uranium is present as uranyl nitrate
which may also appear in gaseous effluents. Thus, the uranium may
be released as both soluble and insoluble aerosols. The discharge
to the environment is through low stacks and vents.
2.1A.4 Model Facility
In order to estimate population and individual radiation
doses, a model facility (table 2.1A-2) was developed. The ORNL Case
2 model UFs plant with low impurity feed (Se77) was chosen for
this analysis. Case 2 control technology represents the practical
limits of technology which are readily available today.
The model facility employs the dry hydrofluor process which
releases radioactivity primarily in the gaseous and solid state
form. The choice of one model facility rather than two simplifies
-------
2.1A-3
Table 2.1A-2. Model uranium conversion facility
Parameter Value
Type: Fluorination-fractionation
(dry hydrofluor) UFg plant
Ore grade: Low impurity plant feed
containing 2800 pCi of
thorium-230 and 200 pCi of
radium-226 per gram of
natural uranium.
Annual capacity: 10,000 metric tons of uranium
Emission control: Primary treatment, secondary
bag filters on dust control
streams and secondary or
tertiary scrubbers on
process off-gas streams
Stack:
Height 10 meters
Plume rise 0.0
the analysis, and the selection of the dry hydrofluor process
results in an analysis of the facility process which releases the
largest amount of radioactive material to the atmosphere.
Model Facility Emissions
The atmospheric emissions of radionuclides from the model
conversion facility are presented in table 2.1A-3. The source terms
were taken from the ORNL Case 2 model UFg plant with low impurity
feed (Se77) and were based on operating data where these data were
available. Where data were not available, assumptions were chosen
which tend to make the source terms slightly high. About two-thirds
of the total airborne losses occur via the untreated building
ventilation effluent. Conservative assumptions which tended to
maximize radon releases were used in estimating radon source terms.
For example, a radon emanation coefficient of one was assumed for
yellow cake (Se77).
-------
2.1A-4
Table 2.1A-3. Atmospheric emissions of radionuclides
from a model uranium conversion facility
Radionuclide
Emissions
(Ci/y)
Uranium-238
Uranium-234
Uranium-235
Thorium-234
Protactinium-234m
Thorium-230
Radium-226
Radon-222
8.3E-2
8.3E-2
2.0E-3
8.2E-2
8.2E-2
9.3E-4
6.7E-5
9.2
2.1A.5 Healtji Impact Assessment of a Model Uranium Conversion
Facility
Annual radiation doses and working level exposures resulting
from atmospheric radioactive emissions from a model uranium
conversion facility are presented in tables 2.1A-4 and 2.1A-5.
These estimates are for a site located in the suburbs of a large
Midwestern city (Site B, Appendix A).
The maximum individual dose equivalent rate occurred 503
meters downwind. The uranium was assumed to be released as
insoluble particulates of uranium oxide and fluoride compounds. The
lung dose equivalent rate to the maximum exposed individual is 88
mrem/yr.
The maximum individual dose equivalent rate for an actual
facility could be lower than that calculated for the model
facility. Consideration is given to additional factors which could
result in a dose reduction, e.g., higher stack height, fraction of
released uranium which is soluble, smaller amounts of home grown
food by the maximum individual, maximum individual located farther
than 503 meters downwind, etc. Further airborne dose reductions
using additional control technology are possible. Two-thirds of the
-------
2.1A-5
airborne releases from the model facility are from the building
ventilation effluent. This effluent could be treated, but this
would be expensive because of the large volume of air that must be
handled.
Individual fatal cancer risks and committed population fatal
cancers are presented in table 2.1A-6. The lifetime fatal cancer
risk to the maximum exposed group of individuals is estimated to be
approximately 2.8E-4. The majority of the cancers are fatal lung
cancers. The lifetime fatal cancer risk of the average individual
living within 80 kilometers of the model facility is estimated to be
2.1E-7. The number of .fatal cancers per year of facility operation
is estimated to be 7.3E-3 to the population living in the region
around the facility.
Table 2.1A-4. Annual radiation doses due to atmospheric
radioactive emissions from a model uranium conversion facility
Organ
Lung
Bone
K i dney
Liver
Thyroid
G.I. tract
Other soft
tissue
Maximum
individual
(mrem/y)
8.8E+1
8.5
2.1
7.6E-1
9.2E-1
1.2
9.4E-1
Average
individual
(mrem/y)
5.0E-2
2.1E-3
5.8E-4
3.0E-4
4.1E-4
2.6E-4
4.2E-4
Population
(person-rem/y)
1.2E+2
5.1
1.5
7.2E-1
1.0
6.5E-1
1.1
-------
2.1A-6
Table 2.1A-5. Working level exposures from radon-222
emissions from a model uranium conversion facility
Maximum Regional
Source individual population
(WL) (person-WL)
Model facility 9.8E-6 9.7E-2
Table 2.1A-6. Individual lifetime risks and number of fatal
cancers due to radioactive airborne emissions from a
model uranium conversion facility
Individual lifetime risks Expected fatal cancers
Source Maximum Average per year of operation
individual individual (Fatal cancers)
Particulates
Radon-222
2.7E-4
1.5E-5
1.5E-7
5.8E-8
5.3E-3
2.0E-3
Total 2.8E-4 2.1E-7 7.3E-3
-------
2.1A-7
REFERENCES
AEC74 U.S. Atomic Energy Commission, 1974, Environmental Survey
of the Uranium Fuel Cycle, WASH-1248 (Directorate of Licensing,
Fuels and Materials, U.S. Atomic Energy Commission, Washington,
DC).
EPA73 Office of Radiation Programs, 1973, Environmental Analysis
of the Uranium Fuel Cycle - Part I - Fuel Supply,
EPA-520/9-73-003-C (U.S. Environmental Protection Agency, Office
of Radiation Programs, Washington, DC).
EPA77 U.S. Environmental Protection Agency, 1977, Subchapter
F-Radiation Protection Programs, Part 190 - Environmental
Radiation Protection Standards for Nuclear Power Operations
Federal Register, Vol. 42, No. 9, Thursday, January 13, 1977.
NRC76 U.S. Nuclear Regulatory Commission, 1976, Final Generic
Environmental Statement on the Use of Recycle Plutonium in
Mixed Oxide Fuel in Light Water Cooled Reactors, NUREG-0002,
Vol. 3 (National Technical Information Service Springfield, VA).
SE77 Sears M.B., Blanco R.E., Finney B.C., Hill G.S., Moore R.E.,
and Witherspoon J.P. 1977, Correlation of Radioactive
Waste Treatment Costs and the Environmental Impact of Waste
Effluents in the Nuclear Fuel CycleConversion of Yellowcake
to Uranium Hexafluoride, Part I. The Fluorination-
Fractionation Process, ORNL/NUREG/TM-7, Nuclear Regulatory
Commission.
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2.1B-1
2.IB Fuel Fabrication
2.1B.1 General Description
LWR fuels are fabricated from uranium that has been enriched
in its content of uranium-235. The enriching process occurs at a
gaseous diffusion plant where the natural uranium is processed to
increase the uranium-235 content from 0.7 up to 2 to 4 percent by
weight. The uranium hexafluoride product is shipped to LWR fuel
fabrication plants where it is converted to solid uranium dioxide
pellets and inserted into zirconium tubes. The tubes are fabricated
into fuel assemblies which are shipped to nuclear power plants
(Pe75).
The NRC currently licenses 16 separate facilities to conduct
fuel, processing and fabrication operations (table 2.1B-1).
Table 2.1B-1. NRC-licensed uranium fuel processing and
fabrication facilities (NRC79)
Company Location
Atomics International Canoga Park, California
Babcock & Wilcox Lynchburg, Virginia
Babcock & Wilcox Apollo, Pennsylvania
Combustion Engineering Hematite, Missouri
Combustion Engineering Windsor, Connecticut
Exxon Nuclear Richland, Washington
General Atomic San Diego, California
General Electric San Jose, California9
General Electric Wilmington, North Carolina
Kerr-McGee Crescent City, Oklahoma
Nuclear Fuel Services Erwin, Tennessee
Texas Instruments Attleboro, Massachusetts
U.S. Nuclear Oak Ridge, Tennessee
United Nuclear Montville, Connecticut
United Nuclear Wood River, Rhode Island
Westinghouse Electric Columbia, South Carolinab
aFacility now reduced to R & D activities on LWR fuel.
bAlso authorized to receive and store sealed mixed oxide
fuel rods and assemble mixed-oxide rods into fuel bundles.
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2. IB-2
2.1B.2 Process Description
The processing technology used for fuel fabrication can be
divided into three basic operations: chemical conversion of UF5 to
U02; mechanical processing including pellet production and fuel
element fabrication; and recovery of uranium from scrap and
off-specification material. These operations are usually identical in
the various facilities. The exception is the chemical conversion of
UFs to U02 which may be accomplished by the ADU (ammonium
diuranate) process or the DC (direct conversion) process. The ADU
process converts UFfj to (NH^l^Oy which is then calcined to U02
powder. The UF5 which is received from the enrichment facility is
vaporized and transferred to the reaction vessels. The UF:5 is
hydrolyzed with water and neutralized with NfyOH at a pH of 8 to 9
to form a slurry of ADU in an aqueous solution of ammonium fluoride
and ammonium hydroxide. The ADU is recovered in a centrifuge and a
clarifier and is subsequently dried and calcined to form 1102
powder.
The DC process hydrolyzes the UF6 and reduces the uranium
directly to U02. Cylinders of UF§ are placed in steam-heated
cabinets to vaporize the contained UF5. The UF5 gas enters into
a bed of U02F2 particles which is fluidized by steam. The gas
reacts with the steam on the hot, wet surface of the particles to
form a coating of U02F2. The reaction is:
UF6 + 2H20 -» U02F2 + 4HF.
The particles of U02F2 overflow to a product hopper at which
point the particles are approximately 120 ym in diameter. After a
given amount is accumulated, the batch is transferred to the next
vessel .where the bed is fluidized by steam and ammonia. A second
reaction follows:
U02F2+H2 * U02+HF.
A high percentage of the U02F2 is converted to U02 in the
second reactor, but the product goes into a third reactor where, by
the same process, the reaction is carried to completion.
The gaseous effluent from each of the three converter vessels
passes through a sintered nickel filter in the top of each vessel
before going to the gaseous effluent treatment system where HF and
particulates are removed from the off-gas stream. The most
significant potential environmental impact results from the release
of UFs and U02F2 during the UF2 to U02 conversion and
during chemical operations in scrap recovery.
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2.1B-3
Control Technology
Airborne emission control technology differs for ADU and DC
type facilities. Treated air streams for both types consist of
process and ventilation components. Control technology presented
here is taken from Pechin as represented by Case 2 ADU and DC
facilities which were used in determining model facility source
terms (Pe75).
In the ADU facility, process gas passes through liquid
(water) scrubbers (90 percent removal of entrained solids) and HEPA
filters (95 percent efficient) before release to the atmosphere.
Ventilation off-gases go through roughing filters and HEPA filters
(95 percent efficient) before release to the atmosphere.
In the DC facility, process gas passes through sintered metal
filters (nickel) with trapped solids returned to process; off-gases
continue to KOH scrubbers (for HF removal), then are diluted (for H
removal) and finally released to the atmosphere. Ventilation
off-gases pass through roughing filters, HEPA filters (95 percent
efficient) and are released.
2.IB.3 Emissions of Radionuclides
Some reported uranium effluents for specific fuel fabrication
plants are given in table 2.1B-2. The data serve to indicate order-
of-magnitude values, but many reports do not relate the effluent
data to the quantities processed.
2.IB.4 Model Fuel Fabrication Facility
A model facility (table 2.1B-3) has been developed for
assessing the impact of uranium fuel fabrication plants. Facility
production values were based on 1500 MTU/y as reported by Pechin
(PE75). The maximum exposed individual was assumed to be located
500 meters downwind.
Other operating parameters (such as control technology) were
taken from the ORNL model (PE75) and were assumed to be the same for
both ADU and DC type facilities.
Estimated radioactive emissions for the model facilities are
shown in table 2.1B-4.
-------
2.1B-4
Table 2.1B-2. Atmospheric emissions of radionuclides from
currently licensed uranium fuel fabrication facilities
Plant
Emissions
(Ci/y)
Exxon (JNC70)
Kerr-McGee (KMC71)
General Electric (Ly78, GEC71)
Westinghouse (WEC72)
Nuclear Fuel Services (NRC78)a
Babcock & Wilcox (NRC78)a
(Lynchburg, Virginia)
0.00015
0.04
0.0028
0.2
0.01
0.0007
aEffluent data reported for January 1 to June 30, 1977, were
doubled to estimate an annual release rate.
Table 2.1B-3. Model uranium fuel fabrication facility
Parameter
Value
Type of facility:
Ammonium diuranate (ADU)
Direct conversion (DC)
Annual capacity
Stack
feed to plant hydrolyzed
in water, uranium precipitated
in ammonia to form ADU. ADU
calcined to form U02.
UF6 feed to plant reacted with
water vapor and hydrogen to form
U02.
1500 metric tons of uranium
10 meters fixed stack height
with no plume rise
-------
2.1B-5
Table 2.1B-4. Atmospheric emissions of radionuclides
from the model uranium fuel fabrication facility
Radionuclide
Emissions (Ci/y)
ADU
DC
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Thorium-231
Thorium- 234
Protactinium-234
7.6E-3
2.6E-4
3.9E-4
9.7E-4
2.6E-4
9.7E-4
9.7E-4
5.9E-3
2.0E-4
3.0E-4
7.5E-4
2.0E-4
7.5E-4
7.5E-4
ADU Ammonium diuranate.
DC Direct conversion.
2.IB.5 Health Impact Assessment for Model Uranium
Fuel fabrication Faci1ity
Annual radiation doses due to airborne radioactive emissions
from model ADU and DC fuel fabrication facilities are presented in
table 2.1B-5. These estimates are for a site located in the suburbs
of a large Midwestern city (Site B, Appendix A). The maximum
individual dose equivalent rate occurred 500 meters downwind.
Table 2.1B-6 estimates the individual lifetime fatal cancer
risks and committed fatal cancers to the regional population. The
lifetime fatal cancer risk to the highest exposed group of
individuals is approximately 1.5E-5 for the ADU facility and 1.2E-5
for the DC facility. The lifetime fatal cancer risk to the average
individual in the region is estimated to be 8.8E-9 for the ADU
facility and 6.5E-9 for the DC-facility.
The estimated number of fatal cancers per year of site
operation to the regional population is estimated to be 3.0E-4 for
the ADU facility and 2.3E-4 for the DC facility.
-------
2.1B-6
Table 2.1B-5. Annual radiation doses from radioactive emissions
from model ADU and DC process fuel fabrication facilities
Organ
Lung
Bone
Kidney
Liver
Thyro i d
G.I. Tract
Other soft tissue
Maximum
individual
(mrem/y)
4.9
4.6E-1
1.2E-1
5.6E-2
7.8E-2
5.9E-2
7.1E-2
ADU Process
Average
individual
(mrem/y)
2.9E-3
1.1E-4
3.5E-5
2.5E-5
4.0E-5
2.1E-5
3.6E-5
Population
(person-rem/y)
7.0
2.7E-1
8.6E-2
6.3E-2
l.OE-1
5.1E-2
8.9E-2
Organ
Lung
Bone
Kidney
Liver
Thyroid
G.I. Tract
Other soft tissue
Maximum
individual
(mrem/y)
3.9
3.6E-1
9.2E-2
4.3E-2
6.1E-2
4.6E-2
5.6E-2
DC Process
Average
individual
(mrem/y)
2.2E-3
8.4E-5
2.7E-5
2.0E-5
3.1E-5
1.6E-5
2.8E-5
Population
(person-rem/y)
5.5
2.1E-1
6.7E-2
4.9E-2
7.8E-2
4.0E-2
7.0E-2
ADU Ammonium diuranate.
DC Direct conversion
-------
2.1B-7
Table 2.1B-6. Individual lifetime risks and population
fatal cancers due to radioactive emissions from model ADU and DC
process uranium fuel fabrication plants
Source
Individual lifetime risks
Maximum Average
individual individual
Expected fatal cancers
per year of operation
(Fatal cancers)
ADU process
DC process
1.5E-5
1.2E-5
8.8E-9
6.5E-9
3.0E-4
2.3E-4
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2.1B-8
REFERENCES
AEC74 U.S. Atomic Energy Commission, 1974, Environmental Survey
of the Uranium Fuel Cycle, WASH-1248 (Directorate of Licensing,
Fuels and Materials, U.S. Atomic Energy Commission, Washington,
DC).
EPA73 Office of Radiation Programs, 1973, Environmental Analysis
of the Uranium Fuel Cycle - Part I - Fuel Supply, EPA-520/9-73-
003-C (U.S. Environmental Protection Agency, Office of
Radiation Programs, Washington, DC).
EPA77 U.S. Environmental Protection Agency, 1977, Subchapter
F-Radiation Protection Programs, Part 190 - Environmental
Radiation Protection Standards for Nuclear Power Operations,
Federal Register, Vol. 42, No. 9, Thursday, January 13, 1977.
GEC71 General Electric Company, 1971, letter dated November 29,
1971, USAEC Docket No. 70-1113.
JNC70 Jersey Nuclear Company, 1970, Applicant's Environmental
Report, Uranium Oxide Fuel Plant, No. JN-14, USAEC Docket No.
70-1257.
KMc71 Kerr-McGee Corporation, 1971, letter dated October 11, 1971,
USAEC Docket No. 70-1113.
Ly78 Lyon R. J., Shearin R. L., and Broadway J. A., 1978, A
Radiological Environs Study at a Fuel Fabrication Facility, EPA-
520/5-77-004 (U.S. Environmental Protection Agency, Office
of Radiation Programs, Washington, DC).
NRC78 U.S. Nuclear Regulatory Commission, 1978, letter from
H. T. Peterson, NRC, to T. W. Fowler, EPA with enclosure of
special nuclear material license effluent reports.
NRC79 U.S. Nuclear Regulatory Commission, 1979, Program Summary
Report, NUREG 0380, Vol. 3, No. 3, March 16, 1979.
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2.1C-1
2.1C Light-Water Reactor Nuclear Power Plants
2.1C.1 General Description
As of June 30, 1978, there were 70 operable nuclear power
reactors in the United States; 91 were being built and 42 were
planned (DOE78). With only two exceptions (an operable high
temperature gas-cooled reactor and a planned breeder reactor), all
of the 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 and it is assumed this two-to-one PWR-BWR ratio
will continue through the year 2000.
Projections of electrical generation by nuclear energy in the
United States are difficult to estimate for a number of reasons--
uncertainty regarding demands for electricity, the availability of
alternative fuels, and national policy regarding nuclear power.
However, projections for the United States by the Department of
Energy estimated an installed nuclear capacity in the year 2000 to
range from 255 GW(e) (low case) to 395 GW(e) (high case)(C178).
2.1C.2 Process Description
A light-water-cooled nuclear power station operates on the
same principle as a conventional fossil-fueled (oil or coal) power
station except that the heat generation is provided by nuclear
fission rather than combustion. The heat liberated in either
process is used to convert water into steam. The steam enters a
turbine which is connected to a generator that produces alternating
electric current. The fission process in the nuclear power plant
produces radioactive gases in the fuel (fission products) and also
in the coolant through the absorption of neutrons by coolant and
structural materials (activation products). A typical power reactor
is expected to experience a certain amount of fuel failures or
defects over its operating life through which radioactive gases may
escape the fuel rod and enter the coolant. As a result, radioactive
fission gases are present to some extent in the coolant of the
reactor at all times.
The major gaseous radionuclide emissions for the BWR result
from noncondensable off-gas removed from the main turbine condenser
by the steam jet air ejector (SJAE). Gaseous waste from the SJAE
enters a holdup system, to allow for decay of short-lived radio-
nuclides, and is then discharged to the atmosphere. Even after
holdup, the activity discharged from the SJAE is typically the pre-
dominant source of gaseous radioactive waste at a BWR.
-------
2.1C-2
The major source of gaseous radioactivity at a PWR results
from degassing of the coolant and shim bleed operations which are
designed to remove dissolved radioactive and stable gases in the
coolant. These gases then enter a gaseous waste system and are
transferred to a holdup system for decay of short-lived
radionuclides before being released to the environment. At a PWR,
gaseous wastes from other sources (e.g., building ventilation) may
contribute comparable amounts of radioactivity.
2.1C.3 Emissions of Radionuclides
EPA evaluated LWR source terms, control technology and
projected health impacts for different levels of effluent treatment
systems (EPA73, EPA76) in developing the uranium fuel cycle standard
(40 CFR 190). EPA has also summarized effluent data from
light-water-cooled nuclear power plants from 1973-1976 (Ph77).
Tables 2.1C-1 and 2.1C-2 summarize the airborne releases from
single-unit BWRs and PWRs for 1976. Variation in annual releases
may be attributed to plant operating parameters such as operating
time, power levels, control technology, and equipment malfunction.
Table 2.1C-1.
Radioactive emissions from operating single-unit
BWRs, 1976
Facility
Emissions (Ci/y)
Noble gases
Tr i t i urn
Halogens
Dresden 1
Big Rock Point
Humboldt Bay 3
La Crosse
Oyster Creek 1
Nine Mile Point 1
Millstone Point 1
Monticello
Vermont Yankee
Pilgrim 1
Cooper
Duane Arnold
Hatch 1
Fitzpatrick 2
460,000
15,200
93,000
124,000
166,000
176,000
507,000
11,400
2,870
183,000
38,100
5,260
3,110
46,200
61.3
7.8
1.3
12.6
1.1
18.5
28.7
76.9
14.2
37.4
67.1
16.4
1.4
15.4
2.34
.02
.37
.10
46.4
8.6
36.5
1.0
.1
2.0
.1
.1
1.0
5.8
-------
Table 2.1C-2.
2.1C-3
Radioactive emissions from operating
single-unit PWRs, 1976
Facility
Emissions (Ci/y)
Noble gases
Tritium
ND Not detectable.
NR Not reported.
Halogens
Yankee (Rowe)
San Onofre 1
Haddam Neck
Ginna, R.E.
Robinson, H.B.
Palisades
Maine Yankee
Fort Calhoun
Kewaunee
Three Mile Island 1
Arkansas One 1
Rancho Seco
Cook, Donald C.
Millstone Point 2
Tro j an
St. Lucie
Beaver Valley 1
Salem 1
27.1
417
492
5,520
791
29.9
1,300
2,150
1,600
2,760
5,690
127
975
1,550
527
1,790
1.08
8.9E-5
2.02
47.2
739
23.6
158
NR
3.7
2.5
0.7
717
6.7
9.1
.11
15.
1.5
2.1
3,720
1.5E-5
1.3E-4
4.5E-3
7.3E-4
3.4E-2
2.5E-1
3.1E-2
1.6E-3
7.2E-1
3.3E-3
8.6E-3
4.1E-2
1.1E-3
1.4E-3
4.9E-2
3.7E-4
2.0E-3
4.3E-6
ND
2.1C.4 Model Facility
Model facilities were developed for the impact analysis of
BWRs and PWRs (table 2.1C-3). The model BWR and PWR characteristics
(with recirculating u-tube type steam generators) as developed by
the NRC are described in the final generic environmental statement
on the use of recycled plutonium in mixed-oxide fuel for light-water
cooled reactors (NRC76).
-------
2.1C-4
Emissions
Atmospheric emissions of radioactive materials from the NRC
model facilities are presented in table 2.1C-4. These models are
meant to illustrate typical cases and source terms and are not
directly applicable to a particular operating reactor (NRC76).
2.1C.5 Health Impact Assessment of Model BWR and PWR Facilities
Radiological impacts of airborne effluents from light-water-
reactors have been estimated using the AIRDOS-II computer code
(Mo77) and emissions reported in table 2.1C-4. The maximum
individual, average individual, and population dose equivalent rates
are presented in table 2.1C-5. The estimates are for a site located
in the suburbs of a large Midwestern city in the United States (Site
B, Appendix A). Food production and consumption assumptions for the
maximum individual were selected for a rural setting. The maximum
individual dose equivalent rate occurred 503 meters downwind.
Table 2.1C-6 estimates the individual lifetime fatal cancer
risks and committed fatal cancers to the regional population. The
lifetime fatal cancer risk to the highest exposed group of
individuals is approximately 2.0E-5 for the BWR and 8.6E-6 for the
PWR. The lifetime fatal cancer risk to the average individual in
the region is estimated to be 3.8E-8 for the BWR and 2.5E-8 for the
PWR. The number of fatal cancers per year of site operation is
estimated to be 1.3E-3 for the BWR and 8.8E-4 for the PWR to the
regional population.
An assessment of the health impact from emissions of
carbon-14 and tritium must also consider the impact to the
population living beyond 80 kilometers since these radionuclides are
dispersed worldwide. The number of fatal cancers committed to the
world population from carbon-14 and tritium releases is estimated to
be 3.7E-2 for the BWR and 2.2E-2 for the PWR (Fo79). These fatal
cancers estimates are the impact over a 100-year period from one
year of carbon-14 and tritium emissions. Risk factors used in
estimating the fatal cancers committed to the world population from
carbon-14 and tritium are presented in table 2.1C-7. The tritium
risk coefficient was calculated using the UNSC^AR (UN77) estimated
collective dose to the global population of 3 x 10-3 person rad
per curie of tritium released to the atmosphere. A fatal cancer
risk coefficient of 200 fatal cancers per 10° person-rem was
utilized since tritium is uniformly distributed throughout body
tissue. The carbon-14 risk coefficients employed estimates of the
global impact of carbon-14 discharges to the atmosphere using the
diffusion model of the carbon cycle developed by Killough (Ki77).
-------
2.1C-5
The 100-year global collective dose commitment factors for carbon-14
are 28 person-rem per curie to the total body and 10.7 person-rem
per curie to the gonads. For 106 person-rem exposure to the total
body from .carbon-14, the following fatal cancers estimates were
used: 58 leukemia deaths and 88 other cancer deaths. The health
effects estimates are for carbon-14 beta irradiation only. No
attempt was made to estimate the effect of carbon-14 to nitrogen-14
transmutation.
In addition, there is an impact to the world population
from krypton-85; however, the impact is small compared to that of
carbon-14. Ellett (E176) estimates 7.0E-5 committed fatal cancers
for annual releases of krypton-85 from a 1000 MW(e) light water
reactor.
-------
2.1C-6
Table 2.1C-3. Model BWR and PWR light-water-reactors
Parameter Value
Type: Boiling water reactor and
Pressurized water reactor
Capacity: 1,000 MW(e)
Fuel: Uranium only
Stack:
Height 20-meter, fixed stack
height with no plume rise
Table 2.1C-4. Atmospheric emissions of radionuclides
from model BWR and PWR facilities (NRC76)
Radionuclide
Argon-41
Krypton -83m
Krypton-85m
Krypton-85
Krypton-87
Krypton-88
Xenon-131m
Xenon-133m
Xenon-133
Xenon-135m
Xenon-135
Xenon-138
Iodine-131
Iodine-133
Carbon-14
Tritium
Emissions
BWR
25
(a)
150
290
200
240
18
(a)
3,200
740
1,100
1,400
0.3
1.1
bg
43
(Ci/y)
PWR
25
1
16
470
3
23
82
120
12,000
(a)
86
(a)
0.025
0.023
b5
1,100
"Source: Fowler (Fo76).
-------
2.1C-7
Table 2.1C-5. Annual radiation doses from model
BWR and PWR facilities
Organ
Organ
BWR
Maximum Average
individual individual Population
(mrem/y) (mrem/y) (Person-rem/y)
Lung
Bone
Kidney
Liver
Thyroid
G. I. tract
Other soft tissue
1.5
2.4
1.4
1.5
11.0
1.4
1.8
2.9E-3
4.8E-3
2.6E-3
2.7E-3
7.6E-3
2.4E-3
3.3E-3
7.3
12.0
6.5
6.7
19.0
6.0
8.3
PWR
Maximum Average
individual individual Population
(mrem/y) (mrem/y) (Person-rem/y)
Lung
Bone
Kidney
Liver
Thyro i d
G. I. tract
Other soft tissue
6.0E-1
1.2
5.9E-1
6.0E-1
1.4
5.7E-1
7.5E-1
1.9E-3
3.7E-3
1.6E-3
1.6E-3
2.7E-3
1.3E-3
2.1E-3
4.6
9.3
4.1
4.1
6.8
3.3
5.3
-------
2.1C-8
Table 2.1C-6. Individual lifetime risks and population
fatal cancers due to radioactive emissions from model BWR
and PWR facilities
Individual lifetime risks Expected fatal cancers
Source Maximum Average per year of operation
individual individual (Fatal cancers)
BWR facilities 2.0E-5 3.8E-8
PWR facilities 8.6E-6 2.5E-8
1.3E-3
8.8E-4
Expected total cancers in the worldwide
Source population over the next 100 years3
Tritium Carbon-14 Total
(Fatal cancers) (Fatal cancers) (Fatal cancers)
BWR facilities
PWR facilities
2.6E-5
6.1E-4
3.7E-2
2.1E-2
3.7E-2
2.2E-2
Calculated as described in Section 2.1C.5.
Table 2.1C-7. Tritium and carbon-14 risk coefficients for
fatal cancers committed to the world population
Radionuclide
Fatal cancers committed to
the world population per curie
released to the atmosphere
(Fatal cancers/Ci)
Tr i t i urn
Carbon-14 (100 years)
6.0E-7
4.1E-3
-------
2.1C-9
REFERENCES
C178 Clark R.G. and Reynolds A.W., 1978, Uranium Market-Domestic
and Foreign Requirements, presented at the Grand Junction Office
Uranium Industry Seminar, Department of Energy, Washington, D.C.
DOE78 Technical Information Center, Department of Energy, 1978,
Nuclear Reactors Built, Being Built, or Planned in the United
States as of June 30, 1978, TID-8200-R38 (National Technical
Information Service, Springfield, VA).
E176 Ellett W. H. M. and Richardson A. C. B., 1976, Estimates
of the Cancer Risk Due to Nuclear-Electric Power Generation,
Technical Note ORP/CSD-76-2 (U.S. Environmental Protection
Agency, Office of Radiation Programs, Washington, D.C.).
EPA73 Office of Radiation Programs, 1973, Environmental Analysis
of the Uranium Fuel Cycle, Part II, Nuclear Power Reactors,
EPA-520/9-73-003-C (U.S. Environmental Protection Agency, Office
of Radiation Programs, Washington, D.C.).
EPA76 Office of Radiation Programs, 1976, 40 CFR 190 Environmental
Radiation Protection Requirements for Normal Operations of
Activities in the Uranium Fuel Cycle, Final Environmental
Statement, Volume I and II, EPA 520/4-76-016 (U.S. Environmental
Protection Agency, Office of Radiation Programs, Washington,
DC).
Fo76 Fowler T.W., Clark R.L., Gruhlke J.M. and Russell J.L.,
1976, Public Health Considerations of Carbon-14 Discharges from
the Light-Water-Cooled Nuclear Reactor Industry, Technical Note
ORP/TAD-76-3 (U.S. Environmental Protection Agency, Office of
Radiation Programs, Washington, D.C.).
Fo79 Fowler T.W., and Nelson, C.B., 1979, Health Impact Assess
ment of Carbon-14 Emissions from Normal Operations of Fuel
Cycle Facilities, Draft EPA Report, Office of Radiation
Programs, Environmental Protection Agency, Washington, D.C.
Ki77 Killough G.G., 1977, A Diffusion-Type Model of the Global
Carbon Cycle for the Estimation of Dose to the World Popula-
tion from Releases of Carbon-14 to the Atmosphere, ORNL-5269
Oak Ridge National Laboratory (National Technical Information
Service, Springfield, Va.).
-------
2.1C-10
REFERENCEScontinued
Mc78 McBride J. P., Moore R. E., Witherspoon J. P. and
Blanco R. E., 1978, Radiological Impact of Airborne Effluents
of Coal and Nuclear Plants, Science, Vol. 202, 8 December 1978.
Mo77 Moore R. E., 1977, The AIRDOS-II Computer Code for
Estimating Radiation Dose to Man from Airborne Radionuclides
in Areas Surrounding Nuclear Facilities, ORNL-5245 (National
Technical Information Service, Springfield, VA).
NRC76 U.S. Nuclear Regulatory Commission, 1976, Final Generic
Environmental Statement on the Use of Recycle Plutonium in
Mixed Oxide Fuel in Light Water Cooled Reactors, NUREG-0002,
Vol. 3 (National Technical Information Service, Springfield,
VA).
Ph77 Phillips J. W. and Gruhlke J., 1977, Summary of Radio-
activity Released in Effluents from Nuclear Power Plants
From 1973 thru 1976, EPA-520/3-77-012 (U.S. Environmental
Protection Agency, Office of Radiation Programs, Washington,
D.C.).
UN77 United Nations, 1977, Sources and Effects of Ionizing
Radiation, United Nations Scientific Committee on the Effects
of Atomic Radiation, 1977 Report to the General Assembly, with
annexes, United Nations Publication Sales No. E.77.IX.1.
-------
2.2-1
2.2 High Temperature Gas Cooled Reactor (HTGR)
2.2.1 General Description
A high temperature gas cooled power reactor uses uranium-235
as fissile fuel in the initial reactor core, thorium-232 as fertile
material which is converted to uranium-233, graphite as the
moderator, cladding structure and reflector, and helium as a
coolant. The heat generated within the fuel element is transferred
through the fuel cladding to the primary coolant, helium gas. The
helium transfers heat to a secondary water system to produce steam.
The resulting steam is used to turn a turbine generator to produce
electricity.
The only operating commercial nuclear power plant using a gas
cooled reactor system in the United States today is Fort St. Vrain
near Greeley, Colorado, which started up in 1978. The first gas
cooled reactor system in the United States was Peach Bottom 1 in
Pennsyslvania which operated between 1966 and 1974 is now
decommissioned. The proposed station in Delaware, Summit Power
Units 1 and 2, has been cancelled.
2.2.2 Process Description
In operation, the gas cooled reactor system converts
thorium-232 to uranium-233, a fissle material. The helium coolant
flows through the reactor core to the steam generators, where it
gives up heat to convert water to steam, and then returned to the
reactor by helium circulators. The generated steam flows to the
turbine to produce electricity. The reactor, with its accessories
(helium circulators and steam generators), is contained in a
prestressed concrete reactor vessel (PCRV), a vertical hexagonal
prism, 19 by 32 meters, with three-foot thick side walls.
The principal source of high activity gaseous waste
originates from the helium purification system. Small amounts of
potentially contaminated gaseous waste also come from the sampling
of the primary coolant, purging of fuel storage and handling
systems, purging of the helium circulator handling-cask and from the
PCRV support floor vent and liquid waste tank vent headers. Gas
enters the gaseous waste system through either of two paths, the low
or high activity vent headers. The low activity vent header
collects gases of sufficiently low activity and flow rate that they
can be discharged through the reactor building vent after passing
through a prefilter, a high efficiency particulate filter (HEPA),
and a charcoal adsorber.
-------
2.2-2
Table 2.2-1. Estimated atmospheric emissions of radionucTides
from a typical gas cooled reactor system
Radio-
Total
Emissionsa
(Ci/y)
nuclide
Krypton-85
Krypton-87
Krypton-88
Xenon-131m
Xenon-133
Xenon-135
Xenon-138
Reactor
building
leakage
0.05
2.0
11.0
0.04
12.0
2.4
0.01
Regenera-
tion
system
947
4.8
9.0
Air
ejection
0.002
1.2
3.2
0.002
0.6
2.4
0.03
PCRV
leakb
25 .,6
34 ..9
.4
13.
7.1
Total
947
29
49
5
22
16
7
1075
Regeneration SystemBased on minimum holdup time of 60 days.
''PCRV Leak--Prestressed concrete reactor vessel. Numbers were
extrapolated from AEC74 by multiplying values by 0.42.
The high activity header collects gases that are normally too
radioactive to be released after treatment by filtration only.
These gases are routed to the helium purification system which
removes fission products and chemical impurities from the primary
coolant. The system, which uses two alternately operating gas
processing trains, consists of a high-temperature filter-adsorber to
remove particulates and halogens (mostly iodines), a helium cooler,
a dryer, low-temperature adsorber, and a hydrogen getter unit
(titanium sponge) which removes gaseous hydrogen and tritium.
The helium dryer and the low-temperature filter-adsorber are
regenerated by passing hot helium through the unit, which strips the
accumulated gases (including the radioactive ones) from the
adsorber. The gases are collected and analyzed prior to controlled
venting to the reactor building vent. The anticipated annual
releases of radioactive material in the regeneration off-gases shown
in table 2.2-1 were based on a minimum holdup time of 60 days.
-------
2.2-3
Before entering the reactor building vent, the effluent of
both treatment systems passes through HEPA filters and charcoal
adsorbers. If the activity is above a predetermined level, the gas
stream is diverted to the vacuum tank of the high activity system
for further treatment.
The stream in the economizer and superheater sections of the
steam generator is at a higher pressure (about 2500 psia) than the
helium (about 693 psia), so any leakage will be essentially steam
into the primary coolant (helium). However, the steam in the reheat
section of the steam generator is at a lower pressure (about 649
psia) so it is possible for radioactive primary coolant to enter the
steam through leaks in this portion of the steam generator. Activ-
ity in the secondary system can be released via the air ejector.
2.2.3 Emissions of Radionuclides
Table 2.2-1 summarizes atmospheric emissions from the model
plant for four release pathways as reported in final environmental
impact statements for Fort St. Vrain (AEC72) and the proposed Summit
Power Station (AEC74). The PCRV leakage source for Summit was
scaled by power level to apply to the model plant. Emissions from
the regeneration systems were based on a minimum holdup time of 60
days. There are no operating data available yet from Fort St. Vrain
which began an operational testing period during 1978.
2.2.4 Typical Facility
Since there is only one gas cooled power reactor operating in
the United States, the characteristics of that facility have been
used to describe the typical plant in table 2.2-2 (AEC72).
2.2.5 Health Impact Assessment of Model Facility
The final environmental impact statement (AEC72) for Fort St.
Vrain estimates an annual total body dose of 6.4 person-rems (table
2.2-3) from atmospheric emissions of radioactive materials to the
population within a 50-mile radius of the site. Estimates are for a
low population density site in Western United States. The
population within 80 km is about 1.4 million. This dose, however,
does not include the small contribution from PCRV leakage. The risk
from this exposure is estimated to be .0013 fatal cancers per year.
The majority of this exposure is from krypton-85 emissions from the
regeneration system. Doses and risks are presented in table 2.2-3
and table 2.2-4.
-------
2.2-4
Table 2.2-2. Typical high temperature gas cooled reactor system
Parameter
Value
Thermal power (MWth)
Electrical power (MWe)
Plant lifetime
Fuel cycle
Fuel temperature (op)
Enrichment (% Uranium-235)
Fuel element material
Coolant pressure (psia)
Coolant loops
Coolant cleanup system:
Tr i t i urn
Noble gases
Containment building type
Radwaste system (gaseous)
Type
Retention days
842
332
30 years
U/Th
2300
93
Graphite
700
2
Titanium sponge
Low temperature adsorber
Confinement only
Pressurized decay tank
60
Table 2.2-3. Annual radiation doses due to atmospheric
emissions from Fort St. Vrain (AEC72)
Source
Nearest
resident
(mrem/y)
Average
individual
(mrem/y)
Population
(person-rem/y)
Fort St. Vrain
0.52
4.5E-3
6.4
-------
2.2-5
Table 2.2-4. Individual lifetime risks and numbers of fatal
cancers due to radioactive emissions from Fort St. Vrain
Individual lifetime risks Expected fatal cancers
Nearby Average individual per year of operation
Source residents in region (Fatal cancers)
Fort St. Vrain 5.5E-6 6.4E-8 1.3E-3
-------
2.2-6
REFERENCFS
AEC72 Atomic Energy Commission, 1972, Final Environmental Statement
related to operation of the Fort St. Vrain Nuclear Generating
Station of Public Service Company of Colorado, Docket Mo. 50-267,
U.S. Atomic Energy Commission, August 1972.
AEC74 Atomic Energy Commission, 1974, Final Environmental Statement
related to the proposed Summit Power Station Units 1 and 2,
Delmarva Power and Light Company, Docket Nos. 40-450 and 50-450
and 50-451, United States Atomic Energy Commission, July 1974.
-------
2.3-1
2.3 RadiopharmaceutJcal Industry
2.3.1 General Description
The radiopharmaceutical industry converts radioactive
chemicals into a form suitable for use in medicine and research.
The radiopharmaceutical industry is composed of producers and users
of radiopharmaceuticals. Radioisotopes used in medical practice
are produced directly in nuclear reactors, particle accelerators,
and in isotope generators by parent-daughter separation. Producers
include those companies which produce isotopes as raw materials for
other pharmaceutical companies as well as the secondary companies
which produce radiopharmaceuticals through tagging and labelling
processes.
Radioactive materials are widely used for medical diagnosis
and therapy. Since 1946, the number of medical institutions in the
United States licensed to use radioactive materials derived from
nuclear chain reactions has grown from 38 to more than 12,000,
including both NRC and Agreement State licensees. These licensees
perform an estimated 30 million nuclear medicine procedures per
year at an estimated cost of $1.6 billion (NRC77).
Producers
Table 2.3-1 lists the principal producers of
radiopharmaceuticals in the United States (AEC74). A more current
list should be available in late 1979 when an EPA contract
requesting this information is completed.
Approximately 12 commercial organizations are currently
providing either cyclotron- or reactor-produced radioisotopes
(AEC71). The Federal government is gradually relinquishing to
industry its role as prime supplier of reactor-produced
radionuclides.
Private industry now produces 65 different, generally used
radioisotopes, as compared to about 50 low-volume, special-use
items still produced in Federal installations. For practical
considerations, therefore, the Federal government is not being
considered as a significant supplier of medical radioisotopes. The
principal industrial suppliers of cyclotron- and reactor-produced
radioisotopes are listed in table 2.3-2.
-------
2.3-2
Table 2.3-1. Principal industrial processors of organic labeled
compounds, radiochemicals and radiopharmaceuticals (AEC74)
Processor
Radio-
immuno- Radio-
Radio- assay pharma-
chemicals reagents^ ceuticals
Abbott Laboratories
North Chicago, IL
Aerotest Operations,
San Ramon, CA X
American Radiochemical Corp
Sanford, FL X
Amersham/Searle,
Arlington Heights, IL X
Ames Company, Division of
Miles Laboratories, Inc.,
Elkhart, IN X
Bio-Chemical & Nuclear Corp.,
Burbank, CA X
Bio-Rad Laboratories
Richmond, CA X
Calatomic, Inc.
Los Alamos, CA X
California Radiochemicals, Inc.,
Los Angeles, CA X
Cambridge Nuclear Corp.,
Billerica, MA, and Princeton, NJ
(Subsidiary of NL Industries,
Inc.) X
Curtis Nuclear Corp.,
Los Angeles, CA X
Dhom Products, Ltd.,
North Hollywood, CA X
X
X
footnote at end of table.
-------
2.3-3
Table 2.3-1. Principal industrial processors of organic labeled
compounds, radiochemicals and radiopharmaceuticalscontinued
Processor
Racho-
immuno- Radio-
Radio- assay pharma-
chemicals reagentsa ceuticals
Diagnostic Isotopes, Inc.
Upper Saddle River, NJ
Virgo Reagents, Electro-Nucle-
onics Lab, Bethesda, MD
General Electric Company,
Pleasanton, CA
Gamma Industries, (Div. of Nuclear
Systems, Inc.) Houston, TX, and
Baton Rouge, LA
Imaj International, Inc., (Nuclear
Medicine Div of Allergan Phar-
maceuticals) Irvine, CA
Industrial Nuclear Company, Inc.,
Overland, MO
International Chemical & Nuclear
Corp., Irvine, CA
Isolab, Inc., Akron, OH
Iso-Med, Inc., Hawthorne, CA,
(Div of New England Nuclear)
Mallinckrodt Chemical Works,
St. Louis, MO
Medi-Physics, Emeryville, CA,
and South Plainfield, NJ
Miles Laboratories, Inc.,
Elkhart, IN
New England Nuclear
North Bill erica, MA
X
X
X
X
X
X
X
X
aSee footnote at end of table.
-------
2.3-4
Table 2.3-1. Principal industrial processors of organic labeled
compounds, radiochemicals and radiopharmaceuticalscontinued
Processor
Radio-
immuno- Radio-
Radio- assay pharma-
chemicals reagents^ ceuticals
Nuclear Associates, Inc.
Westburg, NY X
Nuclear Dynamics, Elmonte, CA X
Nuclear Equipment Chemical
Corp., Farmingdale, NY X
Nuclear Medical Labs, Inc.,
Dallas, TX X
Schartz/Mann, Div. of Becton,
Dickinson & Co., Orangeburg, NY X
E. R. Squibb & Sons,
New Brunswick, NJ
Teledyne Isotopes,
Palo Alto, CA X
Union Carbide Corp.,
Tuxedo, NY X
Worthington Biochemical Corp.,
Freehold, NJ
X
X
aRadioimmunoassay reagents are organic labeled compounds which are
used in isotope dilution techniques for the sensitive measurement
of biological componentsfor example, the measurement of hormone
levels in blood plasma.
-------
2.3-5
Users
Radiopharmaceutical users are defined as hospitals or
private physicians using radionuclides for medical purposes
including the sewage treatment plants which receive and treat the
liquid wastes.
Diagnostic and therapeutic procedures using radionuclides
are listed in table 2.3-3. Data from a Public Health Service
survey indicate that one in every four patients admitted to
hospitals is given a radioactive tracer as part of his medical
diagnostic workup (BRH70). The rapid growth of nuclear medicine is
partly due to the increasing availability of radiopharmaceuticals
in more easily administered forms.
Diagnostic nuclear medicine includes such techniques as
measuring the uptake of radioactive drugs by individual organs (for
such purposes as assessing thyroid function), "imaging" the
distribution of radioactive drugs among organs or within an organ
(to detect the presence of tumors, for example), estimating the
size of certain body pools (such as red blood cell and blood plasma
volumes), and measuring the components in biological samples (such
as protein binding sites and hormones in blood and urine) (NRC77).
Therapeutic techniques include the use of radioactive drugs
internally (for example, in the treatment of thyroid cancers), the
use of radioactive devices both as implants and on the surface of
the body (termed "brachytherapy," or "therapy from a short
distance") and the use of radioactive devices external to the body
(termed "teletherapy," or "therapy from a distance")(NRC77).
Radionuclides are also used in research studies to obtain
basic medical data and to develop a clinical diagnostic or
therapeutic procedure. There is no current information available
on research studies using medical isotopes. A 1966 BRH survey
reported the most frequently conducted research studies at that
time. These studies (table 2.3-4) accounted for 57 percent of all
research studies reported (BRH70).
2.3.2 Process Description
Radioisotopes used by the radiopharmaceutical industries are
produced by accelerators, by small research reactors, and by
radioisotope generators.
-------
2.3-6
Table 2.3-2. Principal industrial suppliers of cyclotron- and
reactor-produced radioisotopes
Principal Industrial Suppliers of Cyclotron Radioisotopes
Amersham-Searle, Arlington Heights, Illinois
Cambridge Nuclear Corporation, Billerica, Massachusetts
Diagnostic Isotopes, Upper Saddle River, New Jersey
International Chemical and Nuclear Corporation, Irvine, California
Mallinckrodt Chemical Works, St. Louis, Missouri
Medi-Physics Inc., Emeryville, California
New England Nuclear Corporation, Billerica, Massachusetts
Commercially Available Sources of Reactor-Produced Radioisotopes
The following facilities offer irradiation services for isotope
production or other irradiation testing:
Northrup Corporate Labs Union Carbide Corp., UCNR
TRIGA Mark-F Sterling Forest, New York
Hawthorne, California
University of Michigan University of Missouri
Ann Arbor, Michigan Columbia, Missouri
Western New York Nuclear
Research Center, Inc.
Pulstar
Buffalo, New York
-------
2.3-7 ,.
Table 2.3-3. Major radiopharmaceuticals and
their uses (FDA76, NRC79)
Radionuclide Use
Phosphorus-32 Bone marrow therapy
Gallium-67 Tumor localization
Rubidium-81 Myocardial imaging
Technetium-99m Bone imaging, brain imaging, liver
imaging, lung perfusion, myocardial
imaging, blood pool, renograms,
thyroid imaging, thyroid uptake
renal imaging
Iodine-123 Thyroid imaging
Thyroid uptake
Iodine-125 Renograms
Iodine-131 Renal imaging, renograms, thyroid
imaging, thyroid uptake, tumor
localization and therapy
Xenon-133 Lung ventilation
Mercury-203 Renograms
Thalium-201 Myocardial imaging
-------
2.3-8
Table 2.3-4. Radioisotopes used in research (BRH70)
Radionuclide
Compound
Study
Iodine-131
Tritium and
carbon-14
Iodine-131
Carbon-14
Tr i t i um
Labeled hormones
Labeled aldosterone
Labeled albumin
Labeled steroids
Labeled cholesterol
Metabolic
Metabolic
Blood volume
Metabol ic
Metabolic
Tr i t i um
Xenon-133
Technetium-99m
Krypton-85
Xenon-133
Krypton-85
Tr i t i um
Iron-59
Iodine-131
Carbon-14
Calcium-47
Indium-113m
Tr i t i um
Iodine-131
Technetium-99m
Labeled aldosterone
Saline solution
Labeled albumin
Gas
Gas
Saline solution
Labeled thymidine
Ferric chloride
Labeled albumin (MAA)
Labeled cortisol
Calcium chloride
Iron complex
Labeled steroids
Labeled human F.S.H.
Labeled albumin
Adrenal secretion
Blood flow
Lung scanning
Brain blood flow
Muscle blood flow
Cardiac output
Autoradiography
Iron absorption
Lung scanning
Plasma clearance
Metabolic
Lung scanning
Metabolic
Radioimriunoassay
Heart scanning
2.3.2 Process Description
Radioisotopes used by the radiopharmaceutical industries are
produced by accelerators, by small research reactors, and by
radioisotope generators.
Reactor Produced
Most radioisotopes are made in nuclear reactors by one of
the reactions shown in table 2.3-5. The most common is the
neutron-gamma reaction because many elements capture neutrons
easily. Such radioisotopes as ^Ha, 5^Fe, 60Co, and
(McG77) are produced by neutron capture.
-------
2.3-9
Table 2.3-5. Nuclear reactions used in radioisotope production
Reaction Examples
(1)
(2)
(3)
Neutron-gamma
Neutron-proton
Neutron-alpha
(n
(
(n
,Y)
n,p)
,a)
59Co + n + 60Co
32S + n -» 32p +
35C1 + n -» 32p
+ Y
P
+ a
The main steps (Ba66) in the production of radionuclides in
a reactor are:
1. A suitable target is prepared and irradiated with
neutrons.
2. The irradiated target is processed by simple
dissolution or by more complicated separations--
including ion exchange, precipitation and
distillationto remove undesirable impurities
or to concentrate the product nuclide.
3. The radionuclides are placed in inventory, dispensed,
and packaged for shipment.
Accelerator Produced
A wide variety of radioisotopes are produced in particle
accelerators, such as the cyclotron. The amount of radioactive
material which can be produced in particle-accelerating machines is
smaller than that made in a nuclear reactor.
The cyclotron is used to produce two principal types of
nuclides. First, those whose decay characteristics are preferable
to other radioisotopes of the same element that are available from
nuclear reactors; and second, isotopes of elements of biological
importance for which no reactor-produced nuclides exist. Examples
of the first category are: iodine-123, iron-52, mercury-199m,
chromium-49, copper-61 and selenium-73; examples of the second are
carbon-11, nitrogen-13, and oxygen-15.
-------
2.3-10
Generator Produced
Because of the difficulties encountered in the preparation
and delivery of radionuclides with short half-lives, there is a
growing trend for hospitals to operate radioisotope generators for
the production of certain isotopes, notably technetium-99m.
These devices make short-lived nuclides available at long
distances from the source of production. They consist of a
longer-lived parent nuclide that produces the short-lived daughter
nuclide as it decays. The daughter nuclide is chemically separated
at intervals and the parent is left to generate a fresh supply of
the daughter (Wa68).
Nuclides that have the potential to be produced by
radioisotope generators are listed in table 2.3-6.
Table 2.3-6.
Potential generator systems from AEC Symposium No. 6
(AEC66)
Daughter
isotope
Arsenic-72
Cesium-131
Indium-113m
Potass ium-42
Praseodymium-144
Rhodium-103m
Scandium-44
Technetium-99m
Tellurium-125m
Tellurium-127
Half-
life
26 hours
9.7 days
1.7 hours
12.4 hours
17.3 minutes
57 minutes
3.9 hours
6.0 hours
58 days
9.3 hours
Parent
isotope
Strontium-72
Barium-131
Tin-113
Argon-42
Cerium- 144
Palladium-103
Titanium-44
Molybdenum-99
Strontium-125
Tellurium-127m
Half-
life
8.4 days
11.6 days
118 days
>3.5 days
285 days
17 days
~103yearS
66.0 hours
2.7 years
105 days
-------
2.3-11
2.3.3 Emissions of Radionuclides
Radiopharmaceutical Production Plants
Three medical isotope producers have submitted current
emission data to EPA. All three firms are large in terms of
production and volume sales, and the three firms together are
representative of the major categories of medical isotope
producers, i.e., radiochemical, radioimmunoassay reagent, or
radiopharmaceutical producer. The emission data represent airborne
releases from normal operations of the facility as measured by
company-owned monitoring systems. The average annual releases are
summarized in table 2.3-7.
Table 2.3-7. Atmospheric emissions of radionuclides
from three major radiopharmaceutical producers
Emissions
Source (Ci/y)
Plant X Plant Y Plant I
Iodine-125
Iodine-131
9.4E-1
8.8E-1
2.6E-3
3.1E-2
2.7E-2
5.7E-1
Hospitals
A study conducted under EPA Contract 68-01-5049 (Te79)
showed that approximately 3,000 hospitals are licensed by the NRC
to administer radioisotopes for diagnostic and/or therapeutic
purposes (Te79). This study indicates that iodine-131 and
xenon-133 are released as airborne emissions due to normal
preparation and administration procedures. Technetium-99m is not
released. Source terms were estimated to be 1E-5 Ci/y for
iodine-131 and 2 Ci/6 for xenon-133.
-------
2.3-12
Sewage Treatment Plants
Most radioisotope releases from hospitals occur via the
liquid pathway. When sewage containing radioisotopes is then
treated in facilities employing sludge drying and combustion,
radionuclides can be emitted into the air. EPA contract 68-01-5049
(Te79) studied the air emissions from a sewage treatment plant and
estimated airborne source terms of 5E-4 Ci/y for iodine-131, and
8E-4 Ci/y for technetium-99m. The study further estimated that
approximately 21 percent of sewage treatment facilities (about 4000)
in the United States use incineration/pyrolysis for sludge treatment.
2.3.4 Typical Facility
Radiopharmaceutical Production Plant
Demographic, meteorological and emission parameters were
developed from actual site characteristics of one of the plants in
table 2.3-7 and by using rural food source parameters for Sites C-F
in table A-2, Appendix A. All of the major medical isotope
production facilities are located in urban settings with fairly high
population densities.
The typical facility produces technetium-99m, xenon-133,
iodine-131, iodine-125, and molybdenum-99/technetium-99m generators
using a reactor-produced neutron activation process and discharges
all airborne releases from a single stack. Atmospheric emissions
from the typical facility are listed in table 2.3-9.
Production figures are considered proprietary but. preliminary
figures (Te79) estimate that the entire industry produces 800 to
1350 curies per year of iodine-131, 1600 to 3000 curies per year of
xenon-133, and 15,600 to 30,600 curies per year of techrietium-99m.
It is reasonable to assume, therefore, that the typical facility
(because it is one of the three largest) produces one-third of the
entire industry's output. We have assumed production values of 350
curies per year of iodine-131, 800 curies per year of xenon-133, and
23,000 curies per year of technetium-99m.
The typical facility employes charcoal bed and HEPA filters
as gaseous effluent control technology; annual emissions after such
treatment are 0.94 curies of iodine-125 and 0.88 curies of
iodine-131.
Hospital
The typical hospital facility is a general care hospital of
more than 500 beds located in the suburbs of a large Midwestern city
-------
2.3-13
(Site B, Appendix A). It is assumed that the hospital has nuclear
medicine capabilities and administers an average of 0.6 curies per
year of iodine-131 and 2.7 curies per year of xenon-133.
Radiopharmaceuticals are prepared in exhaust hoods in the
nuclear medicine section; the exhaust from the building ventilation
system enters the atmosphere without further treatment.
Diagnostic administrations occur in patients' rooms while
therapeutic administrations occur in the nuclear medicine
department. Emissions may be collected by the hospital ventilation
system and enter the atmosphere without further treatment.
Atmospheric emissions are listed in table 2.3-9.
Sewage Treatment Facility
The typical sewage treatment plant developed for assessing
health impacts is located in the suburbs of a large Midwestern city
(Site B, with rural food source fractions, Appendix A). The plant
incinerates the sludge after it has been dried. Atmospheric
emissions are listed in table 2.3-9.
-------
2.3-14
Table 2.3-8. Model facilities of typical producers
and users of radiopharmaceuticals
Parameter
Value
Radiopharmaceutical production
Product line:
Product volume:
Emission control:
Stack:
Height
Diameter
Effluent velocity
Rate of heat emission
Hospital
Size:
Volume of administrations:
Emission control:
plant
Sewage treatment plant
Process:
Iodine-131, xenon-133,
technetium-99m, molybdenum-99/
technetium-99m generators
Iodine-131 - 350 Ci/y
Xenon-133 - 800 Ci/y
Technetium-99m - 23,000 Ci/y
Charcoal/HEPA filters with
release through a single
elevated stack
15.2 meters
1.22 meters
20.1 meters/s
0.0 cal/s
500+ beds
Iodine-131 - 0.6 Ci/y
Xenon-133 - 2.7 Ci/y
Exhaust hoods in nuclear
medicine department; release
through building ventilation
system. In patients' rooms,
release is through building
ventilation system.
Incineration following
sludge drying.
-------
2.3-15
Table 2.3-9. Atmospheric emissions of radionuclides
from typical facilities of producers and users of
radi opharmaceut i cals
Radionuclide
Emissions
(Ci/y)
Radiopharmaceutical production plant
Iodine-125
Iodine-131
Hospital
Iodine-131
Xenon-133
Sewage treatment plant
Iodine-131
Technetium-99m
9.4E-1
8.8E-1
l.OE-5
2.0
5.0E-4
8.0E-4
2.3.5 Health Impact Assessment of Producers and Users of
Radiopharmaceuticals
Table 2.3-10 presents estimates of annual radiation doses
resulting from radioactive emissions from a typical radiopharma-
ceutical production plant, a typical hospital administering
radioisotopes for diagnostic and/or therapeutic purposes, and a
typical sewage treatment plant receiving and treating liquid wastes.
Estimated individual lifetime risks and the number of fatal
cancers to the population resulting from doses from these facilities
are shown in table 2.3-11.
Because the highest doses from these facilities is to the
thyroid, the fatal cancer risk is primarily a risk of fatal thyroid
cancer. The risk of nonfatal cancer of the thyroid is approximately
an order of magnitude higher.
-------
2.3-16
Table 2.3-10. Annual radiation doses due to radioactive emissions
from typical facilities of producers and users of
radiopharmaceuticals
Organ
Maximum
Individual
(mrem/y)
Average
individual
(mrem/y)
Population
(peron-rem/y)
Typical radiopharmaceutical production plant
Lung
Bone
Kidney
Liver
Thyroid
G.I.tract
Other soft tissue
Lung
Bone
Kidney
Liver
Thyroid
G.I. tract
Other soft tissue
Lung
Bone
Kidney
Liver
Thyroid
G.I. tract
Other soft tissue
1.5E-1
2.9E-1
1.5E-1
1.3E-1
8.6E+1
9.0E-2
2.0E-1
3.1E-5
6.2E-5
3.2E-5
2.7E-5
1.3E-2
1.9E-5
4.3E-5-
Typical hospital
7.1E-5
1.5E-4
6.1E-5
5.9E-5
1.3E-4
3.9E-5
7.7E-5
2.9E-7
6.1E-7
2.4E-7
2.4E-7
5.3E-7
1.5E-7
3.1E-7
Typical sewage treatment plant
5E-5
6E-5
2E-5
1E-5
2.6E-2
3.7E-5
4.7E-5
2.6E-8
3.3E-8
2.4E-8
2.3E-8
5.8E-6
2.0E-8
2.7E-8
4.7E-1
9.3E-1
4.8E-1
4.0E-1
2.0E+2
2.8E-1
6.5E-1
7.1E-1
1.5
6E-1
5.9E-1
3.8E-1
1.3
7.7E-1
6.3E-2
8.1E-2
6E-2
5.6E-2
1.4E+1
4.9E-2
6.6E-2
-------
2.3-17
Table 2.3-11. Individual lifetime risks and number of fatal cancers
due to radioactive emissions from typical facilities of
producers and users of radiopharmaceuticals
Source
Individual lifetime risks3
Maximum Average
individual individual
Expected fatal cancers
per year of operation3
(Fatal Cancers)
Typical
production
plant
8.1E-06
4.4E-10
2.9E-4
Typical
hospital
Typical sewage
treatment
plant
9.4E-10
2.3E-09
3.7E-12
7.0E-13
1.3E-7
2.5E-8
aThe fatal cancer risk is primarily a risk of fatal thyroid
cancer. The risk of nonfatal thyroid cancer is approximately and
order of magnitude higher.
-------
2.3-18
REFERENCES
AEC66 Atomic Energy Commission, 1966, AEC Symposium Series No. 6,
Radioactive Pharmaceuticals, CONF-651111, April 1966.
AEC71 Atomic Energy Commission, 1971, The Nuclear Industry, WASH
1174-71, 1971.
AEC74 Atomic Energy Commission, 1974, The Nuclear Industry, WASH
1174-74, 1974.
Ba66 Baker P. J., 1966, Reactor-Produced Radionuclides in AEC
Symposium Series No. 6, Radioactive Pharmaceuticals, CONF651111,
April 1966.
BRH70 Bureau of Radiological Health, 1970, Survey of the Use of
Radionuclides in Medicine, BRH/DMRE 70-1, January 1970.
FDA76 Food and Drug Administration, 1976, A Pilot Study of Nuclear
Medicine Reporting through the Medically Oriented Data System,
HEW(FDA) 76-8045, June 1976.
McG77 McGraw-Hill Encyclopedia of Science and Technology, Vol. 11,
1977, New York.
NRC77 Nuclear Regulatory Commission, 1977, Annual Report---1977, NRC,
Washington, D.C.
NRC79 Nuclear Regulatory Commission, 1979, Private communication from
G. Wayne Kerr, NRC, Washington, D.C.
Te79 Teknekron, Inc., 1979, Draft Final Report, A Study of Airborne
Radioactive Effluents from the Radiopharmaceutical Industry, EPA
Contract No. 68-01-5049, March 1979.
Wa68 Wagner H. N., Jr., M.D., 1968, Principles of Nuclear
Medicine, W. B. Saunders Co., Philadelphia, 1968.
-------
2.4-1
2.4 Test Reactors
2.4.1 General Description
The Department of Energy divides this category into three
types: 1) general irradiation test reactors, 2) high-power research
and test reactors, and 3) safety research and test reactors.
A general irradiation test reactor is defined as having: 1)
a thermal power level exceeding 10,000 kW, 2) test loops or
experimental facilities within or in proximity to the core, and 3)
the use of nuclear radiation for testing the life or performance of
reactor components as its major function.
A high-power research and test reactor is defined as having
relatively high thermal power level (5000 kW or more) but not
classified as a general irradiation test reactor.
A safety research and test reactor is defined as a reactor
associated with nuclear safety research or engineering-scale test
program conducted for the purpose of developing basic information or
demonstrating safety characteristics of terrestrial and aerospace
nuclear reactor systems (DOE78).
As of June 30, 1978, there were 15 reactors of this category
in operation within the United States (DOE78). Two of these
reactors, the National Bureau of Standards reactor and the Union
Carbide Corporation reactor, are facilities not operated by the
Department of Defense or Department of Energy (table 2.4-1).
The National Bureau of Standards reactor (NBSR) complex is
located northwest of Washington, D.C., near Gaithersburg, Md. The
reactor is an enriched-uranium, heavy-water cooled and moderated
vessel type unit. The facility began operation in 1967 and has an
available power capacity of 10 M watts. The high fluxes generated
by the NBSR are used primarily to measure fundamental properties of
matter and to develop new standards and measuring techniques.
2.4.2 Process Description
Test reactors have been constructed and operated in the
United States since the early 1950's. Several types of test
reactors, such as heavy water, graphite, tank and pool, have been
constructed and used. These reactors have been used primarily to
test new reactor designs, reactor components, safety features and to
develop material standards.
-------
2.4-2
Table 2.4-1. Test reactors (other than DOD or DOE)
Startup
Facility Type Power date
National Bureau of Standards Heavy water 10 MW 1967
Gaithersburg, Maryland
Union Carbide Corp.a
Tuxedo, N. Y.
aThis reactor is used to produce radiopharmaceuticals and is
discussed in section 2.3 "of this report.
Table 2.4-2. Atmospheric emissions from test reactors
(other than DOD or DOE facilities)
Facility Emissions
(Ci/y)
National Bureau of Standards
Tritium 155
Argon-41 465
Union Carbide (a)
FIRNot reported.
aThis reactor is used to produce radiopharmaceuticals and is
discussed in section 2.3 of this report.
-------
2.4-3
2.4.3 Emissions of Radionuclides
The airborne emission of radioactive materials from test
reactors is usually confined to argon-41, tritium, and small
quantities of fission products and noble gases.
The National Bureau of Standards reactor facility released
approximately 155 curies of tritium and 465 curies of argon-41 to
the environment as gaseous waste in 1977 (table 2.4-2). These were
the only detectable radioisotopes released (To78).
2.4.4 Typical Facility
To analyze the impact of test reactors, the parameters of a
typical facility (table 2.4-3) have been used in projecting
population dose rates. The atmospheric emissions of radioactive
materials are listed in table 2.4-4.
2.4.5 Health Impact Assessment of Typical Facility
Table 2.4-5 presents estimated annual radiation doses
resulting from radioactive emissions from the typical test reactor
facility. These estimates are for a site located in the suburbs of
a large Midwestern city (Site B, Appendix A).
Individual fatal cancer risks and committed population fatal
cancers are presented in table 2.4-6. The lifetime fatal cancer
risk to the maximum exposed group of individuals is estimated to be
about 2.4E-6. The lifetime fatal cancer risk to the average
individual living within 80 kilometers of the typical facility is
estimated to be 5.0E-9.
The estimated number of fatal cancers to the population
living in the region around the facility is estimated to be 1.8E-4
per year of reactor operation. The world-wide fatal cancer
commitment (tritium) is estimated to be 9.0E-5.
-------
2.4-4
Table 2.4-3. Typical test reactor facility
Parameter Value
Type of facility: Enriched uranium, heavy
water cooled and
moderated test reactor
Power level: 10 M watt
Stack:
Height 23.2 meters
Diameter 1.5 meters
Effluent velocity 12.7 meters/s
Rate of heat emission 0.0
Table 2.4-4. Atmospheric emissions of radionuclides
from the typical test reactor facility
Radionuclide Emissions
(Ci/y)
Argon-41 450
Tritium 150
-------
2.4-5
Table 2.4-5. Annual radiation doses due to radioactive
emissions from a typical test reactor facility
Organ
Lung
Bone
Kidney
Liver
G.I. Tract
Thyroid
Other soft
Maximum
Individual
(mrem/y)
2.1E-1
2.1E-1
2.1E-1
2.1E-1
2.1E-1
2.1E-1
tissue 2.1E-1
Average
Individual
(mrem/y)
4.4E-4
4.4E-4
4.4E-4
4.4E-4
4.4E-4
4.4E-4
4.4E-4
Population
(Person-rem/y)
1.1
1.1
1.1
1.1
1.1
1.1
1.1
Table 2.4-6. Individual lifetime risks and number of fatal
cancers due to radioactive emissions from a typical
test reactor facility
Source
Individual
Maximum
individual
lifetime risks
Average
individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Typical facility 2.4E-6
5.0E-9
1.8E-4
Source
Expected total cancers
in the worldwide population^
over the next 100 years
(Fatal cancers)
Typical facility
9.0E-5
aFrom 150 Ci/y of tritium.
-------
2.4-6
REFERENCES
AEC67 Major Activities in the Atomic Energy Programs, January-
December 1967.
DOE78 Department of Energy, 1978, Nuclear Reactors Built, Being
Built or Planned in the United States as of June 30, 1978,
TID-8200-R-38.
To78 Torrence, J., 1978, NBS Private Communication.
Un77 United Nations, 1977, Sources and Effects of Ionizing Radia-
tion, United Nations Scientific Committee on the Effects of
Atomic Radiation, 1977, Report to the General Assembly, with
annexes, United Nations Publication Sales No. E-77, IX.1.
-------
2.5-1
2.5 Research Reactors
2.5.1 General Description
A research reactor, excluding those located at a university,
uses nuclear radiation as a tool for basic or applied research, and
has a thermal power level less than 5,000 kW (DOE78). As of June
1978, there were 24 research reactors operating in the United
States. The 13 reactors not located in Department of Energy or
Department of Defense facilities are listed in table 2.5-1 (DOE78).
2.5.2 Process Description
Several reactor designs are used in research reactors, as
noted in table 2.5-1. The rated power output of these reactors
ranges from negligible to 2,000 kW(t). A research reactor is
generally designed to provide a source of neutrons and/or gamma
radiation for research into basic or applied physics, biology, or
chemistry or to aid in the investigation of the effects of radiation
on materials.
2.5.3 Emissions of Radionuclides
The NRC regulations, under which all research reactors
operate, do not require the submission of annual effluent release
data. Therefore, there are no effluent release data for research
reactors (NRC77). However, some licensees do submit release data to
the NRC in conjunction with annual operating reports. All available
data from NRC is shown in table 2.5-1 for 13 facilities (NRC77).
The variation in the emission data shown in table 2.5-1 may
be attributed to parameters such as power level, operating time, and
emission control systems which influence annual emissions for
research reactors.
Because this category of reactors is used for a variety of
research purposes, operating times and power levels vary
accordingly. These parameters have not been compiled for each
facility listed in table 2.5-1.
2.5.4 Model Facility
In order to estimate population and individual radiation
doses, a model facility (table 2.5-2) was developed by assigning the
various parameters that are important in assessing impacts.
-------
2.5-2
Table 2.5-1. Research reactors (other than DOD or DOE)
Facility
Type
Airborne Releases
Power Startup Argon-41 Time
kW(t) date Ci Period
Aerotest Operations, Pool
San Ramon, Calif. TRIGA core
Babcock & Wilcox Pool
Lynchburg Pool Reactor
Lynchburg, Va.
Dow Chemical Co. U Zr
Midland, Mich. hydride
General Atomic Co. U Zr
TRIGA-Mk 1 hydride
Prototype Reactor
LaJolla, Calif.
General Atomic Co. U Zr
Advanced TRIGA-MK 1 hydride
Prototype Reactor
LaJolla, Calif.
General Electric Light
Nuclear Test Reactor water
Pleasanton, Calif.
Northrop Corp. Labs. U Zr
(Space Radiation Lab.) hydride
Hawthorne, Calif.
Nuclear Examination Homog.
Reactor (Rockwell
International),
Santa Susana, Calif.
Omaha Veterans Adm. U Zr
Hospital hydride
Omaha, Neb.
250 1965 3.0E-3 1972
1,000 1958 (a)
1974
100 1967 l.OE-3 1972
250 1958 6.6E-2 1976
1,500 1960 1.61E-1 1976
100 1957 (b) 1972
1,000 1963 9.152 1976
3 1952 NA NA
18 1959 3.0E-1 1972
See footnotes at end of table.
-------
2.5-3
Table 2.5-1. Research reactors (other than DOD or DOE)--continued
Airborne Releases
Power Startup Argon-41 Time
Facility Type kW(t) date Ci Period
Rhode Island Nuclear Pool 2,000 1964 247 1972
Science Center
Fort Kearney, RI
Rockwell Inter- Homog. NEG. 1958 NA NA
national, Canoga Park,
Calif.
U.S. Geological U Zr 1,000 1969 NA NA
Survey Lab. (Dept. hydride
of the Interior),
Denver, Colo.
Westinghouse Nuclear 10 1972 NA NA
Training Center
Zion, 111.
NA not available.
NEG. Negligible.
aAmount of argon-41 not available but 1.3E-4 Ci of other isotopes were
released.
^Amount of argon-41 not specified but 27 Ci of noble gases were
released.
Emissions
The model facility was assumed to release 200 curies of argon-41
and 100 curies of tritium per year through a 23-meter stack at a velocity
of 12.7 m/s (table 2.5-3). These releases were assumed even though they
are higher than any shown in table 2.5-1. Thus, they represent a
reasonable upper bound for emissions.
-------
2.5-4
Table 2.5-2. Model research reactor facility
Parameter Value
Type: Enriched uranium,
light-water-cooled and
moderated research reactor
Power level: 1 M Watt
Stack:
Height 23.2 meters
Diameter 1.5 meters
Effluent velocity 12.7 meters/s
Rate of heat emission 0.0
Table 2.5-3. Atmospheric emissions from the model research
reactor facility
Emissions
Radionuclide (Ci/y)
Argon-41 200
Tritium 100
-------
2.5-5
2.5.5 Health Impact Assessment of a Model Facility
Table 2.5-4 estimates the annual radiation doses resulting
from radioactive emissions from a model research reactor facility.
These estimates are for a site in the suburbs of a large Midwestern
city (Site B, Appendix A).
Individual fatal cancer risks and committed population fatal
cancers are presented in table 2.5-5. The lifetime fatal cancer
risk to the maximum exposed group of individuals is estimated to be
approximately l.OE-6. The lifetime fatal cancer risk to the average
individual living within 80 kilometers of the model facility is
estimated to be 2.4E-9.
The number of fatal cancers committed to the population
living in the region near facility is estimated to be 8.2E-5 per
year of reactor operations, and the worldwide fatal cancer
commitment is estimated to be 6.0E-5 per year of reactor operation.
Table 2.5-4. Annual radiation doses due to radioactive
emissions from a model research reactor facility
Organ
Lung
Bone
Kidney
Liver
S.I. Tract
Thyroid
Other soft
tissue
Maximum
Individual
(mrem/y)
9.1E-2
9.1E-2
9.1E-2
9.1E-2
9.1E-2
9.1E-2
9.1E-2
Average
Individual
(mrem/y)
2.1E-4
2.1E-4
2.1E-4
2.1E-4
2.1E-4
2.1E-4
2.1E-4
Population
(person-rem/y)
5.1E-1
5.1E-1
5.1E-1
5.1E-1
5.1E-1
5.1E-1
5.1E-1
-------
2.5-6
Table 2.5-5. Individual lifetime risks and population
fatal cancers due to radioactive emissions
from a model research reactor facility
Source
Individual lifetime risks
Maximum Average
individual individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Model facility l.OE-6
2.4E-9
8.2E-5
Source
Expected total cancers
in the worldwide population
over the next 100 years
(Fatal cancers)
Model facility
6.0E-5
-------
2.5-7
REFERENCES
DOE78 Department of Energy, 1978, Nuclear Reactors Built, Being
Built or Planned in the United States as of June 30, 1978,
TID-8200-R-38.
NRC77 Letter from J. Kastner, NRC, to W. Mills, EPA, with 59-page
enclosure of test, research, and university reactor effluent data,
12/7/77.
-------
2.6-1
2.6 University Reactors
2.6.1 General Description
University reactors are defined as those located at a
university and usually operated for the primary purpose of training
in the operation and utilization of reactors and for the instruction
in reactor theory and performance (DOE78). As of June 30, 1978,
there were 54 such reactors (DOE78) operable in the United States
(table 2.6-1).
2.6.2 Process Description
University reactors are used primarily as a teaching tool or
for basic and applied research. Several types and designs are in
use as shown in table 2.6-1. The capable power levels of these
reactors is generally low, ranging from negligible to 10 MW.
2.6.3 Emissions of Radionuclides
Due to the relatively low power levels and in some cases
intermittent operation, the amounts of radioactive emissions from
university reactors are relatively small. University reactors,
unlike power reactors, are not required as part of their operating
license to report effluent release data to the Nuclear Regulatory
Commission (NRC). However, some universities do submit these data
to the NRC as a part of their annual operating report and these data
are shown in table 2.6-1 (NRC77).
The predominant waste product released by university reactors
is argon-41 which is discharged through the facility stack. Argon-
41 is produced by neutron activation of argon-40, a naturally
occurring component of air. The maximum permissible air
concentration (MPC)a for environmental levels of argon-41 is 4.0E-8
microcuries/cc as specified in 10 CFR 20. Most university reactors
usually restrict their stack releases to levels below this (MPC)a
which is quite conservative since dilution by air in the environment
is not taken into account. As a result, the atmospheric
concentration of argon-41 in the vicinity of university reactors is
generally well below (MPC)a levels.
The differences in reported emission values shown in table
2.6-1 are probably due to variations in operating time, power
levels, and control systems. Universities with large student
-------
2.6-2
demand for reactor time obviously operate their reactors longer during
a given year than do universities with smaller programs. Specific
data concerning operating parameters of the reactors listed in
table 2.6-1 are not compiled.
Table 2.6-1. University reactors (other than DOD or DOE)
Facility
Brigham Young
University
Provo, Utah
California State
Polytechnic
San Luis Obispo, Calif.
Catholic Univ.
of America
Washington, D.C.
Columbia Univ.
New York, N.Y.
Cornell Univ.
Ithaca, N.Y.
Cornell Univ.
Zero Power Reactor
Ithaca, N.Y.
Georgia Institute
of Technology
Atlanta, Ga.
Georgia Tech.
Research Reactor
Type
Homog.
Homog.
solid
Homog.
solid
U Zr
hydride
U Zr
hydride
Tank
Homog.
solid
Heavy
water
Power
kW(t)
NEG.
NEG.
NEG.
250
100
NEG.
NEG.
10,000
Startup
date
1967
1973
1957
1977
1962
1962
1957
1964
Airborne
Argon-41
(Ci)
NA
NA
NA
NA
1.16
NA
NA
466.87
Releases
Time
Period
NA
NA
NA
NA
1972
NA
NA
1976
Atlanta, Ga
See footnotes at end of table.
-------
2.6-3
Table 2.6-1. University reactors (other than DOD or DOE)continued
Facility
Idaho State Univ.
Pocatello, Idaho
Iowa State Univ.
Ames, Iowa
Kansas State
University
Manhattan, Kan.
Manhattan College
New York, N.Y.
Massachusetts
Institute of Tech.
Cambridge, Mass.
Memphis State
Memphis, Tenn.
Michigan State
University
East Lansing, Mich.
North Carolina
State University
Raleigh, N.C.
Nuclear Science
Center Reactor
Texas A&M Univ.,
College Station, Texas
Ohio State Univ.
Columbus, Ohio
Oregon State
University
Corvallis, Ore.
Type
Homog.
solid
Graphite
water
U Zr
hydride
Tank
Heavy
water
reflected
Homog.
solid
U Zr
hydride
Pool
Pool
TRIGA
core
Pool
Homog.
solid
Power
kW(t)
NEG.
10
250
NEG.
5,000
NEG.
250
1,000
1,000
10
NEG.
Startup
date
1967
1959
1962
1964
1958
1977
1969
1972
1961
1961
1958
Airborne Releases
Argon-41 Time
(Ci) Period
NA NA
0.4 1972
6.0E-4 1972
NA NA
a7795.0 7/76-6/77
NA NA
1.1E-3 1976
NA NA
3.35 1976
3.4E-2 1972
NA NA
See footnotes at the end of the table.
-------
2.6-4
Table 2.6-1. University reactors (other than DOD or DOE)--continued
Facility
Oregon State
Corvallis, Ore.
Penn. State TRIGA
Reactor, Penna.
State University,
Univ. Park, Pa.
Purdue University
West LaFayette, Ind.
Reed College
Portland, Ore.
State Univ. of
New York
(Western New
York Nuclear
Research Center, Inc.)
Buffalo, N.Y.
Texas A&M Univ.
College Station,
Texas
Tuskegee Institute
Tuskegee, Ala.
Univ. of Arizona
Tucson, Ariz.
Univ. of Calif.
Berkeley, Calif.
Univ. of Calif.
Santa Barbara,
Calif.
Type
U Zr
hydride
Pool
TRIGA
core
Pool
U Zr
hydride
Pool
Homog.
solid
Homog.
solid
U Zr
hydride
U Zr
hydride
Homog.
Power
kW(t)
1,000
1,000
10
250
2,000
NEG.
NEG.
250
1,000
NEG.
i ^
Airborne
Startup Argon-41
date (Ci)
1967 20.06
1965 1
1962 NA
1968 NA
1961 6.9
1957 NA
1957 NA
1958 NA
1966 32
None
Releases
Time
Period
7/76-6/77
1972
NA
NA
1972
NA
NA
7/76-5/77
1972
See footnotes at the end of the table.
-------
2.6-5
Table 2.6-1. University reactors (other than DOD or DOE)--continued
Facility
Univ. of Calif.
Irvine, Calif.
Univ. of Calif
at Los Angeles
School of
Engineering &
Applied Science
Los Angeles, Calif.
Univ. of Delaware
Newark, Del a.
Univ. of Florida
Gainesville, Fla.
University of
Illinois
Urbana, 111.
University of
Illinois
Urbana Champaign, 111.
Univ. of Kansas
Lawrence, Kansas
Univ. of Lowell
Lowell, Mass.
Univ. of Md.
College Park, Md.
Univ. of
Michigan
(Ford Nuclear Reactor)
Ann Arbor, Mich.
Type
U Ar
hydride
Graphite
water
Homog.
solid
Gaphite/
water
U Zr
hydride
U Zr
hydride
Pool
Pool
Tank
Pool
Airborne
Power Startup Argon-41
kW(t) date (Ci)
250 1969 4.2E-2
100 1960 33.0
NEG. 1958 NA
100 1959 5.03
10 1971 NA
1,500 1960 b2.60
10 1961 NA
1,000 NA
350 1960 NA
2,000 1957 (c)
Releases
Time
Period
7/77-6/77
1976
NA
9/75-8/76
NA
1976
7/76-6/77
NA
NA
1976
See footnotes at the end of the table.
-------
2.6-6
Table 2.6-1. University reactors (other than DOD or DOE)--continued
Facility
Univ. of
Missouri
Columbia, Mo.
Univ. of
Missouri at Rolla
Rolla, Mo.
Univ. of
New Mexico
Albuquerque, N.M.
Univ. of
Oklahoma
Norman, Ok la.
Univ. of Texas
Austin, Texas
Univ. of Utah
Salt Lake City,
Type
Tank
Pool
Homog.
solid
Homog.
solid,
Pool
U Zr
hydride
U Zr
hydride
Power
kW(t)
10,000
200
NEG.
NEG.
250
250
Startup
date
1966
1961
1957
1958
1963
1975
Airborne
Argon-41
(CD
d2,406
(e)
NA
NA
NA
NA
Releases
Time
Period
6/76-7/77
1976
NA
NA
1976
NA
Utah
Univ. of
Virginia
Charlottesville, Va.
Univ. of Pool
Virginia
Charlottesville, Va.
NEG.
2,000 1960
Univ. of
Washington
Seattle, Wash.
Graphite/ 100 1961
water
f0.442
NA
13.0
University of Pool
Wisconsin TRIGA
Madison, Wise. core
1976
NA
1976
1,000 1960 1.77 7/77-6/76
See footnotes at end of the table.
-------
2.6-7
Table 2.6-1. University reactors (other than DOD or DOE)--continued
Airborne Releases
Power Startup Argon-41 Time
Facility Type kW(t) date (Ci) Period
Virginia Graphite/ 100 1959 180.14 1976
Polytechnic water
Institute
Blacksburg, Va.
Washington Pool 1,000 1961 7.67 7/76-6/77
State Univ. TRIGA
Pullman, Wash. core
Worcester Pool 10 1959 NA. NA
Polytechnic
Institute
Worcester, Mass.
NA not available.
NEG. Negligible.
aAlso 7.04 Ci of tritium.
bAlso l.OE-3 Ci of tritium.
CA total of 107.9 Ci of unspecified radioactivity.
dAlso 4.8 Ci of tritium.
eAlso 0.115 of unspecified radioactivity.
fAlso 0.013 Ci of krypton-85.
2.6.4 Model Facility
In order to estimate population and individual radiation doses,
a model facility (table 2.6-2) has been developed.
Emissions
Atmospheric emissions from the model facility are reported in
table 2.6-3. The argon-41 releases are somewhat high when compared
with the majority of the source terms given in table 2.6-1; however,
two reported values are greater than 1,000. The assumed annual
-------
2.6-8
release for tritium is higher than any shown in table 2.6-1. The
argon-41 release term assumed here, therefore, represents a
reasonable upper bound for this source category.
Table 2.6-2. Model university reactor facility
Parameter Value
Type: Enriched uranium, heavy-
water moderated and cooled
university reactor
Power level: 10,000 kW(t)
Stack:
Height 23.2 meters
Diameter 1.5 meters
Effluent velocity 12.7 meters/s
Rate of heat emission 0.0
Table 2.6-3. Atmospheric emissions of radionuclides
from the model university reactor facility
Radionuclide Emissions
(Ci/y)
Argon-41 1,000
Tritium 100
-------
2.6-9
2.6.5 Health Impact Assessment of A Model University
Reactor Facility
Table 2.6-4 presents estimates of annual radiation doses
resulting from radioactive emissions from a model university reactor
facility. These estimates are for a site located in the suburbs of
a large Midwestern city (Site B, Appendix A).
Individual fatal cancer risks and committed population fatal
cancers are presented in table 2.6-5. The lifetime fatal cancer
risk to the maximum exposed group of individuals is estimated to be
approximately 4.9E-6. The lifetime fatal cancer risk to the average
individual living within 80 kilometers of the model facility is
estimated to be 1.1E-8.
The number of fatal cancers committed to the population
living in the region around the facility is estimated to be 3.9E-4
per year of reactor operation. The estimated number of fatal
cancers committed to the world population is 6.0E-5 per year of
reactor operation.
-------
2.6-10
Table 2.6-4. Annual radiation doses due to radioactive emissions
from a model university reactor facility
Organ
Lung
Bone
Kidney
Liver
G.I. Tract
Thyroid
Other soft
tissue
Maximum
individual
(mrem/y)
4.3E-1
4.3E-1
4.3E-1
4.3E-1
4.3E-1
4.3E-1
4.3E-1
Average
individual
(mrem/y)
9.7E-4
9.7E-4
9.7E-4
9.7E-4
9.7E-4
9.7E-4
9.7E-4
Population
(person-rem/y)
2.4
2.4
2.4
2.4
2.4
2.4
2.4
Table 2.6-5. Individual lifetime risks and population
fatal cancers due to radioactive emissions from a
model university reactor facility
Source
Individual lifetime risks
Maximum Average
individual individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Model facility 4.9E-6
1.1E-8
3.9E-4
Source
Expected total cancers
in the worldwide population
over the next 100 years
(Fatal cancers)
Model facility
6.0E-5
-------
2.6-11
REFERENCES
DOE78 Department of Energy, 1978, Nuclear Reactors Built, Being
Built or Planned in the United States as of June 30, 1978,
TID-8200-R-38.
NRC77 Letter from J. Kastner, NRC, to W. Mills, EPA, with 59-
page enclosure of test, research, and university reactor
effluent data. 12/7/77.
-------
2.7-1
2.7 Shallow Land Burial of Low-Level Radioactive Wastes
2.7.1 General Description
Current practice in the United States is to dispose of solid
low-level radioactive waste by shallow land burial. These wastes
are generated primarily by the nuclear fuel cycle, nuclear medicine
facilities, commercial manufacturers, universities, and private and
governmental research institutions.
Table 2.7-1 lists the six commercial low-level waste burial
facilities and five major active sites operated by Department of
Energy (Ad78). Three of the commercial facilities are currently
inoperative. Maxey Flats, Ky, and West Valley, N.Y., are closed
indefinitely and the responsibility for perpetual care is expected
to be shifted to the respective States. The Sheffield, Illinois,
site is awaiting approval from NRC and the State of Illinoisto
expand its storage capacity.
2.7.2 Process Description
In general the wastes are placed as received in a trench
excavated in the existing till or soil and the material thus removed
is used to cover the wastes once the trench is filled to capacity.
This overburden is sometimes compacted and usually mounded to
promote surface water runoff. The dimensions of trenches and their
proximity to adjacent trenches vary within and between sites.
The character, source, and amount of specific radionuclides
contained in the waste differ somewhat from site to site. However,
the operational characteristics and the compositions of the wastes
are generally similar.
2.7.3 Emissions of Radionuclides
Emissions from Normal Waste Handling Operations
The routine procedures of transporting, packaging, and
burying the waste at the low-level radioactive waste burial site may
result in the release and subsequent dispersion of radioactive
material into the atmosphere. All radionuclides contained in the
waste are potential sources.
-------
2.7-2
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-------
2.7-3
Resuspension of Contaminated Soil
The surface soil of a waste burial site can become
contaminated with radionuclides during normal operations. The
contaminated soil particles may then be resuspended and transported
offsite. Those particles which are respirable may contribute a dose
to the respiratory tract of persons in the general population.
These contaminants may also expose persons in the surrounding
population through uptake and transport in other pathways, e.g.,
uptake into food crops. All radionuclides routinely received at the
waste burial site could be resuspended.
Emissions from Trenches
Long-lived radionuclides either existing in a gaseous state
or those forming gaseous compounds in the trench gas may migrate
through the trench cover and be dispersed to areas off the waste
burial sites and result in doses to the public. Radionuclides which
may be released are krypton-85, radon-222, carbon-14, and tritium.
Emissions from Leachate Evaporator
Waste trenches excavated in soil (till) which is impervious
to water in its undisturbed state retain most surface water entering
through a fractured trench cap. Thus, trenches in humid (high
precipitation) areas may overflow if the water collected (leachate)
is not periodically removed. At the Maxey Flats site, after pumping
and storage, the leachate is concentrated to solids by an
evaporator. The vaporized liquid containing volatized radionuclides
is released to the atmosphere.
2.7.4 Typical Facility
To estimate airborne emissions and the health impact of
low-level radioactive waste burial sites, the Maxey Flats site was
chosen because it is considered to be representative of sites
located in humid areas and most of the data needed for calculations
are available. However, since there were no data on gaseous
releases from trench caps, West Valley data was used because that
site is similar in geology, meteorology, and in the type of buried
wastes. In addition both sites use leachate evaporators to avoid
trench overflow. The data on emissions from leachate evaporators
are presented separately to assess its impact.
Typical facility parameters are given in table 2.7-2. A
comprehensive inventory of the radioactivity buried at Maxey Flats
-------
2.7-4
Table 2.7-2. Typical low-level radioactive waste burial site
Parameter
Value
Area
Radionuclide inventory
Annual rainfall
Evaporator stack height
Regional population
1.3
(See table 2.7-3)
125 cm/y
10 meters
387,961
Table 2.7-3. Annual atmospheric emissions of radionuclides
and inventory of principal radionuclides at a typical
low-level radioactive waste burial site
Radionuclide
Total in place
at site (Ci)
Emissions (Ci/y)
Normal
Operations
Trenches
Leachate
Evaporator
Tr i t i urn
Carbon-14
Iron-55
Cobalt-58
Cobalt-60
Strontium-90
Technetium-99
Iodine-129
Cesium-134
Cesium-137
Radium- 226
Uranium-233
Uranium-235
Uranium-238
Plutonium-238
Plutonium-239
5.4E+5
3.6E+4
1.2E+2
7.0E-2
8.1E+4
1.5E+4
9.3
1.1E-2
3.9E-1
2.5E+4
4.9E+3
9.5E+2
3.5E-1
8.5E+1
4.3E+4
a3.8E+3
5.4E-03 80.0
3.6E-04 5.0
1.2E-06
7.0E-10
8.1E-04
1.5E-04
9.3E-08
1.1E-10
3.9E-09
2.5E-04
4.9E-05
9.5E-06
3.5E-09
8.5E-07
4.3E-04
3.8E-05
1.9E+4
5.6E-3
_
7.8E-4
2.4E-3
1.9E-3
-
-
8.8E-4
3.5E-2
-
-
-
-
4.3E-4
8.1E-6
alncludes 2.1E+02 of plutonium of uncertain isotopic origin.
-------
2.7-5
was compiled from radioactivity shipment records by Gat et al.
(GA75) in 1975 and was recently updated by Prairie (PR76). The
inventory of the principal radionuclides is shown in table 2.7-3.
Emissions from Normal Waste Handling Operations
Adams has calculated emissions for major radionuclides from
normal waste handling operations at Maxey Flats (Ad78). The same
methodology has been used in calculating the releases in table
2.7-3, but the nuclide inventory is from Prairie (Pr76).
Emissions from Trenches
The Office of Radiation Programs is currently studying
gaseous releases from trenches at four low-level burial sites. The
trench pathway was identified by Davis et al. (Da76) in studies at
West Valley. The only data currently available are carbon-14 and
tritium as methane in the forms of ^CH4 and ^HCH3,
respectively. These data have been reported by Lu, Matuszek, et al.
(Lu78, Ma78) and are given in table 2.7-3. The ORP studies will
identify the magnitude of carbon-14, tritium, krypton-85, and
radon-222 releases at four sites as well. Carbon-14 and tritium
releases at Maxey Flats are not expected to differ greatly from
those at West Valley because of the similarities of the two sites.
Emissions from Leachate Evaporator
This pathway has been studied by Blanchard, et al. (B178).
The average annual release is calculated by using the product of the
average discharge rates and the annual operating time for the
evaporator (table 2.7-3).
2.7.5 Health Impact Assessment of a Typical Low-Level Radioactive
Waste Burial Site
Table 2.7-4 estimates annual radiation doses resulting from
radioactive emissions from the typical low-level radioactive waste
burial site. The estimates are for a rural, low population density
site in Fleming County, Kentucky.
Table 2.7-5 estimates the individual lifetime risks and fatal
cancers to the population from these doses. The lifetime cancer
risk to the highest exposed group of individuals is approximately
8.5E-5 resulting primarily from emissions from the leachate
evaporator. The lifetime cancer risks to the average individual in
the region is estimated to be 5.0E-8.
-------
2.7-6
Table 2.7-4. Annual radiation doses from radioactive emissions
from a typical low-level radioactive waste burial site
Organ
Lung
Bone
Kidney
Liver
Thyroid
G.I. Tract
Other soft tissue
Maximum
individual
(mrem/y)
1.8E+1
2.5E+1
1.9E+1
1.9E+1
1.9E+1
1.7E+1
1.9E+1
Leachate evaporator
Average
individual
(mrem/y)
1.1E-2
1.2E-2
1.1E-2
1.1E-2
1.1E-2
1.1E-2
1.1E-2
Population
(person-rem/y)
4.1
4.6
4.2
4.2
4.2
4.1
4.2
Organ
Lung
Bone
Kidney
Liver
Thyro i d
G.I. Tract
Other soft tissue
Maximum
individual
(mrem/y)
6.8E-1
5.4
9.6
1.1
5.5E-1
6.4E-1
8.9E-1
Area sources
Average
individual
(mrem/y)
4.9E-4
2.4E-2
6.4E-4
7.2E-4
4.9E-4
7.5E-4
9.5E-4
Population
(person-rem/y)
1.9E-1
9.3E-1
2.5E-1
2.8E-1
1.9E-1
2.9E-1
3.7E-1
Organ
Lung
Bone
Kidney
Liver
Thyro i d
G.I. Tract
Other soft tissue
Maximum
individual
(mrem/y)
1.9E+1
3.0E+1
2.0E+1
2.0E+1
2.0E+1
1.8E+1
2.0E+1
Total
Average
individual
(mrem/y)
1.1E-2
1.4E-2
1.2E-2
1.2E-2
1.1E-2
1.2E-2
1.2E-2
Population
(person-rem/y)
4.3
5.5
4.5
4.5
4.4
4.4
4.6
-------
2.7-7
The number of fatal cancers per year of site operation is
estimated to be 7.5E-4 to the population in the region around the
site and 3.3E-2 to the world population. These result primarily
from emissions from the leachate evaporator. The number of fatal
cancers committed to the world population is the impact over a
100-year period of time from emissions of carbon-14 and tritium
during a one-year period.
Health impacts were assessed for a plant with an assumed
operating lifetime of 25 years.
Table 2.7-5. Individual lifetime risks and population fatal cancers
due to radioactive emissions from a typical low-level
radioactive waste burial site
Individual lifetime risks3
Area Maximum Average
individual individual
Total
Expected fatal cancers
per year of operation
(Fatal cancers)
Leachate
evaporator
Area sources
8.0E-5
6.8E-6
4.5E-8
4.3E-9
6.8E-4
6.6E-5
8.5E-5
5.0E-8
7.5E-4
Radionuclide
Expected total cancers in the
worldwide population over the
next 100 years
Leachate Area
evaporator sources
Tritium
Carbon-14
Totalb
1.2E-2
2.5E-5
1.2E-2
4.8E-5
2.1E-2
2.1E-2
over the lifetime of an individual for an exposure
to the dose rates shown in table 2.7-5 for a 25-year period.
^Impact over 100-year period from one-year release of carbon-14
and tritium.
-------
2.7-8
REFERENCES
Ad78 Adams J. A., and Rogers V. L., 1978, A Classification
System for Radioactive Waste DisposalWhat Waste Gas Where?,
NUREG-0456, Nuclear Regulatory Commission, Washington, D.C.
B178 Blanchard R. L., Montgomery D. M., Kolde H. E., and
Meyer G. L., 1978, Supplementary Radiological Measurements
at the Maxey Flats Radioactive Waste Burial Site--1976-1977,
EPA-520/5-78-011, Environmental Protection Agency, Eastern
Environmental Radiation Facility, Montgomery, Ala.
Da76 Davis J. F., Fakundiny R. H., Pferd J. IV, 1976, Evaluation
of Radionuclide Pathways at a Shallow Low-Level Radioactive
Waste Burial Site in Western New York, presented at the National
Meeting of the Geological Society of America, November 10, 1976.
Ga75 Gat I)., Thomas J. D., and Clark D. T., 1975, Radioactive
Waste Inventory at the Maxey Flats Nuclear Waste Burial Site,
Health Physics, 30, 281.
Lu78 Lu A. H. and Matuszek J. M., 1978, Transport through a
Trench Cover of Gaseous Tritiated Compounds from Buried
Radioactive Wastes, presented at the International Symposium
on the Behavior of Tritium in the Environment, San Francisco,
Calif., October 16-20, 1978.
Ma78 Matuszek J. M., Wahlen M. U.. Kunz, C. 0., and Hutchinson
J. H., 1978. Is the Removal of 14C from Nuclear Reactor
Effluents a Plausible Regulatory Practice?, presented at a
conference on Effluent and Environmental Radiation Surveillance,
Atlanta, Ga.,July 9-14, 1978.
Mo77 Montgomery D. E., Kolde H. E., and Blanchard R. L., 1977,
Radiological Measurements at the Maxey Flats Radioactive Waste
Burial Site--1974-1975, EPA-520/5-76/020, January 1977, Eastern
Environmental Radiation Facility, Montgomery, Ala.
Pr76 Richard Prairie, Personal Communication.
-------
2.8-1
2.8 Plutonium Fuel Fabrication Facilities
2.8.1 General Description
The NRC currently licenses seven facilities within their
Plutonium Fuel and Fabrication (Special Nuclear Materials)
category. These facilities and their locations are shown in table
2.8-1. The missions of these facilities range from laboratory-scale
Plutonium fuel research to fabrication of light-water plutonium
recycle reactor fuel. In general, these facilities use processes
which are still under development with regard to the technology and
production of plutonium or plutonium-uranium mixed oxide fuel
elements. The overall objective is to recover plutonium and uranium
from depleted uranium process residues and fabricate these isotopes
as oxides into fuel pellets.
2.8.2 Process Description
Fuel pellets are fabricated from plutonium oxide and natural
uranium oxide powders which are received as feed materials
(ORNL75). Another possible feed material is aqueous plutonium
nitrate solution which is converted to plutonium oxide powder.
Scrap metal alloys bearing plutonium may also be used.
Plutonium-bearing solids are mixed with nitric acid to
dissolve the plutonium. Impure plutonium nitrate is then pumped
through ion exchange resin beds for selective sorption of the
plutonium. The plutonium on the resin column is eluted and
transferred to the product evaporator. The resin is reconditioned
and the scrap recovery cycle is repeated.
Plutonium Oxide Conversion Process
Plutonium is precipitated either as the peroxide or oxalate
which produces a filter cake. The filter cake is calcined
batch-wise in a muffle furnace to produce PuC^. The oxide is
milled, screened, and packaged for transfer to the fabrication
process.
Mixed Oxide Preparation
Mixed oxides in plutonium and uranium are prepared by
combining U02 with PuC>2 in a V-blender.
-------
2.8-2
Pellet Production
The mixed oxide blend is pressed, sintered, and
dry-centerless-ground into cylindrical pellets.
Fuel Rod Production
Pellets are put into fuel rods. Their subassembly is then
decontaminated and welded into a final fuel rod assembly.
Control Technology
Radiological control systems consist of single, double, or
triple-stage HEPA filtration on process stream ventilation systems
before there is release to the atmosphere.
2.8.3 Emissions of Radionuclides
Because plutonium is valuable, great care is taken to avoid
losing any of this material; filters from the HEPA systems are
returned to the scrap recovery portion of the process.
Differences in reporting by facilities preclude listing all
available data on emissions of radionuclides in one table. Some
facilities report gross alpha emissions only, while others provide
specific isotope emissions. Table 2.11-2 shows data which are
available on a comparable basis.
2.8.4. Model Plutonium Fuel Fabrication Facility
A model facility (table 2.8-3) was developed for assessing
the impact of airborne emissions from plutonium fuel fabrication
plants. The model parameters and source terms were taken from
Groenier, et al. (ORNL75) based on their Case 1 (base case, current
practice) data. The source terms used for the model are shown in
table 2.8-4.
-------
2.8-3
Table 2.8-1. Plutonium fuel fabrication facilities (NRC79)
i
Facility Location
Babcock and Wilcox Parks Township, Pa.
Battelle Memorial Institute West Jefferson, Ohio
Battelle P. N. Laboratories Richland, Wash.
Exxon Nuclear Co., Inc. Richland, Wash.
General Electric Co. Vallecitos, Calif.
Westinghouse Electric Corp. Cheswick, Pa.
Westinghouse Electric Corp. Waltz Mill, Pa.
Table 2.8-2. Atmospheric emissions of radionuclides from currently
licensed plutonium fuel fabrication facilities3
Facility
Exxon Nuclear
Battelle (Memorial)
Westinghouse
(Cheswick)
Release
interval
1976
1976
7-12/78
Plutonium emi
(CD
4.8E-3
6.3E-7
<1.4E-4
ssionsb
^Source: Facility reports to NRC.
"Contain plutonium-238, plutonium-239, plutonium-240,
plutonium-24l, and plutomum-242.
-------
2.8-4
Table 2.8-3. Model plutonium fuel fabrication facility
Parameter Value
Type of facility Plutonium fuel fabrication
Annual capacity 300 MT
Stack 10-meter fixed stack height
with no plume rise
Table 2.8-4. Atmospheric emissions of radionuclides from
the model plutonium fuel fabrication facility
Radionuclide Emissions
(Ci/y)
Uranium-234 l.OE-09
Uranium-235 3.3E-11
Uranium-238 7.1E-10
Plutonium-238 4.9E-05
Plutonium-239 2.6E-06
Plutonium-240 5.8E-06
Plutonium-241 1.3E-06
Plutonium-242 3.6E-08
Americium-241 4.3E-06
-------
2.8-5
2.8.5. Health Impact Assessment for Model Plutonium Fuel Fabrication
Facility
Annual radiation doses due to airborne radioactive emissions
from the model plutonium fuel fabrication facility are presented in
table 2.8-5. These estimates are for a facility located in the near
suburbs of a large Midwestern city (Site B with rural food source
fractions, Appendix A). The maximum individual dose equivalent rate
occurred 503 meters downwind of the site.
Table 2.8-6 estimates the individual lifetime fatal cancer
risks and committed fatal cancers to the regional population. The
lifetime fatal cancer risk to the highest exposed group of
individuals is approximately 1.1E-6. The lifetime fatal cancer risk
to the average individual in the region is estimated to be 6E-10.
The estimated number of fatal cancers per year of site
operation to the regional population is estimated to be 2.1E-5.
Table 2.8-5. Annual radiation doses from radioactive emissions from
the model plutonium fuel fabrication facility
Maximum Average
individual individual Population
Organ (mrem/y) (mrem/y) (person-rem/y)
Lung 1.7E-2 9.6E-6 2.4E-2
Bone 4.0E-1 2.2E-4 5.6E-1
Kidney 4.5E-2 2.5E-6 6.3E-2
Liver 6.1E-2 3.3E-6 8.3E-2
Thyroid 1.1E-2 6.4E-7 1.6E-2
G.I. Tract 9.5E-4 4.8E-7 1.2E-3
Other soft tissue 1.1E-2 6.4E-6 1.6E-2
-------
2.8-6
Table 2.8-6. Individual lifetime risks and population fatal cancers
due to radioactive emissions from the model plutonium fuel
fabrication facility
Individual lifetime risks Expected fatal cancers
Maximum Average per year of operation
Source individual individual (Fatal cancers)
Model
facility l.OE-6 5.6E-10 2.0E-5
-------
2.8-7
REFERENCES
NRC79 U.S. Nuclear Regulatory Commission, 1979, Program Summary
Report, NUREG 0380, Vol. 3, No. 3, March 16, 1979.
ORNL75 Oak Ridge National Laboratory, 1975, Correlation of
Radioactive Waste Treatment Costs and the Environmental Impact
of Waste Effluents in the Nuclear Fuel Cycle for Use in
Establishing "as low as Practicable" GuidesFabrication of
Light-Water Reactor Fuels Containing Plutonium, ORNL-TM-4904,
May 1975.
-------
2.9-1
2.9 Industrial Users and Other Categories
2.9.1 General Description
Table 2.9-1 lists the categories of industrial licenses and
the number of licenses issued in each category by the Nuclear Regula-
tory Commission (NRC78). Information from the NRC Office of
Standards Development (TC78) indicate that almost all of the
radiation sources covered by these licenses are sealed
(encapsulated) sources, i.e., the industrial application of the
radionuclide uses electromagnetic radiation from a sealed source.
The remaining sources can emit radionuclides into the air.
These include facilities which make static eliminators (nonsealed
alpha or beta emitters), self-illuminating devices (such as tritium
activated signs, watch dials), and facilities engaged in research
and development.
2.9.2. Process Description
The industrial uses of radionuclides are many; the following
are examples.
Radioisotope Gauges
A sealed radionuclide source is placed on one side of a
production line with a radiation detection device on the other
side. These gauges automatically measure and control the thickness
of steel sheet, paper, tire cords, textiles, plastics and other
materials. They may also control sugar, fat, and meat content in
food packaging; measure soil density and moisture content; and
measure snow density for water run-off forecasting.
These devices do not normally emit radionuclides into the air.
Static Eliminators
Electroplated alpha or beta sources are placed in close
proximity to production components prone to the build-up of static
electricity (such as moving conveyor belts). The ionizing
characteristics of the radiation reduce static charges. They are
used in production systems where static charges may degrade product
quality or pose safety hazards (inflammable or explosive
environments).
These devices do not normally emit radionuclides into the air.
-------
2.9-2
Table 2.9-1. NRC Industrial License Categories
Program area Licenses Licensees
Industrial :
Well logging
Other measuring systems*
Manufacturing, distribution
and service-broad
Manufacturing, distribution
and service-other
General license distribution
Exempt quantities
Exempt watches
Other exempt distribution
Nuclear laundry
Leak test
Waste disposal (Burial)
Waste disposal (other)
Radiography-single location
Radiography-multiple location
Power source
Irradiator (<10,000 curies)
Irradiator (>10,000 curies)
Research & development broad
Research & development other
Total
73
2,373
60
258
69
46
63
104
5
31
3
10
147
214
1
177
54
83
425
4,196
67
1,985
47
214
11
28
55
54
5
28
3
7
132
177
1
41
11
65
343
3,274
*0ther measuring systems include thickness gauges, level
indicators, densitometers, etc.
-------
2.9-3
Radiographic Devices
These devices use shielded sealed gamma sources as portable
radiography units for nondestructive testing. They are popular as
field quality control and inspection devices in such areas as
pipeline construction, building, and aircraft inspection to detect
flaws, voids, and other defects in materials and equipment. There
are no routine emissions of radionuclides into air.
Manufacturing and Research and Development Facilities
These facilities use radionuclides, sometimes in large
quantities, in manufacturing such product as self illuminating watch
dials using tritium, smoke detectors using americium-241, and static
illuminators. These are also many facilities which radionuclides in
research and development processes.
Emission data from this category of sources is not now
available.
2.9.3 Emission of Radionuclides
No data are available.
2.9.4. Model Facility
No model was developed, pending availability of release data.
2.9.5. Health Impact Assessment
An assessment will be made when emission data are available.
2.9.6. Other Licensed Categories
Other NRC license categories may be potential significant
sources of airborne emissions of radionuclides. These categories
include the academic license categories (university laboratories)
and byproduct material licenses for facilities performed in research
which are not included in the industrial license category by NRC.
Radionuclide emission data are not now available. Information on
these categories has been requested from NRC and an analysis of them
will be included in future reports.
-------
2.9-4
REFERENCES
NRC78 Memorandum from NRC, Office of Standards Development to
Robert S. Call is, EPA, dated November 28, 1978.
TC78 Telephone conversation with Lewis Battist, NRC, Environmental
Protection Standards Branch, November 27, 1978.
-------
CHAPTER 3
SOURCES OF EMISSIONS FROM DEPARTMENT OF ENERGY FACILITIES
Department of Energy (DOE) facilities support the development
of weapons in response to Department of Defense requirements. DOE
is also responsible for research programs on the biomedical, environ-
mental, physical and safety aspects of nuclear energy. Twenty-seven
prime contractor sites are discussed in this section. Data
presented here h^ve been obtained from reports by the contractor to
DOE during 1977.
The assessments for DOE facilities therefore were made on a
site-by-site basis. The diversity of operations at the different
sites makes it difficult to assess DOE facilities on a generic
basis. In general, the annual doses reported are 50-year dose
commitments, i.e., the dose delivered to an organ over a 50-year
period from one year's exposure. For most radionuclides, however,
the dose commitment is delivered in the f~' t year. Several sites
report 70-year dose commitments. These at- appropriately annotated
in the text. Additional information about the method by whic1" the
reported doses were calculated can be found in the references to
each section. The maximum exposed individual at DOE sites
represents either a hypothetical individual at the site boundary or
the nearest resident or group of individuals.
For the purpose of consistency of notation in the summary
table S-2, the 50-year or 70-year dose commitments are listed under
the annual dose rates; these are appropriately notated in the table
and in the individual site sections.
3.1 Hanford Site
3.1.1 General Description
The Hanford site was established in 1943 to produce plutonium
for nuclear weapons. Today, operations at the facilities on the
Hanford site include plutonium production, reactor fuel
reprocessing, waste management, reactor fuel manufacturing and
research and development.
Hanford site consists of 145,000 hectares of Government-owned
land in south-central Washington State. The site is about 320
kilometers east of Portland, Oregon, 270 kilometers southeast of
Seattle, Washington, and 200 kilometers southwest of Spokane,
Washington. The Columbia River flows through the northern area and
along the eastern boundary.
-------
3.1-2
The Hanford environs are categorized as a desert plain. The
rural, agricultural region is sparsely populatedonly 250,000
people live within 80 kilometers of the site. The nearest
population center includes the tri-city area of Richland, Pasco, and
Kenewich which have a combined population of 80,000. These cities
are southeast and downriver of the Hanford Site.
3.1.2 Process Description
100 Area
The 100 Area of the Hanford site, located to the north along
the Columbia River, is the location of the historic plutonium
production reactors. Eight reactors are deactivated and only the N
Reactor remains operationalthe principal source of radioactive
atmospheric releases in the area.
The exhaust air from the reactor building undergoes absolute
and activated charcoal filtration before being released. The stack
gases are monitored for gamma radiation and continuously sampled.
Radioactive atmospheric releases originating from the reactor heat
exchangers are not treated. However, the exhaust air is sampled
continuously. Argon-41 constitutes 95 percent of the radioactivity
released into the atmosphere from the 100 Area.
200 AREA
The 200 Area in the center of the reservation is the most
isolated area and the most removed from surface and subsurface
water. This area contains the fuel processing and waste-management
facilities and is subdivided into the 200 East and 200 West Areas.
200 EAST AREA
The major facilities located within the 200 East Area are the
Purex plant, B plant, Chemical Processing Plant, Critical Mass
Laboratory and the Tank Farm Facility. The Purex plant is
maintained in a standby status and is used only for plutonium
recovery when required by production from the N Reactor. The B
plant facility is used for waste fractionization, encapsulation and
storage. Cesium-137 and strontium-90 are removed from process
wastes, then doubly encapsulated and stored in retrievable
water-cooled storage. The Chemical Processing Plant is presently
shut down. At the Critical Mass Laboratory, research concerning
criticality safety of plutonium is conducted.
-------
3.1-3
High-level radioactive wastes, produced as a result of the
chemical reprocessing of irradiated Hanford reactor fuels, are
stored in large underground tanks. Operations at the 200 East Area
Tank Farm facilities involve evaporating nonboiling waste to salt
cake. There are three evaporators in operation, two of which are
permanently installed in tanks for in-tank solidification.
Most of the radioactive atmospheric effluents from the 200
East Area facilities undergo some form of HEPA filtration.
Effluents are either monitored or regularly sampled prior to
release. The 200 East Area facilities release small quantities of
gross alpha activity, assumed to be plutonium-239, and 0.12 curies
unidentified beta and gamma activity from mixed fission products,
which were assumed to be strontium-90.
200 WEST AREA
The U plant, Redox plant, T plant, Z plant and the Tank Farm
are the major facilities located within the 200 West Area. The U
plant is no longer used to recover uranium. However, the adjacent
uranium oxide plant produces powdered U03 by calcining uranyl
nitrate hexahydrate (U02(N03)26H20). The Redox fuel
processing plant is also shut down, only its analytical laboratory
remains in operation. The T plant is now used for various
decontamination projects, equipment repair, and plutonium storage.
The Z plant has only a plutonium laboratory which operates when
required by the N Reactor. Operations include processing and
preparation of plutonium materials for the N Reactor.
The Tank Farms and solidification facilities in the area
consist of 149 underground waste storage tanks that have capacities
for 190,000 to 3,800,000 liters of high-level liquid radioactive
waste, and two evaporators for the reduction of liquid waste to a
solid salt cake.
The radioactive atmospheric emissions from the 200 West Area,
like those from the East Area, undergo one or more stages of HEPA
filtration before being released. Effluent release points are
monitored by routine sampling. The 200 West Area emits small
amounts gross alpha activity, assumed to be plutonium-239, and
millicurie quantities of unidentified beta and gamma radioactivity,
assumed to be strontium-90.
-------
3.1-4
300 AREA
The 300 Area is located approximately two kilometers north of
the city of Richland, Washington, in the southeastern section of
Hanford site. Because the demand for production reactor fuel has
been reduced at Hanford, reactor fuel fabrication operations have
become secondary to the research and development activities in the
area.
The Pacific Northwest Laboratory (PNL) conducts research in
the physical, life, and environmental sciences, environmental
surveillance, and advanced methods of nuclear waste management. The
Hanford Engineering Development Laboratory (HEDL) is involved in
advanced reactor developments.
Each facility within the area has its own treatment system
for exhausted air. Radioactive particulates are removed from the
exhaust gases by various stages of HEPA filters located near the
hood, glovebox or cell where they are generated. Charcoal filters
are used where needed to absorb radioiodines. Radioactive
atmospheric releases from the 300 Area total less than 1 milicurie.
400 AREA
The 400 area of the Hanford site, the newest area to be
developed, is located just north of the 300 area, about 20
kilometers from Richland, Washington, and 7 kilometers from the west
bank of the Columbia river.
At the present time there are no sources emitting
radioactivity to the atmosphere. However, future sources of
radioactive atmospheric emissions include the Fast Flux Test
Facility and the High Performance Fuel Laboratory. Future
development in support of the Liquid Metal Fast Breeder Reactor
program call for a fuels and materials examination facility,
maintainance and storage facilities, fuel storage facility and
support facilities.
3.1.3 Emission of Radionuclides
Table 3.1-1 summarizes atmospheric emissions from the Hanford
site reported in the annual environmental surveillance report for
1977.
-------
3.1-5
Table 3.1-1. Atmospheric emissions of radionuclides,
Hanford Site, 1977 (Ho78)
Radionuclide
Tr i t i um
Sodium-21
Phosphorus-32
Argon-41
Chromium-51
Maganese-54
Maganese-56
Iron-59
Cobalt-58
Cobalt-60
Zinc- 65
Arsenic-76
Krypton-85m
Krypton-87
Krypton-
Rub idium-88
Strontium-89
Strontium-90
Strontium-91
Zirconium-95
Niobium- 95
Emissions
(Ci/y)
100 Area 200 Area 300 Area 400 Areaa
1.8E+1 - 9.0 1.6E+1
1.8E-1
_
1.3E+5
1.7E-2
1.6E-2
2.4 - -
1.8E-2
2.9E-3
2.9E-2 - &1.2E-4
4.5E-4
6.6E-1
8.3E+2
2.5E+3
2.8E+3
7.9E-3
1.7E-4 C2.1E-1 c2.5E-4
5.8E-1
3.8E-3
3.0E-3
Zirconium-
Niobium-97 2.1E-3
Molybodenum-
Technetium-99
Ruthenium-103
Ruthenium-106
Antimony-122
Antimony-124
Antimony-125
Tellurium-132
Iodine-129 2.1E-7
6.2E-1
8.2E-3
1.9E-2
5.4E-3
3.3E-3
4E-4
6E-3
1,
5.
See footnotes at the end of table
-------
3.1-6
Table 3.1-1. Atmospheric emissions, of radionucTides
Hanford Site, 1977--continued
Radionuclide
Iodine-131
Iodine- 132
Iodine- 133
Iodine- 135
Xenon-133
Xenon- 135
Cesium- 137
Cesium- 138
Barium-140
Lanthanum-140
Cerium-141
Cerium-
Praseodymium- 144
Neodymium-147
Samarium- 153
Europium- 154
Europium-155
Tungsten-187
Uranium-238
Neptunium-239
Plutonium-238
Plutonium-239
Plutonium-240
Americium-241
Plutonium-241
Plutonium-242
Curium- 244
100 Area
5.5E-1
9.6
4.0
8.6
6.8E+2
3.4E+3
1.5E-3
1.3E+4
2.0E-1
3.6E-1
1.3E-3
3.0
1.3E-2
1.7E-3
l.OE-2
6.2E-3
6.9E-2
-
3.2E-3
1.1E-6
5.7E-5
_
4.0E-3
-
_
-
Emissions
(Ci/y)
200 Area 300 Area 400 Areaa
4.4E-4
_
_
_
_
_
_ _
_
_ _
_ _
mm
_
-
_
_
_
- -
5.2E-5
_
1.2E-5
d2.8E-3 d3.2E-5 1.1E-6
2.2E-6
1.4E-6
1.4E-6
5.3E-4
8.4E-8 1.1E-8
aEstimated from available environmental statements.
bActually reported as mixed activation products. Cobalt-60 was
assumed to be the radionuclide for dose calculations.
GActually reported as mixed fission products. Strontium-90 was
assumed to be the radionuclide for dose calculations.
^Actually reported as gross alpha. Plutonium-239 was assumed to be
the radionuclide for dose calculations.
-------
3.1-7
3.1.4 Health Impact Assessment at Hanford Site.
The maximum site boundary dose from 1977 atmospheric
emissions at the Hanford site occurred along the northwest
boundary. The annual total body dose to an individual at the site
boundary was reported to be 0.14 mrem. The principal radionuclides
contributing to this dose were cesium-138 and argon-41 from N
reactor operations in the 100 Area
Table 3.1-2 estimates individual doses resulting from one
year of operations and following 50 years of operations at the
site. The estimated maximum individual dose would be received at a
location 1.6 kilometers east of the 300 Area and was .06 mrem to the
thyroid. The 50-year dose commitment to the maximum exposed
individual was .07 mrem to the bone.
Because the population affected by the emissions from each
area differs greatly, the 80-kilometer radius population dose was
calculated for each area. The annual 50-year dose commitments are
reported in table 3.1-3. The highest population dose was estimated
to be 4.1 person-rems to the bone from N reactor operations in the
100 Area following 50 years of operations.
Table 3.1-4 estimates the individual lifetime risks and the
number of fatal cancers resulting from these doses. The lifetime
cancer risk to the maximum individual is estimated to be 2.7 x
10-'. jhg estimated number of fatal cancers per year of Hanford
operation to the population within 80 kilometers is estimated to be
about 6.5 x 10~4.
-------
3.1-8
Table 3.1-2. Doses from radioactive emissions,
Hanford Site, 1977 (Ho78)
Maximum individuals
Organ (mrem)
50-year dose
Annual commitment
Lung
Bone
Thyroid
G.I. Tract
Skin
Total body
.03
.02
.06
.03
.03
.03
.03
.07
.06
.03
.03
.03
aThe maximum individual is located 1.6 km east of 100 Area,
Table 3.1-3.
Population doses from radioactive emissions,
Hanford Site, 1977 (Ho78)
Organ
Population dose (person-rem)'
Annual
100-N 200 300
Area Area Area
50-year Commitment
100 200 300
Area Area Area
Lung
Bone
Thyroid
G.I. Tract
Total body
2.1
2.0
4.0
1.9
2.0
.02
.02
.01
.02
.01
.01
.01
.03
.01
.01
2.1
2.4
4.1
1.9
2.0
.06
1.5
.01
.02
.16
.02
.15
.03
.01
.02
aPopulation within 80 kilometers:
Area, 258,000; 300 Area, 171,000.
100-N Area, 236,000; 200
-------
3.1-9
Table 3.1-4. Individual lifetime risks and number of fatal cancers
from radioactive emissions, Hanford Site, 1977
Individual lifetime risks Expected fatal cancers
Organ Maximum individual3 per year of operation^
(Fatal cancers)
Lung
Bone
Thyroid
G.I. Tract
Other soft tissue
8.4E-8
2.8E-8
4.2E-9
4.2E-8
1.1E-7
8.5E-5
8.1E-5
4.0E-6
3.9E-5
4.4E-4
Total 2.7E-7 6.5E-4
aThe maximum individual is located 1.6 km east of 300 Area.
bThe population is within 80 km.
-------
3.1-10
REFERENCES
AEC72 Atomic Energy Commission, 1972, Environmental Statement,
WASH-1510, Fast Flux Test Facility, Richland, Washington.
AEC74 Atomic Energy Commission, 1974 Draft Environmental Statement,
WASH-1538, Waste Management Operations, Hanford Reservation,
Richland, Washington.
ERDA77a Energy Research & Development Administration, 1977 Draft
Environmental Impact Statement, ERDA 1556-D, High Flux Neutron
Source Facility, Hanford Reservation, Richland, Washington.
ERDA77b Energy Research and Development Administration, 1977 Final
Environmental Impact Statement, ERDA-1550, High Performance Fuel
Laboratory, Hanford Reservation, Richland, Washington.
Ho78 Houston J. R. and Blumer P. J., 1978, Environmental
Surveillance at Hanford for CY-1977, PNL-2614, Battelle Pacific
Northwest Laboratories, Richland, Washington
-------
3.2-1
3.2 Savannah River Plant
3.2.1 General Description
Savannah River Plant (SRP) facilities were established in the
early 1950's to produce nuclear materials for national defense
requirements, principally plutonium and tritium. The site is 150
kilometers from Savannah, Georgia, on a 78,000-hectare tract along
the Savannah River in Barnwell, Allendale, and Aiken Counties, South
Carolina. The population density varies from 4 to over 154 people
per square kilometer in the surrounding area, which is predominantly
rural with forested countryside and diversified farming.
Major operating facilities include three nuclear reactors,
two chemical separations plants, a fuel and target fabrication
plant, a heavy water production plant, and the Savannah River
Laboratory. These facilities, which are the principal sources of
radioactive airborne effluents, are centrally located on the site,
while the administration and other non-nuclear support facilities
are located nearer the site periphery.
3.2.2 Process Description
Nuclear Reactors
Three operating nuclear reactors (C, K, and P) produce
plutonium and tritium for nuclear weapons. Two additional reactors
have been shut down and placed in standby operational status. All
five reactors are fueled with uranium and are moderated with heavy
water. During 1977, the threexoperating reactors released large
amounts of tritium, argon-41, krypton and xenon gases, and carbon-14
to the atmosphere. At each reactor the radioactive atmospheric
effluents are treated by HEPA and charcoal filtration before being
released from a 61-meter stack. This filtration system removes only
radioactive particles and iodines and has no effect on the noble
gases and tritium (which is released as tritiated water vapor).
Chemical Separations Facilities
The chemical separations facilities consist of two separate
areas, the F- and H-areas, which process irradiated fuel and uranium
targets. In the F-area, plutonium-239 is recovered using the Purex
solvent extraction process. The F-area also contains the main
analytical laboratory, the plutonium metallurgical laboratory, and
the Plutonium Fuel Form Facility.
-------
3.2-2
The H-area canyon building is similar to the F-area
separations building. Special radionuclides and uranium-235 are
recovered using a solvent extraction process like the Purex process,
but with modifications depending on the nature of the fuels and the
radionuclide to be recovered. The H-area also contains the Tritium
Processing Building, the Receiving Basin for Offsite Fuels and the
Resin Regeneration Facility.
The major releases from the separations area include tritium,
krypton-85, carbon-14, xenon gases, and other fission products. The
primary sources of radioactivity in the effluent are from the main
process canyon air, the vessel vent system off-gases, and the
Plutonium powder-handling cabinets. The effluent is passed through
a deep-bed sand filter and is then released. Ventilation air from
the separations area may undergo several stages of filtration for
removal of particulate radioactivity prior to release.
The F-area main processing, or canyon building, is served by
a 61-meter stack, which releases the combined filtered effluents
from several sections of the building. Similarly, the H-area
processing canyon building is serviced by a 61-meter stack and the
main process canyon air and the process vessel vent off-gases are
routed through a deep-bed sand filter. Both stacks are continuously
monitored.
Other sources of releases of radioactivity to the atmosphere
from the H-area include: (1) three tritium facilities which are
serviced by three 61-meter stacks and one 23-meter stack; (2) the
receiving basin for off-site Fuels and Resin Regeneration Facility
which releases fission product gases from shipping casks and fuel
cutting and cleaning operationsthese effluents are released from a
short stack, 16 meters above grade; and (3) the ventilation air from
the metallurgical building and analytical laboratory which is
filtered through HEPA filters and released from 23-meter stacks.
Fuel and Target Fabrication Facilities
The Fuel and Target Fabrication Facilities make enriched
uranium-aluminum alloy fuel and canned depleted uranium metal
targets for the Savannah River Plant reactors. During 1977, the
three major buildings in this area released small amounts of natural
uranium and unidentified alpha activity. Off-gas exhausts from the
various operations and laboratories are filtered through HEPA
filters, monitored, and then released from 15-meter, 23-meter, and
several 30-meter stacks.
-------
3.2-3
Heavy Water Production Facility
The heavy water production and recovery facilities emit
tritium into the atmosphere. Evaporated tritiated heavy water from
the rework unit pump room and drum handling building is vented to a
21-meter stack. Radioactivitiy releases are monitored
continuously. Monthly estimates of atmospheric releases from the
drum cleaning facility and the analytical laboratory are based on
periodic grab samples.
Savannah River Laboratory
The Savannah River Laboratory, which is responsible for
research and development activities at SRP, releases small amounts
of tritium, cobalt-60, -and iodine-131. HEPA and charcoal filtration
treat the radioactive effluents which are then monitored and
released from eight stacks, 6 to 23 meters high.
3.2.3. Emission of Radionuclides
Emissions from the Savannah River Plant in 1977 are given in
Table 3.2-1.
-------
3.2-4
Table 3.2-1. Atmospheric emissions from the Savannah
River Plant, 1977 (DOE77)
Emissions
Facility and Radionuclide (Ci/y)
C Reactor Plant
Argon-41 3.1E+4
Carbon-14 1.2E+1
Tritium 8.3E+4
Iodine-131 4.9E-3
Krypton-85m 1.8E+2
Krypton-87 2.2E+2
Krypton-88 3.3E+2
Unidentified Alpha 2.6E-6
Unidentified
Beta and Gamma 1.8E-4
Xenon-133 1.2E+3
Xenon-135 7.8E+2
K Reactor Plant
Argon-41 1.6E+4
Carbon-14 1.3E+1
Tritium 4.2E+4
Iodine-131 3.3E-4
Krypton-85m 4.9E+1
Krypton-87 6.4E+1
Krypton-88 7.2E+1
Unidentified Alpha 1.9E-6
Unidentified
Beta 1.8E-4
Xenon-133 2.4E+2
Xenon-135 1.1E+2
P Reactor Plant
Argon-41 1.8E+4
Carbon-14 1.2E+1
Tritium 3.9E+4
Iodine-131 1.2E-3
Krypton-85m 6.1E+2
Krypton-87 3.1E+2
Krypton-88 2.7E+2
Unidentified Alpha 2.4E-6
-------
3.2-5
Table 3.2-1. Atmospheric emissions from the Savannah River
Plant, 1977 (DOE77)continued
Emissions
Facility and Radionuclide (Ci/y)
P Reactor Plantcontinued
Unidentified
Beta and Gamma 1.1E-4
Xenon-133 9.0E+2
Xenon-135 6.5E+2
200 Separations Area
Tritium 1.9E+5
Iodine-129 1.4E-1
Krypton-85 4.4E+5
Xenon-131m 1.2E+1
Xenon-133 1.3E-2
F-Area Separations Plant
Americium-241 1.5E-4
Carbon-14 1.3E+1
Cerium-141 2.1E-4
Cerium-144 4.8E-3
Curium-244 1.5E-4
Cesium-134 2.0E-6
Cesium-137 5.6E-4
Iodine-131 3.8E-2
Niobium-95 3.3E-2
Plutonium-238 2.8E-4
Plutonium-239 l.OE-4
Ruthenium-103 1.5E-3
Ruthenium-106 l.OE-2
Strontium-89
and Strontium-90 2.0E-3
Uranium-238 1.2E-3
Unidentified
Beta and Gamma 1.6E-5
Ziconium-95 7.6E-3
-------
3.2-6
Table 3.2-1. Atmospheric emissions from the Savannah River
Plant, 1977 (DOE77)continued
Emissions
Facility and Radionuclide (Ci/y)
H-Area Separations Plant
Americium-241 1.9E-4
Carbon-14 1.3E+1
Cerium-141 1.4E-4
Cerium-144 l.OE-2
Curium-244 1.9E-4
Cesium-134 4.3E-4
Cesium-137 1.3E-3
Iodine-131 1.2E-2
Niobium-95 4.4E-3
Plutonium-238 4.7E-3
Ruthenium-103 2.3E-2
Ruthenium-106 1.3E-1
Strontium-89 and
Strontium-90 2.2E-3
Uranium-238 3.2E-4
Unidentified
Beta & Gamma 4.0E-6
Zirconium-95 6.8E-3
Fuel and Target Fabrication Plant
Uranium-238 1.8E-5
Unidentifed
Alpha 1.7E-5
Heavy Water Production Plant
Tritium 2.9E+3
Savannah River Laboratory
Cobalt-60 3.8E-4
Tritium 4.0
Iodine-131 5.0E-3
Unidentified Alpha l.OE-6
Unidentified
Beta & Gamma 7.0E-6
-------
3.2-7
3.2.4 Dose Assessment of the Savannah River Plant
The 70-year dose commitment from a one-year release of
radioactive materials from the Savannah River Plant, based on 1977
radioactive atmospheric releases, was calculated for the maximum
exposed individual at the site boundary and for the population
within 80 kilometers. The major contributors to the maximum exposed
individual whole-body dose at the perimeter were tritium, 78
percent; argon-41, 15 percent; carbon-14, 5 percent; and from
krypton, xenon, iodine-129, iodine-131, and other radionuclides, 2
percent. The population dose commitment calculated for the
population within an 80-kilometer radius was 114 person-rem, with
tritium accounting for 86 percent; argon-41, 7 percent; and
carbon-14, 6 percent.
Table 3.2-2 summarizes the contribution of each nuclide to
the maximum site boundary dose and the 80-kilometer population
dose. The doses shown are the 70 year lifetime dose commitment
based on 1977 atmospheric emissions.
Table 3.2-3 estimates the individual lifetime risks and the
number of fatal cancers resulting from these doses. The lifetime
cancer risk to the maximum individual is estimated to be 1.5E-5.
The estimated number of fatal cancers per year of SRP operation to
the population within 80 kilometers is estimated to be about 2.2E-2.
-------
3.2-8
Table 3.2-2. Site boundary and population dose commitments3
from atmospheric emissions at Savannah River Plant, 1977 (DP78)
Radionuclide
Tr i t i urn
Carbon-14
Argon-41
Krypton-85m
Krypton-85
Krypton-87
Krypton-88
Xenon-131m
Xenon-133
Xenon-135
Iodine-129
Iodine-131
Cobalt-60
Strontium-89
and Strontium-90
Zirconium-95
Niobium-95
Ruthenium-103
Ruthenium-106
Cesium-134
Cesium-137
Cerium- 141
Cerium-144
Uranium
Plutonium-238
Plutonium-239
Americium-241
Curium-244
Maximum
individual
(mrem)
8.5E-1
6.0E-2
1.8E-1
4.9E-4
2.2E-3
1.7E-3
4.3E-3
l.OE-5
8.6E-4
1.5E-3
1.3E-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
l.OE-5
5.3E-4
l.OE-5
l.OE-5
l.OE-5
Totals 1.1
Population
(person-rem)
9.8E+1
6.9
8.3
3.2E-2
2.7E-1
6.4E-2
2.4E-2
8.0E-4
8.2E-2
1.1E-1
l.OE-1
1.2E-3
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
l.OE-4
3.6E-2
8.0E-4
8.0E-4
8.0E-4
1.1E+2
aA 70-year dose commitment based on 1977 atmospheric emissions
of radioactive materials. The maximum exposed individual is at
the site boundary; the population is within 80 kilometers of SRP.
-------
3.2-9
Table 3.2-3. Individual lifetime risks and number of fatal cancers3
from radioactive emissions, Savannah River Plant, 1977
Source
Individual
Maximum
individual^
lifetime risks
Average individual
Region
Expected fatal cancers
per year of operation
(Fatal cancers)
SRP
1.5E-5
3.3E-6
2.2E-2
aTo the population within 80 kilometers.
bThe maximum individual is at the site boundary.
-------
3.2-10
REFERENCES
DOE77 Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for 1977, EIS-51,
Washington, D.C.
ERDA77 Energy Research & Development Administration, 1977, Final
Environmental Impact Statement, Waste Management Operations,
Savannah River Plant, Aiken, South Carolina, ERDA-1537, UC-2-11-70,
DP78 E.I. du Pont de Nemours & Company, 1978 Environmental Monitoring
in the Vicinity of the Savannah River Plant, Annual Report for 1977,
DPSU 78-30-1.
-------
3.3-1
3.3 Idaho National Engineering Laboratory
3.3.1 General Description
The Idaho National Engineering Laboratory (INEL) was
established in 1949 for operating and testing nuclear reactors and
critical facilities. Today, a broad scope of engineering activities
are conducted there. Seventeen reactors and critical facilities
support the following programs: naval propulsion, fast breeder
development, lightwater reactor safety testing, organic moderator
and coolant development, materials testing, portable military power
space development, decontamination and decommissioning, spent fuel
reprocessing, waste management and miscellaneous research.
The INEL site, occupying more than 230 thousand hectares on
the Upper Snake River Plain in southeastern Idaho, is 48 kilometers
west of Idaho Falls, Idaho. The surrounding area is a grazing range
for livestock with some irrigation farming to the north and
northeast. The total population within 80 kilometers of INEL is
approximately 94,000.
3.3.2 Process Description
There are currently four primary contractors which operate
the facilities for the Department of Energy: EG & G Idaho, Allied
Chemical Corporation, Westinghouse Electric Corporation, and Argonne
National Laboratory.
EG & G Idaho Facilities
EG & G Idaho operates the Test Reactor Area, the Test Area
North, the Loss-of-Fluid-Test Facility, the Power Burst Facility,
the Auxilary Reactor Area, the INEL Radioactive Waste Management
Complex and the Central Facilities Area. The major programs
conducted by EG & G Idaho include test irradiation and
light-water-cooled reactor safety testing and research.
Test Reactor Area
The Test Reactor Area, in the south central part of INEL,
provides facilities for studying the performance of reactor
materials and equipment components under high neutron flux
conditions. The reactor area contains three reactors, the Materials
Testing Reactor (MTR), the Engineering Test Reactor (ETR) and the
Advanced Test
-------
3.3-2
Reactor (ATR). Also, there are three low-power reactors located in
this area, the Engineering Test Reactor Critical facility, the
Advanced Test Reactor Critical facility and the Advanced Reactivity
Measurement Facility.
The MTR is in a standby status. Exhaust air from these
laboratories is treated by caustic scrubbing and HEPA filtration.
Exhausted air is released from a 26-meter stack after being
monitored. No detectable release of radioactivity has been observed
from the MTR since its shutdown.
Ventilation air from the ATR and ETR is discharged from
76-meter stacks at each building. No waste treatment system is
employed at either of these facilities. The stacks are monitored
continuously for both gaseous and particulate activity. Large
amounts of argon-41 and radioactive isotopes of krypton and xenon
are released from these facilities.
Test Area North (TAN)
The Technical Support Facilities (TSF) occupy 28 hectares and
consist of about 20 buildings. The TSF supports water reactor
safety programs, particularly the Loss-of-Fluid Test activities.
Primary sources of radioactive airborne releases from the TSF area
include hot shops area, radioactive material storage pool, warm
shop, cleaning rooms and various laboratories.
Each area that has a potential for generating radioactive
airborne waste has its own effluent control system which includes
roughing and HEPA filters. No provisions have been made for removal
of radioiodines or for monitoring radioactive gaseous wastes such as
isotopes of krypton and xenon and argon-41. The effluents are
monitored for alpha and beta activity. Small amounts of fission
products constitute typical releases from the TSF area.
The Low Power Test (LPT) and Experimental Beryllium Oxide
Reactor (EBOR) facilities are also located in the TAN area. There
is little potential for radioactive airborne releases from these
facilities because of minimal reactor operations at the LPT and only
non-nuclear testing conducted at the EBOR. However the exhaust
systems at each facility are equipped with prefliters and absolute
HEPA filters.
Loss-of-Fluid Test (LOFT) Facility
The LOFT facility, located in the TAN complex, is the only
reactor in operation in this area. Radioactive airborne effluents
-------
3.3-3
from reactor operations undergo several stages of filtration which
include roughing and HEPA filters. Charcoal absorbers are used to
remove halogens. Airborne effluents are exhausted from a 46-meter
stack which is monitored for gases and particulates. No radioactive
airborne effluents were reported in 1977.
Power Burst Facility
The Power Burst Facility is a high-performance, water-cooled,
uranium oxide fueled reactor used in support of the light-water
reactor safety testing program. Airborne effluents are vented
through roughing and HEPA filters. Silver zeolite absorbers are
used for iodine removal. Ventilation air is then released from a
24-meter stack equipped with a gas monitor and particulate sampler.
Small quantities of activation and fission products are released.
Auxiliary Reactor Area (ARA)
The ARA area is located to the southeast of the PBF in the
south-central section of the INEL site. This was the location of
the Army reactors which have been shutdown. Only the ARA-1 hot
cells remain as a source of radioactive airborne releases. The
exhaust system serving the ARA-1 hot cells is equipped with roughing
and HEPA filters. One cell utilizes charcoal filters for iodine
removal. The exhaust stack is continuously monitored for iodine,
particulate, alpha, and gaseous activity. The annual release of
mixed activation and fission products is very small.
Radioactive Waste Management Complex
Located in the southwest corner of the INEL site, this
complex disposes solid low-level radioactive waste materials
generated by laboratory operations. There is a subsurface disposal
area used for burial of nontransuranic solid wastes. Transuranic
wastes in containers having less than 10 nCi/g of activity are
stacked on an asphalt pad and covered with earth in the separate
Transuranic Disposal Area. A third area, the Transuranic Storage
Area, provides interim storage for containers having greater than 10
nCi/g of transuranic acitivity. Operations in this area are
monitored to detect any release of hazardous materials. There are
no routine releases reported.
Central Facilities Area
This area, located in the south-central portion of the site,
provides support services for outlying operational areas. The only
-------
3.3-4
potential source for airborne radioactivity release is from the
laundry facility. The off-gas vent from the dryer is equipped with
screens and filters to prevent the spread of contaminated lint. The
area is surveyed frequently.
Allied Chemical Corporation
The Allied Chemical Corporation operates the Idaho Chemical
Processing Plant (ICPP), the Waste Calcining Facility (WCF), the
Fuel Storage Basin Facility and the Tank Farm Facility. Of these
facilities, the ICPP and the WCF are the primary sources of airborne
radioactivity.
At the ICPP, enriched uranium is recovered from spent nuclear
fuel by a solvent extraction process. The major sources of
radioactive airborne effluent are the process dissolvers, process
vessels and areas, analytical facilities, sample stations, waste
solvent burners and process ventilation air. Process off-gas
streams from the dissolvers are routed through a reflux condenser,
entrainment separator, demister, superheater and HEPA filters before
being released from the stack. Two of the three dissolver off-gas
streams can be sent to the Rare Gas Plant where radioactive xenon
and krypton can be recovered by a cryogenic distillation process.
The remaining off-gases are collected, then routed through a
demister, superheater and HEPA filter prior to discharge. Process
sampling stations exhaust to the stack through fiber glass filters.
The waste solvent burner exhaust and process ventilation air flow
directly to the stack without any treatment.
The Waste Calcining Facility (WCF) solidifies liquid waste
generated by the ICPP. This process results in significant amounts
of airborne activity. The WCF employs an extensive cleanup system
to remove particulates and volatile fission products. The cleanup
system includes a cyclone, scrubbing system, silica gel absorber
beds and HEPA filters. Radiation detectors monitor the process
off-gases which are then exhausted to the ICPP stack.
The 76-meter ICPP stack is monitored for gross activity, and
radiochemical analyses are conducted on daily filter samples. From
these measurements, the total activity in the emissions from the
stack are estimated. In 1977, krypton-85 and tritium were the
principal nuclides contributing to the total activity released.
Westinghouse Electric Corporation
Westinghouse Electric Corporation operates the Naval Reactor
Facility (NRF), which consists of three operating naval reactor
-------
3.3-5
prototypes (S1W, AlW and S5G) and the Expended Core Facility (ECF).
The NRF area is located in the south-central area just north of the
Test Reactor Area.
The AIM reactor plant radioactive waste system building, the
S5G reactor building and the ECF are equipped with HEPA filters in
the exhaust systems. The principal nuclides released from the NRF
complex were several curies of krypton-85 and xenon-133 which
originated from the ECF; very small quantities of mixed fission and
activation products were released from the reactor buildings.
Argonne National Laboratory-West (ANL-W)
The Argonne National Laboratory-West facilities are located
in the southeastern portion of the INEL site. There are five major
complexes which comprise ANL-W; the Experimental Breeder Reactor
No. 2, the Transient Reactor Test Facility, the Zero Power Plutonium
Reactor, the Hot Fuel Examination Facility, and the Laboratory and
Office and support complex.
Experimental Breeder Reactor No. 2 (EBR-II)
The EBR-II is an experimental liquid-metal cooled fast
breeder reactor used to irradiate samples of reactor fuels and
structural material for the LMFBR development program. Associated
with the EBR-II are the sodium component cleanup facility and a fuel
assembly and storage building.
The primary tank of the reactor is filled with an argon cover
gas. Cover gas that leaks into the containment building is
withdrawn along with the ventilation air through HEPA filters. The
argon covergas can be purged directly to the atmosphere, bypassing
the containment building and the HEPA Filters. The containment
building discharge may be initiated by the gamma monitoring system.
Radioactive airborne effluents from the EBR-II consist
primarily of fission product noble gases, which consist of isotopes
of xenon, krypton and argon-41.
The sodium component cleanup facility releases small amounts
of radioactivity. Airborne effluents are treated by HEPA filtration
before release from a local stack. Monthly filter samples are used
to determine gross alpha and beta activity released.
-------
3.3-6
The Fuel Assembly and Storage Building (FASB) airborne
effluents are treated by HEPA filtration. Airborne effluents
resulting from fuel assembly operations are monitored by monthly
filter samples for gross alpha, beta and gamma activity.
Transient Reactor Test Facility (TREAT)
The TREAT facility reactor produces short extreme pulses of
nuclear energy for melt-down studies of prototype and experimental
fuel elements. Airborne effluents from the reactor operations are
exhausted through HEPA filters and an 18-meter stack. The exhaust
air is continuously monitored and cold trap samples are obtained
periodically. The TREAT radioactive gaseous effluent consists of
isotopes of krypton and xenon and argon-41.
Zero Power Plutonium Reactor (ZPPR)
The ZPPR experiments provide information about reactor
physics in support of the plutonium fuel fast breeder reactor
program. During reactor operations the cooling air is recirculated
through HEPA filters. When the ZPPR is not in operation, the
cooling air is exhausted through HEPA filters and an 18-meter
stack. The cooling air is continuously monitored both upstream and
downstream of the HEPA filters. The radioactive atmospheric
effluents from the ZPPR are composed primarily of noble gas fission
product isotopes of krypton and xenon.
Hot Fuels Examination Facility (HFEF)
The Hot Fuels Examination Facility consists of two separate
buildings. The HFEF-South is used mainly as an irradiation
subassembly-disassembly, inspection and assembly point and the
HFEF-North, for diagnostics and inspections.
The HFEF-South contains two cells used for processing
operations. The facility is used for examination of materials and
fuels irradiated in the EBR-II and TREAT. Ventilation air from this
facility is treated by HEPA filters and exhausted from the 61-meter
EBR-II stack. There are two work areas used for cleanup that have
separate exhaust and multiple HEPA filtration systems.
The HFEF-North is used for interim and final examination of
fast reactor fuel and structural specimens. The main cell has an
argon atmosphere. The ventilation system passes all radioactive
airborne effluents through HEPA filters. The exhaust is
-------
3.3-7
continuously monitored for alpha and beta activity and selectively
for cesium-137, iodine-131, xenon-133 and krypton-85. Only small
releases were detected in 1977.
Laboratory and Office Complex
Potential sources of radioactive airborne emissions are from
radiochemical hoods in the several laboratories of the Laboratory
and Office Complex. The sodium chemistry laboratory is the only
building in this complex using its own short stack. Exhaust air
from all the other laboratories undergoes HEPA filtration before
release from the main EBR-II stack.
3.3.3 Emissions of Radionuclides
Table 3.3.1 summarizes the radioactive airborne emissions
from the INEL facilities in 1977.
-------
3.3-8
Table 3.3-1. Atmospheric emissions of radionuclides from
the Idaho National Engineering Laboratory Site, 1977
Source
Radionuclide
Emissions
(Ci/y)
EG & G IDAHO
Test Reactor Area
Test Area North
Power Burst Facility
Argon-41
Barium-139
Cesium-138
Krypton-85m
Krypton-87
Krypton-88
Rubidium-88
Xenon-133
Xenon-135
Xenon-135m
Xenon-138
Cerium-144
Cobalt-58
Cobalt-60
Cesium-134
Cesium-137
Iodine-131
Lanthanum-140
Manganese-54
Ruthenium-103
Strontium-90
Unidentified
Alpha
Unidentiffied
Beta & Gamma
Silver-llOm
Barium-140
Cerium-141
Cerium-144
Cobalt-58
Cobalt-60
Chromium-51
3.2E+3
4.9Ef2
7.3E+1
1.1E+3
4.8E+3
3.4E+3
4.3EH
3.1E+2
9.6E+1
3.6E+3
1.1E+4
l.OE-7
5.1E-6
2.5E-5
6.2E-7
9.7E-5
8.9E-7
6.7E-9
8.7E-6
5.9E-7
5.7E-7
5.9E-6
7.5E-5
6.3E-7
9.8E-7
9.3E-7
1.3E-6
1.5E-7
4.7E-6
2.4E-6
-------
3.3-9
Table 3.3-1. Atmospheric emissions of radionuclides
from the Idaho National Engineering Laboratory Site, 1977continued
Source Radionuclide
Power Burst Facilitycontinued
Cesium-134
Cesium-137
Iron-59
Mercury-203
Iodine-131
Lanthanum-140
Manganese-54
Niobium-95
Antimony-125
Strontium-90
Unidentified
Alpha
Unidentified
Beta & Gamma
Tungsten-187
Zircon ium-95
Auxilary Reactor Area Cerium-141
Cerium-144
Cobalt-58
Cobalt-60
Cesium-134
Cesium-137
Europium-152
Iodine-131
Manganese-54
Niobium-95
Strontium-90
Unidentified
Alpha
Unidentified
Beta & Gamma
Emissions
(Ci/y)
1.7E-6
3.5E-5
1.1E-6
1.7E-6
9.7E-6
5.1E-6
3.0E-6
1.1E-6
3.5E-7
2.1E-6
2.2E-7
2.5E-5
9.4E-7
2.4E-7
2.0E-9
1.6E-8
1.8E-8
1.9E-7
l.OE-8
3.2E-7
2.8E-7
1.3E-6
6.1E-8
3.1E-8
1.6E-7
6.1E-9
4.2E-7
-------
3.3-10
Table 3.3-1. Atmospheric emissions of radionuclides
from the Idaho National Engineering Laboratory Site, 1977--continued
Source
Allied Chemical Corporation
Idaho Chemical Processing
Plant (ICPP) and the Waste
Calcining Facility (WCF)
Westinghouse Electric Corp.
S1W Reactor Plant
Radionuclide
Cerium- 144
Cobalt-60
Cesium- 134
Cesium- 137
Europium-154
Tritium
Krypton-85
Manganese-54
Niobium-95
Praseodymium-144
Plutonium-238
Plutonium-239
Plutonium-240
Rhodium-106
Rubidium-106
Antimony-125
Strontium-90
Thorium- 2 32
Yttrium-90
Zirconium-
Niobium- 95
Zirconium-95
--Naval Reactor Faci
Argon-41
Cobalt-60
Cesium-137
Iodine-131
Iodine-132
Iodine-133
Iodine-134
Iodine-135
Krypton-85
Krypton-88
Emissions
(Ci/y)
1.02E-1
3.00E-5
6.30E-3
2.03E-1
1.51E-4
3.08E+3
1.08E+5
3.73E-5
2.96E-3
1.02E-1
9.93E-4
3.57E-4
6.30E-5
1.39E-1
1.39E-1
3.44E-2
6.16E-2
2.64E-5
6.16E-2
2.00E-3
1.71E-3
lity
3.9E-4
2.6E-6
2.4E-6
2.3E-7
8.5E-8
3.9E-7
7.1E-8
2.5E-7
7.5E-5
2.5E-5
-------
3.3-11
Table 3.3-1. Atmospheric emissions of radionuclides
from the Idaho Engineering Laboratory Site, 1977--continued
Emissions
Source Radionuclide (Ci/y)
Westinghouse Electric Corp.--Naval Reactor Facilitycontinued
S1W Reactor Plantcontinued
Unidentified
Beta & Gamma 5.8E-6
Xenon-133 1.7E-3
Xenon-138 4.1E-6
A1W Reactor Plant and Radioactive
Waste System Building
Argon-41 1.3E-3
Cobalt-60 1.8E-5
Iodine-131 4.5E-7
Iodine-132 1.9E-7
Iodine-133 8.9E-7
Iodine-134 1.7E-7
Iodine-135 5.7E-7
Krypton-85 1.2E-4
Krypton-88 5.7E-5
Unidentified
Beta & Gamma 3.3E-6
Xenon-133 3.4E-3
Xenon-138 9.7E-6
S5G Reactor Plant
Argon-41 1.2E-3
Cobalt-60 5.4E-7
Iodine-131 2.0E-7
Iodine-132 7.6E-8
Iodine-133 3.6E-7
Iodine-134 6.9E-8
Iodine-135 2.3E-7
Krypton-85 3.0E-5
Krypton-88 2.3E-5
Xenon-133 1.5E-3
Xenon-138 3.9E-6
-------
3.3-12
Table 3.3-1. Atmospheric emissions of radionuclides
from the Idaho Engineering Laboratory Site, 1977continued
Source
Expended Core Facility
Radionuclide
Cobalt-60
Tritium
Iodine-131
Krypton-85
Xenon-133
Emissions
(Ci/y)
2.4E-6
4.1E-3
1.7E-3
1.8
8.2
Argonne National Laboratory-West
EBR-II; HFEF-S; Laboratory
& Office
Sodium Components
Cleanup Facility
Fuel Assembly & Storage
Area
Argon-41
Bromine-82
Tritium-3
Krypton-85
Krypton-85m
Krypton-87
Krypton-88
Rubidium-88
Xenon-133
Xenon-135
Xenon-135m
Xenon-138
Cesium-137
Unidentified
Alpha
Unidentified
Beta & Gamma
Unidentified
Alpha
Unidentified
Beta & Gamma
7.4
1.2E-1
6.7E-1
5.0
2.1
1.2
1.7
8.4E-3
3.4E+2
.3E+2
1E-3
1
2.
9.4E-3
1.3E.-7
1.8E-7
1.1E-6
2.1E-7
1.5E-6
-------
3.3-13
Table 3.3-1. Atmospheric emissions of radionuclides
from the Idaho Engineering Laboratory Site, 1977--continued
Source
Radionuclide
Emissions
(Ci/y)
Transient Reactor
Test
Argon-41
Krypton-88
Xenon-133
Xenon-135
Zero Power Plutonium Reactor
Facility
Krypton-85M
Krypton-87
Krypton-88
Xenon-133
Xenon-135
Cesium-137
Unidentified
Alpha
Unidentified
Beta & Gamma
Radiological & Environmental
Sciences Laboratory Cesium-134
Hot Fuel Examination
Facility-North
1.4E+2
2.2
2.7E-1
5.4E-1
5.8E-1
1.1
1.2
1.1E-2
4.3E-1
2.1E-7
2.5E-7
1.5E-5
l.OE-1
-------
3.3-14
3.3.4 Health Impact Assessment of the INEL Site
The maximum total body dose to a hypothetical individual at
the site boundary from radioactive emissions was reported to be 0.10
mrem for 1977. This calculated dose is the 50 year dose commitment
for chronic exposure during 1977. The maximum concentrations
leading to this dose occurred along the southern site boundary and
about 94 percent of this dose was found to be due to noble gases and
particulates of isotopes having half-lives of less than 10 hours.
The maximum potential dose to an individual of population
group was reported to be 0.074 mrem/yr. Terreton-Hamer was the
population group reported to have the greatest potential dose from
site operations. The estimated life time risk of fatal cancer for a
resident of this town is about l.OE-6.
The estimated dose of 0.90 person-rem/y to the population
within 80 kilometers of INEL resulted from radioactive emissions
from operations at the Test Reactor Area and the Idaho Chemical
Processing Plant which emit more than 98 percent of the total
radioactive emissions from the Idaho site. The number of cancers
per year of operations is estimated to be 1.8E-4 for the population
living in the region of INEL.
Table 3.3-2. Radiation dosesa due to radioactive emissions
from Idaho National Engineering Laboratory (DOE77)
Source
Maximum
Site
boundary
(mrem)
individual
Resident of
Terreton-Hamer
(mrem)
Population^3
(person-rem)
INEL 0.10 0.074 0.90
^Fifty-year dose commitment.
"The population is within 80 km.
-------
3.3-15
Table 3.3-3. Individual lifetime risks and number of fatal cancersa
from INEL operations, 1977
Individual lifetime risks Expected fatal cancers
Source Site Resident of per year of operation
boundary Terreton-Hamer Region (Fatal cancers)
INEL 1.4E-6 l.OE-6 1.3E-7 1.8E-4
aTo population within 80 kilometers
-------
3.3-16
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No 02, Narrative Summary Data Base Master List (EIS 02).
DOE77b Department of Energy, 1977, Effluent Information System
Report No 51. Release Point Analysis Report for Calendar year 1977
(EIS 51).
DOE78 Department of Energy, 1978, 1977 Environmental Monitoring
Program Report for Idaho National Engineering Laboratory Site.
ERDA77 Energy Research and Development Administration, 1977, Final
Environmental Impact Statement, ERDA-1536, Waste Management
Operations, Idaho National Engineering laboratory, Idaho.
-------
3.4-1
3.4 Los Alamos Scientific Laboratory
3.4.1 General Description
The Los Alamos Scientific Laboratory (LASL) was established
in 1943 for the purpose of nuclear weapons research and
development. Today, in addition to weapons research, LASL conducts
national security programs including laser fusion, nuclear materials
research, and laser isotope separation. Research is also conducted
in areas of power reactors, magnetic fusion, radiobiology and
medicine, astrophysics, earth sciences, energy resources, lasers and
the environment.
LASL is located in Los Alamos County, New Mexico, in the
north-central area of the State, near the towns of White Rock and
Los Alamos, about 100 kilometers north-northeast of Albuquerque and
40 kilometers northwest of Sante Fe. LASL facilities occupy 11,100
hectares in a relatively undeveloped area on top of the Pajarito
Plateau. About 98,000 people reside within 80 kilometers of the
laboratory facilities.
3.4.2 Process Description
A wide variety of facilities support the research and dev-
opment programs at Los Alamos Scientific Laboratory. These
facilities are located within 30 Technical Areas throughout the LASL
site. The major facilities at LASL include an 800 MeV proton
accelerator (TA-53), a Tandem Van de Graaff accelerator (TA-3), the
laser and magnetic fusion laboratories, and a 10 MW research reactor
(TA-2).
Airborne radioactive effluents are discharged from a number
of facilities. Those facilities which conduct operations that could
have significant releases use appropriate treatment methods to
control releases of radioactivity. Hot cells, laboratory hoods and
glove boxes are equipped with filtration systems to remove
particulate radioactivity. HEPA filters are the principal type of
filter used where plutonium and uranium handling operations are
conducted. Charcoal filters, bag filters and cyclone separators are
used where applicable. Where feasible, tritium releases are reduced
from effluents by catylysts, microsieves, and adsorbers. Short-
lived activation gas releases are delayed to allow reduction by
radioactive decay.
-------
3.4-2
3.4.3 Emissions of RadionucTides
Radioactive airborne emissions are released in stack exhausts
from eleven (11) principal Technical Areas. The quantities shown in
table 3.4-1 may have significant year-to-year variations depending
on the research being conducted.
3.4.4 Health Impact Assessment of Los Alamos Scientific Laboratory
The maximum whole body dose to a hypothetical individual
located at the site boundary of LASL was calculated to be 69 mrem
for 1977 (table 3.4-2). This dose was primarily due to activation
products released in airborne effluents from the proton
accelerator. This dose would correspond to a lifetime risk of fatal
cancer of 9.7E-4 (table 3.4-3).
The estimated whole body dose to the nearest individual in a
population group was calculated to be 19 mrem for 1977. This dose
occurred to the north of Technical Area 53 and was due to the
activation gases in the effluent from the proton accelerator
facility. The lifetime risk to an individual receiving this dose is
estimated to be 2.7E-4.
The whole body population dose to the residents within an 80-
kilometer radius of LASL was calculated to be 11.1 person-rem for
1977. The activation products argon-41, carbon-11, nitrogen-13 and
oxygen-15 were the principal contributors to the dose. The number
of fatal cancers per year of LASL operation is estimated to be
2.2E-3 to the population within 80 kilometers.
-------
3.4-3
Table 3.4-1. Atmospheric emissions of radionuclides,
Los Alamos Scientific Laboratory, 1977 (DOE77b)
Source
Radionuclide
Emissions
(Ci/y)
Technical Area 2 (TA-2)
Omega Site
Technical Area 3 (TA-3)
Chemical Metallurgical
Research Building
Press Building
Sigma Building
Technical Shops
Addition
Rolling Mill Building
Argon-41
3.2E+2
Iodine-131
Mixed Fission
Products
Plutonium-238 and
Plutonium-239
Plutonium-239
Uranium-235 and
Uranium-238
Uranium-235
Thorium-234
Uranium-235
Uranium-235and
Uranium-238
Uranium-238
Uranium-235 and
Uranium-238
Uranium-235 and
Uranium-238
Van de Graaff Facility Tritium
Technical Area 9 (TA-9)
Anchor Test Site Tritium
See footnotes at end of table.
8.8E-5
4.8E-4
2.6E-5
6.4E-6
3.6E-5
2.4E-6
5.2E-3
3.9E-6
1.8E-6
2.8E-4
3.3E-6
6.6E-6
4.0E+2
2.6
-------
3.4-4
Table 3.4-1. Atmospheric emissions of radionucTides,
Los Alamos Scientific Laboratory, 1977 (DOE77b)--continued
Source
Radionuclide
Emissions
(Ci/y)
Technical Area 21 (TA-21)
DP Site
Tritium 1.3E+2
Mixed Fission
Products 3.3E-6
Plutonium-238 2.7E-7
Plutonium-238 and
Plutonium-239 3.9E-6
Plutonium-239 5.8E-6
Uranium-235 3.2E-4
Technical Area 33 (TA-33)
HP Site Tritium
Technical Area-35 (TA-35)
Ten Site
Technical Area-43 (TA-43)
Health Research
Laboratory
Technical Area-46 (TA-46)
WA Site
Technical Area-48 (TA-48)
Radiochemistry Site
Technical Area-50 (TA-50)
Liquid Waste
Treatment Plant
Tritium
Plutonium-239
Phosphorus-32
Plutonium-239
Plutonium-239
Uranium-238
Mixed Fission
Products
Plutonium-239
Uranium-235
Mixed fission
Products
Plutonium-239
a3.7E+4
7.9E+2
8.2E-7
3.0E-4
4.7E-6
3.0E-9
4.0E-9
2.2E-3
8.4E-6
5.5E-5
8.6E-5
7.0E-5
See footnotes at end of table.
-------
3.4-5
Table 3.4-1. Atmospheric emissions of radionuclides,
Los Alamos Scientific Laboratory, 1977 (DOE77b)--continued
Source
Radionuclide
Emissions
(Ci/y)
Technical Area-50 (TA-50)
Liquid Waste
Treatment Plant
Technical Area-53 (TA-53)
Meson Physics Facility
Mixed fission
Products
Plutonium-239
Argon-41
Beryllium-7
Carbon-11
Tr i t i urn
Mixed Activation
Products
Nitrogen-13
Oxygen-15
8.6E-5
7.0E-5
4.8E+2
2.8E-7
1.4E+4
2.9E+2
5.0E-9
1.4E+3
3.2E+4
^Includes 30800 Ci accidental release on October 6, 1977.
Table 3.4-2. Annual radiation doses due to radioactive emissions
from Los Alamos Scientific Laboratory, 1977 (La78)
Critical organ
and
radionuclide
Maximum individual
Site boundary Nearest resident Population9
(mrem/y) (mrem/y) (person-rem/y)
Total body
Tritium
Carbon-11,
Nitrogen-13
and Oxygen-15
Argon-41
Lung
Plutonium-239
Total
0.42
67.0
2.1
.06
69.5
.09
19.0
.9
.06
20
0.4
7.1
3.6
NR
11.1
I W I* M I t-r\J \*IJ J S*
NR Not reported.
-------
3.4-6
Table 3.4-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Los Alamos Scientific Laboratory, 1977
Individual lifetime risks Expected fatal cancers
Source Site Nearest Average per year of operation
boundary resident individual^ (Fatal cancers)b
LASL 9.7E-4 2.7E-4 1.6E-6 2.2E-3
region extends to 80 kilometers.
°To the population within 80 kilometers.
-------
3.4-7
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Reports Nos. 02 and 05, Narrative Summary Data Base Master list
and Narrative Summary Data Base Master List Updated Records Report
for 1977, (EIS 02 and EIS 05), (Computer Listing).
DOE77b Department of Energy 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977 (EIS-51), (Computer Listing).
DOE78 Department of Energy, 1978, Draft Environmental Impact
tatement, Los Alamos. Scientific Laboratory Site, Los Alamos, New
Mexico, DOE/EIS-00/8-D.
LA78 Los Alamos Scientific Laboratory, 1978, Environmental
Surveillance at Los Alamos During 1977, UC-41, LA-7263-MS, Los
Alamos, N.M.
-------
3.5-1
3.5. Lawrence Livermore Laboratory
3.5.1 General Description
Lawrence Livermore Laboratory (LLL) was established in 1952
for nuclear weapons research and development. In addition to its
prime mission, programs are carried out in the areas of magnetic
fusion research, non-nuclear energy research, biomedical studies,
laser fusion research, and laser isotope separation research.
Lawrence Livermore Laboratory is located about 64 kilometers
east of San Francisco, California, on a 254-hectare site in the
Livermore Valley of southern Alameda County. The City of Livermore
is 5 kilometers to the west of the site.
LLL is situated on an alluvial flood plain bordered by the
hills of the Livermore Uplands. The area surrounding the valley is
primarily used for pasture lands. The principal agricultural
products of the area include grapes and wine, cattle, poultry and
eggs. The population within 80 kilometers of LLL is about 4.3
mi 11i on.
3.5.2 Process Description
There are six principal facilities that release radioactivity
into the air at Lawrence Livermore Laboratory.
Light Isotope Handling Facility
Tritium is the principal nuclide released from this facility
which is involved with research and development in the area of light
isotopes. There is no system employed to reduce tritium from the
airborne effluents. The two stacks from this facility are monitored.
Livermore Pool Type Reactor
The Livermore Pool Type Reactor operates at a thermal power
of 3 MW to provide neutron irradiations for basic and applied
research supporting LLL programs.
Argon-41, produced by air flowing through beam ports and
irradiation cells, is essentially the only radioactive stack
effluent under normal operating conditions. Argon-41 is released
without treatment but the stack is continuously monitored. In the
-------
3.5-2
event the reactor building cannot be maintained at negative
pressure, air is rerouted through a bank of HEPA filters,
activated-charcoal filters, and KOH scrubbers.
Insulated Core Transfer Accelerator (ICT)
The ICT facility houses many medium energy accelerators.
However, the ICT accelerator is the only unit having off-site impact.
The ICT accelerator is an air-insulated variable energy
machine which accelerates protons and deuterons up to 500 keV. The
accelerator uses tritium targets for production of 14 MeV neutrons
in support of the Magnetic Fusion Energy Program. Tritium is
released from the facility without treatment. The effluent is
continuously monitored.
Electron Positron Linear Accelerator (LINAC)
Operation of the 100 MEV LINAC for neutron physics research,
produces activation of nitrogen, oxygen, and dust particles in the
air of the facility. The activation gases, primarily oxygen-15 and
nitrogen-13, are released without treatment. HEPA filters are used
to reduce particulate radioactivity in the airborne effluent
stream. The effluent stream is continuously monitored before
release to the atmosphere from a 30-meter high stack.
Decontamination Facility
HEPA filters are used to reduce particulate radioactivity
from exhaust air. The radioactivity in air effluents originate from
various decontamination operations. Stack effluents are
continuously sampled.
Solid Waste Disposal Facility
Radioactive solid waste packaging, holding, and shipping
activities are conducted at this facility. Transfer and compacting
operations of dry waste may result in particulate activity being
released into the facility ventilation and process air. This air is
passed through HEPA filters before release to the atmosphere.
During operations the stack effluent is sampled.
3.5.3 Emissions of Radionuclides
Table 3.5-1 identifies radioactive emissions from the
facilities at Lawrence Livermore Laboratory in 1977.
-------
3.5-3
Table 3.5-1. Atmospheric emissions of radionuclides,
Lawrence Livermore Laboratory, 1977 (DOE77b)
Source
Radionuclide
Emissions
(Ci/y)
Light Isotope
Handling Facility
Livermore Pool Type
Reactor
ICT Accelerator
Electron Positron
Linear Accelerator
(LINAC)
Decontamination
Facility
Solid Waste Disposal
Facility
Tritium
Argon-41
Tr i t i urn
Nitrogen-13
Oxygen-15
Unidentified
Beta & Gamma
Unidentified
Beta & Gamma
Unidentified
Alpha
Unidentified
Beta & Gamma
3.1E+3
3.8E+2
2.1E+3
5.9E+2
3.9E+2
4.1E-5
3.5E-7
2.4E-9
1.1E-8
3.5.4 Health Impact Assessment
Table 3.5-2 summarizes doses from radioactive emissions from
Lawrence Livermore Laboratory in 1977. The maximum total body dose
to a hypothetical individual at the site boundary of LLL was
calculated to be 4.3 mrem from nitrogen-13 and oxygen-15 released
from the electron positron linear accelerator. These emissions also
resulted in a 1.0 mrem dose to the nearest resident. A population
dose of 2.94 person-rem was estimated for the population within 80
kilometers of LLL.
-------
3.5-4
Table 3.5-3 estimates the individual lifetime risks and
number of fatal cancers to the population resulting from these
doses. The lifetime risk of fatal cancer to the nearest resident is
estimated to be 1.4E-5. The number of fatal cancers per year of ILL
operation is estimated to be 5.9E-4 to the population within 80
kilometers.
Table 3.5-2. Annual radiation doses from atmospheric emissions of
radionuclides from Lawrence Livermore Laboratory, 1977 (Si78)
Maximum individuals
Source and Site boundary Nearest resident Population
Radionuclide (mrem/y) (mrem/y) (Person-rem/y)
Light Isotope
Handling Facility
Tritium 0.3 0.03 0.3
Livermore Pool
Type Reactor
Argon-41 3.2 0.4 2.1
Insulated Core
Transfer Accelerator
(ICT)
Tritium 3.0 .04 0.3
Electron Positrom
Lincear Accelerator
(LINAC)
Nitrogen-13,
Oxygen-15 4.3 1.0 0.2
Total 2.9
aDoses are not additive since they occur at different locations.
-------
3.5-5
Table 3.5-3. Individual lifetime risks and number of fatal
cancers9 due to radioactive emissions from
Lawrence Livermore Laboratory, 1977
Source
Individual lifetime risks
Site Nearest Average
boundary resident individual
Expected fatal cancers
per year of operation
(Fatal cancers)
LLL
6.0E-5
1.4E-5
9.4E-9
5.9E-4
aTo the population within 80 kilometers.
-------
3.5-6
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List (EIS 02),
(Computer Listing).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, (EIS 51), (Computer Listing).
DOE78 Department of Energy, 1978, Draft Environmental Impact
Statement, Livermore Site, Livermore, California, DOE/EIS-0028-D.
Si78 Silver W.J., C.L. Llndenken, K.M. Wong, E.H. Willes, J.H.
White, 1978, Environmental Monitoring at the Lawrence Livermore
Laboratory 1977 Annual Report, Livermore, California.
-------
3.6-1
3.6 Rocky Flats Plant
3.6.1 General Description
The Rocky Flats Plant is part of the national nuclear weapons
research, development, and production program. Its primary mission
is to produce plutonium components for nuclear weapons.
The Rocky Flats Plant is located in Jefferson County,
Colorado, approximately 26 kilometers northwest of Denver. The
facilities are located within a 155-hectare security area which is
situated on 2650 hectares of Federally-owned land. The site is on
the eastern edge of a geological bench, with the foothills of the
Rocky Mountains to the west. The area immediately surrounding the
plant is primarily agricultural or undeveloped. However, about 1.8
million people reside within 80 kilometers.
3.6.2 Process Description
The Rocky Flats Plant is primarily a radioactive metal
fabrication and chemical processing plant. Its mission involves
foundry and fabrication of plutonium and uranium components,
chemical processing and recovery of plutonium from scrap material,
and other transuranic purification operations. These activities are
supported by other disciplines such as nuclear safety, engineering,
health physics, environmental control, and research and development.
Plutonium at the Rocky Flats Plant is stored in closed
containers in a vault with an inert atmosphere. Ingots of plutonium
taken from the vault undergo metallurgical processes which include
reduction rolling, blanking, forming and heat treating. Sma.ller
pieces of plutonium are drilled or broken to provide samples for the
Analytical Laboratory and for casting operations. The formed pieces
are then machined into the various components which are then
assembled. Assembly operations include cleaning, brazing, marking,
welding, weighing, matching, sampling, heating and monitoring.
Nuclear weapons are not assembled at this plant.
Solid residue generated during plutom'um-related operations
are recycled through one of two plutonium recovery processes; the
process selected depends on the purity and content of plutonium in
the residue. Both processes result in a plutonium nitrate solution
from which the metal can be extracted. The recovered plutonium is
returned to the storage vault for use in foundry operations. A
secondary objective of the process is the recovery of americium-241.
-------
3.6-2
Rocky Flats Plant also conducts operations involving the
handling of uranium. Depleted uranium-alloy scrap is consolidated
and recycled at one of the foundries. The depleted uranium alloys
are ore-melted into ingots for further metallurgical processing.
Rocky Flats also has the capabilities to machine and assemble
enriched uranium pieces. Enriched uranium components, returned
because of age, are disassembled. The enriched uranium is separated
and then sent to Oak Ridge, Tennessee, for recycling.
Research and development operations at Rocky Flats Plant are
directed towards improving the methods by which plutonium components
are produced. Scientific investigations are conducted in all the
production oriented fields. These operations are carried out in
many areas throughout the plant site.
Because of its toxicity, plutonium is stored and processed
under strictly controlled conditions. Much of the plutonium
processing equipment is enclosed in glove boxes with an inert,
nitrogen atmosphere. The glove boxes are maintained at a slight
negative pressure relative to the surrounding area. This allows
ventilation air to flow toward areas of greater radioactive
contamination instead of away from them.
Many operations, such as handling oxide powder, machining
metalic materials, incinerating scrap and waste materials, and
chemical recovery processes, have the capability of releasing fine
particles of radioactive material into the ventilation air or
process gas stream. These effluent streams are passed through HEPA
filters to remove the contaminants.
The Rocky Flats Plant uses a minimum of two stages of HEPA
filtrati'on for all general building air where plutonium handling is
conducted. Plutonium glove box and process air undergoes four
stages of HEPA filtration prior to release. Three or four stages of
HEPA filters may be expected to provide decontamination factors up
to 1C)9 to 10^. In buildings where only uranium is handled, at
least one stage of HEPA filtration is employed.
3.6.3 Emissions of Radionuclides
Since the upgrading of ventilation and HEPA filtration
systems at the plutonium processing facilities was completed in
1970, plutonium releases to the atmosphere have been held below 100
microcuries per year from normal operations. Based on current
performance, the expected yearly atmospheric emissions of plutonium
from Rocky Flats would be about 10 microcuries per year. Table
3.6.1 shows that the 1977 releases were below that level.
-------
3.6-3
Following improvements made to the filtration systems in the
uranium handling buildings in 1970, yearly releases of depleted and
enriched uranium have been less than 50 microcuries per year each.
The tritium released in 1977 from Rocky Flats was due to
residuals from a release in 1973 when tritium contaminated material
was inadvertently processed.
Table 3.6-1. Atmospheric emissions of radioactive materials from
the Rocky Flats Plant, 1977 (DOE77b)
Radionuclide
Tritium
Plutonium-239 and
Plutonium-240
Uranium 235
Uranium-238
Emissions
(Ci/y)
5.3E-1
4.2E-6
2.1E-5
1.9E-5
3.6.4 Health Impact Assessment of Rocky Flats Plant
The estimated doses resulting from radioactive emissions from
the Rocky Flats Plant in 1977 are listed in table 3.6-2. The
maximum potential dose at the site boundary, based on continuous
exposure to measured concentrations of plutonium in the air, was to
the east of the plant. The dose of about 0.2 millirem to a
hypothetical resident of Denver in 1977 was based on exposure to
concentrations of plutonium in air in the Denver area in excess of
background concentrations. The Rocky Flats population dose was
based on atmospheric concentrations measured in surrounding
communities. The total dose to the population within 80 kilometers
was about 200 person-rems.
Table 3.6-3 estimates the individual lifelime risks and fatal
cancers in the population resulting from these doses. The number of
fatal cancers for plant operation in 1977 is estimated to be about
7.2E-3 to the population within 80 kilometers.
-------
Table 3.6-2.
3.6-4
Annual radiation doses3 from radionucTides
from Rocky Flats Plant, 1977
Organ
Lung
Bone
Kidney
Liver
Total body
Maximum
Site
boundary
(mrem/y)
8.0E-2
3.5E-2
4.0E-3
5.4E-3
8.6E-4
individual
Denver
resident
(mrem/y)
1.3E-1
5.5E-2
6.3E-3
8.5E-3
1.4E-3
Populationb
(Person-rem/y)
1.3E+2
5.8E+1
6.6
8.9
1.4
aDoses are based on measured concentrations in the air in excess
of background concentrations. Denver concentrations are higher than
site boundary concentrations.
''The population is residing within 80 kilometer of the plant.
Table 3.6-3. Individual lifetime risks and number of fatal
cancers3 due to radioactive emissions from Rocky Flats Plant, 1977
Organ Individual lifetime risks
Maximum individual Denver
at site boundary resident
Expected fatal cancers
per year of operation
(Fatal cancers)
Lung
Bone
Kidney
Liver
Other soft
tissue
2.2E-7
7.4E-8
2.8E-9
3.8E-9
3.0E-9
3.5E-7
1.2E-7
4.4E-9
5.9E-9
4.7E-9
5.3E-3
1.7E-3
6.6E-5
8.9E-5
7.1E-5
Total
3.0E-7
4.9E-7
7.2E-3
3To the population within 80 kilometers.
-------
3.6-5
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report Nos. 02 and 05, Narrative Summary Data Base Master list and
Narrative Summary Data Base Master List Updated Report for 1977,
(EIS 02 and EIS 05), (Computer Listing).
DOE775 Department of Energy, 1977, Effluent Information System
Report No. 50, Release Point Analysis Report (EIS 51), (Computer
Listing).
Ro78 Rockwell International, 1978, Annual Environmental Monitoring
Report, January-December 1977, RFP-ENV-77, Rocky Flats, Colorado.
-------
3.7-1
3.7 Mound Facility
3.7.1 General Description
The Mound Facility has been in operation since 1949. Its
primary functions include research, development, engineering,
production and surveillance of components for the DOE weapons
program. Other operations involve the separation, purification, and
sale of stable noble gas isotopes and the fabrication of
radioisotopic heat sources for thermoelectric generators.
The Mound facility, located in Miamisburg, Ohio, about 16
kilometers southwest of Dayton, occupies a 73-hectare site in the
Great Miami River Valley. This area is highly industrialized. The
surrounding region is mostly agricultural with some light industry
and scattered residential communities. About 2.8 million people
live within 80 kilometers of the Mound Facility.
3.7.2 Process Description
Nine buildings at the Mound Facility released radioactivity
into the atmosphere in 1977. Operations at these facilities
resulted in the release of tritium and plutonium-238.
Tritium was released in atmospheric effluents from the HH and
SW buildings. Operations at the HH building involve the recovery of
helium-3 which is contaminated with tritium. Gaseous wastes
generated here are stored and transfered to the SW building. At the
SW building operations involve disassembly, analysis and development
of nuclear components containing tritium, and the recovery of
tritium wastes. Tritium in gaseous effluents streams of the SW
building are treated before release by the effluent removal system,
which oxidizes elemental tritium and then removes the resulting
tritiated water by molecular sieve drying beds.
Plutonium-238,was released in airborne effluents from PP, R,
WD, WDA, 41, H and SM buildings. Plutonium processing and other
related activities are conducted at the PP building. At the R
building plutonium heat source production is the principal
activity. Operations at the WD, WDA and 41 buildings involve
radioactive waste disposal processes. Contaminated clothing is
laundered at the H building, and the SM building is an idle
contaminated facility. At all these facilities, particulate
radioactivity is removed from process air streams by HEPA filters.
-------
3.7-2
The airborne effluents undergo filtration at their point of
generation and again at the stack just prior to release.
3.7.3 Emissions of Radionuclides
In 1977 a total of 4.9E+3 curies of tritium and 0.1
microcuries of plutonium-238 were released from the Mound Facility.
Table 3.7.1 shows the atmospheric emissions from each building.
Table 3.7-1. Emissions of radionuclides from
the Mound Facility, 1977 (DOE77b)
Facility
Radionuclide
Emissions
(Ci/y)
H Building
HH Building
PP Building
R Building
SM Building
SW Building
WD Building
WDA Building
41 Building
Plutonium-238
Tritium
Plutonium-238
Plutonium-238
Plutonium-238
Tritium
Plutonium-238
Plutonium-238
Plutonium-238
4.9E-13
9.5E+01
4.1E-06
3.7E-07
.6E-06
.8E+03
.7E-08
.7E-06
1.2E-07
3.7.4 Health Impact Assessment of the Mound Facility
The maximum individual annual dose to a person at the site
boundary of the Mound Facility due to 1977 radioactive atmospheric
emissions was 0.11 rem/y to the lung from inhalation of
plutonium-238. These same releases resulted in a maximum dose to
the lung of .04 mrem/y to the nearest resident. Dose estimates to
the bone and whole body from exposure to atmospheric releases of
Plutonium and tritium from the Mound Facility are shown in Table
3.7-2. The maximum lifetime risk of fatal cancer to the nearest
person is 2.8E-7.
The dose to the population within 80 kilometers from airborne
tritium was calculated to be 3.7 person-rem for 1977. This would
correspond to 7.2E-4 fatal cancers per year of operation to the
population surrounding the Mound Facility.
-------
Table 3.7-2.
3.7-3
Annual radiation doses from atmospheric emissions,
Mound Facility, 1977 (Fa77)
Maximum individual a
Organ Site boundary Nearest resident
(mrem/y) (mrem/y)
Population
(Person-rem/y)
Bone
Lung
Total body
4.5E-2
1.1E-1
4.0E-2
1.5E-2
4.0E-2
2.0E-2
NR
NR
3.6
aDoses are not additive since they occur at different locations.
NR Not reported.
Table 3.7-3. Individual lifetime risks and number of fatal cancers
due to radioactive emissions from Mound Facility, 1977
Individual lifetime risks Expected fatal cancers
Organ Nearest Average individual per year of operation
resident Region (Fatal cancers)
Bone
Lung
Total body
3.2E-8
l.OE-7
2.8E-7
NR
NR
1.8E-8
NR
NR
7.2E-4
NRNot reported.
-------
3.7-4
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report Nos. 02 and 05, Narrative Summary Data Base Master List and
Narrative Summary Data Base Master List Updated Report for 1977,
(EIS 02 and EIS 05).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977 (EIS 51).
Fa78 Farmer B., Robinson B. M., Carfagno D.G., 1978, Annual
Environmental Monitoring Report: Calendar Year 1977, MLM-2515.
-------
3.8-1
3.8 Pantex Plant
3.8.1 General Description
The Pantex Plant's mission includes atomic weapons assembly,
retirement, and stockpile surveillance. The plant also fabricates
and tests chemical explosives.
The Pantex Plant, located on 3683 hectares of land in Carson
County, Texas, is in the Panhandle Plains area about 27 kilometers
northeast of Amarillo. The Panhandle is predominately an
agricultural area; the population is about 236,000 persons within 80
kilometers.
3.8.2 Process Description
High explosive test firings, which are conducted
intermittently, release small quantities of depleted uranium.
Disassembly, shipping, and receiving operations within the plant
release small amounts of tritium. There are no sources of
continuous releases of radioactive material to the atmosphere. With
few exceptions, radioactive material is handled in sealed
containers. These containers are not opened, thus avoiding the
possiblity of release during normal operations. There are no
systems employed to treat the radioactive emissions to the
atmosphere.
3.8.3 Emissions of Radionuclides
The estimated release of radioactive materials for 1977 from
Pantex Plant are shown in Table 3.8-1.
Table 3.8-1. Atmospheric emissions of radionuclides
from the Pantex Plant, 1977 (A178)
Facility Radionuclide Emissions
(Ci/y)
Firing site Uranium-238 l.OE-3
Assembly area Tritium l.OE-2
-------
3.8-2
3.8.4 Health Impact Assessment of the Pantex Plant
Table 3.8-2 summarizes the estimated radiation doses due to
radioactive emissions from the Pantex Plant. The principal
radionuclide contibuting to these doses is uranium-238.
The total body dose to the the population within 80 kilometers
of the Pantex Plant in 1977 was 4.0E-4 person-rem from uranium-238 and
2.0E-7 person-rem from tritium. These doses would result in an
estimated 8.0E-8 fatal cancers per year of plant operations to the
population within 80 kilometers. Table 3.8-3 summarizes the
individual lifetime risks and health effects associated with Pantex
Plant operations.
Table 3.8-2. Annual radiation doses due to radioactive emissions
from Pantex Plant, 1977 (A178)
Maximum individual
Organ
Lung
Kidney
Total body
Site
boundary
(mrem/y)
9.0E-5
2.0E-4
3.0E-6
Nearest
resident
(mrem/y)
6.0E-5
l.OE-4
2.0E-6
Population
(person-rem/y)
NR
NR
4.0E.-4
NRNot reported.
Table 3.8-3. Individual lifetime risks and number of fatal cancersa
from Pantex Plant, 1977
Individual lifetime risks Expected fatal cancers
Organ Nearest Average individual per year of operation
resident Region (Fatal cancers)
Lung
Kidney
Other soft
tissue
Total
1.7E-10
7.0E-11
7.0E-12
2.5E-10
NR
NR
NR
2.4E-11
NR
NR
NR
8.0E-8
aTo the population within 80 kilometers.
NR Not reported.
-------
3.8-3
REFERENCES
A178 Alexander Ronald E., and C. Newlyn Morton, 1978, Environmental
Monitoring Report for Pantex Plant Covering 1977, MHSMP-78-7.
DOE77a Department of Energy, 1977, Effluent Information System
Reports Nos. 02 and 05. Narrative Summary Data Base Master List
and Narrative Summary Data Base Master List Updated Report for
1977, (EIS 02 and EIS 05), (Computer Printout).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report For Calendar Year
1977, (EIS 51), (Computer Printout).
-------
3.9-1
3.9 Pinellas Plant
3.9.1 General Description
Pinellas Plant is operated by the Neutron Devices Department
of the General Electric Company. Operations involve the design,
development, and manufacture of special electronic and mechanical
components for nuclear weapons.
Pinellas Plant, located on a 39.2 hectare-tract in Pinellas
County, Florida, is on Florida's West Coast, north of St. Peters-
burg. The county has had a rapid population growth and is the most
densely populated county in Florida with about 1.7 million people
residing within 80 kilometers of the plant.
3.9.2 Process Description
The principal operations causing atmospheric releases of
radioactive materials are not described in the literature. However,
they involve neutron generator development and production, testing,
and laboratory operations. Areas utilizing radioactive materials
are connected to a special exhaust system which is designed to trap
tritium and reduce the amount released to the atmosphere. In this
system tritium gas is converted to the oxide form by passage through
heated copper oxide beds. Then the tritiated water vapor is
absorbed by silica! gel.
Small sealed plutonium capsules are used as heat sources in
the manufacture of radioisotopic thermoelectric generators at
Pinellas Plant. These sources are triply encapsulated so as to
prevent release of plutonium to the atmosphere.
3.9.3 Emissions of RadionucTides
Small amounts of tritium gas, tritium oxide, and krypton-85
were released from one of two 30-meter stacks at Building 100. The
releases for 1977 are summarized in Table 3.9-1.
3.9.4 Health Impact Assessment of Pinellas Plant
Table 3.9-2 summarizes doses from radioactive atmospheric
emissions from the Pinellas Plant in 1977. The maximum site
boundary dose occurred along the western perimeter while the nearest
residential area is located approximately 1.6 kilometers to the
south-southeast of the plant. A population dose of 0.4 person-rem
from atmospheric emissions of tritium was estimated for the nearly
1.7 million people living within 80 kilometers of the Pinellas Plant.
Estimates of the individual lifetime risks and the number of
fatal cancers resulting from these doses are given in table 3.9-3.
-------
3.9-2
Table 3.9-1. Atmospheric emissions of radionuclides
from Pinellas Plant, 1977 (DOE77b)
Source Radionuclide Emissions
(Ci/y)
Building 100-Main stack Tritium 1.6E+2
Krypton-85 2.8E+1
Building 100-Laboratory
stack Tritium 1.3E+2
Table 3.9-2. Annual radiation doses from atmospheric emissions of
radioactive materials from Pinellas Plant, 1977 (GE78)
Maximum individual
Radionuclide Site boundary Nearest resident Population^
(mrem/y) (mrem/y) (person-rem/y)
Krypton-85
Tritium
Total
1.3E-3
l.OE-2
1.1E-2
NR
3.8E-3
3.8E-3
NR
NR
0.4
aThe population is within 80 kilometers.
NR Not reported.
-------
3.9-3
Table 3.9-3. Individual lifetime risks and number of fatal
cancers resulting from atmospheric emissions of radioactive
materials from Pinellas Plant, 1977
Radio-
nuclide
Krypton-85
Tritium
Individual lifetime risks
Maximum Nearest Average
Individual Resident individual
1.8E-8 NR NR
1.4E-7 5.3E-8 3.3E-9
Expected fatal cancers
per year of operation
(Fatal cancers)
NR
8.0E-5
aTo the population within 80 kilometers.
NR Not reported.
-------
3.9-4
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System Re-
ports Nos. 02 and 05, Narrative Summary Data Base Master List and
Narrative Summary Data Base Master List Updated Report for 1977,
(EIS 02 and EIS 05), (Computer Printout).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, (EIS 51), (Computer Printout).
GE78 General Electric Company, 1978, Pinellas Plant Environmental
Monitoring Report, St. Petersburg, Florida.
Rh73 Rhinehammer T.B. and P.H. Lamberger, 1973, Tritium Control
Technology, WASH-1269, U.S. Atomic Energy Commission, Miamisburg,
Ohio.
-------
3.10-1
3.10 Sandia Laboratories
3.10.1 General Description
Sandia Laboratories is a nuclear ordnance laboratory which
combines nuclear weapons developed by Los Angeles Scientific
Laboratory and Lawrence Livermore Laboratory with delivery systems
needed by the military services. This responsibility includes
performing weapons testing, quality control and assurance, arming
and fusing, safety, delivery system modification, and safeguards.
Components are tested for proper operation under a variety of
environmental conditions involving parameters such as shock,
vibration, temperature, moisture and radiation.
The Sandia Laboratories, are located in Albuquerque, New
Mexico, and Livermore, California. The Livermore site is adjacent
to Lawrence Livermore Laboratory. In Albuquerque, Technical Area V
is where much of the radiation testing is performed. This area is
located approximately 10 kilometers south of the city in a sparsely
populated regionabout 380,000 people live within 80 kilometers,
mostly inside a 20-kilometer radius.
3.10.2 Process Description
Albuquerque Site
As part of the environmental testing capability, Sandia
Laboratories operates two research reactors which are located in
Technical Area V. The Sandia Pulsed Reactor is an unreflected,
cylindrical, enriched-uranium assembly. The Annular Core Pulse
Reactor is a modified TRIGA-type reactor. Reactor operations
release small amounts of fission and activation product gases,
primarily argon-41. Both reactor building exhaust effluents undergo
HEPA filtration prior to monitoring and release.
Also located within Technical Area V is the Relativistic
Electron Beam Accelerator Facility which is used for electron beam
fusion research. Small amounts of tritium gas are emitted from the
vacuum pump exhaust. Tritium gas in the effluent stream is
catalyzed to the oxide from which it is then removed by molecular
sieves.
Technical Area I has research, design, administrative and
support facilities. The principal operation emitting radioactivity
to the atmosphere in this area is from neutron activation
experiments. Small quantities of tritium are released
-------
3.10-2
Table 3.10-1. Atmospheric emissions of radionuclides
from Sandia Laboratories, 1977 (DOE77b)
Source
Radionuclide
Emissions
(Ci/y)
Albuquerque Site
Annular Core Pulsed
Reactor
Sandia Pulsed Reactor
Relativistic Electron
Beam Accelerator
Neutron Generator
Livermore Site
Building 913
Argon-41
Argon-41
Tritium
Tr i t i urn
3.9
9.2E-01
4.0E-07
l.OE-02
Unidentified Alpha 1.8E-10
Unidentified Beta
& Gamma 1.8E-09
Tritium 8.2E-01
Building 968
Tr i t i urn
l.OE-02
from a neutron generator housed in building 805. The building
ventilation system is not equipped with any radioactive waste
treatment system.
Livermore Site
Laboratories at the Livermore Site include those for
electronics, telemetry, nucleonics, optical electronics, powder
metallurgy, hydrogen effects, and microelectronics.
-------
3.10-3
Two buildings at this site release radioactive airborne
effluents, building 913 and building 968. In building 913 assembly
and dissasembly of radioactive materials are conducted along with
cutting and polishing operations. The process air from these
operations is passed through HEPA filters before being released and
the stacks are sampled for alpha and beta radiation.
Building 968, the Tritium Research Laboratory, is designed
for experiments involving kilocurie quantities of tritium. The
facility uses containment and cleanup capabilities. Tritium from
glovebox operations and vacuum pump exhaust is removed by a gas
purification system. This is accomplished by catalytic or chemical
oxidation of the tritium gas to the water form, then collecting it
on a molecular sieve bed.
3.10.3 Emissions of Radionuclides
Radioactive releases to the atmosphere from Sandia include
argon-41 from Area V reactors and tritium from research activities
in both Areas I and V. Table 3.10-1 summarizes the releases for
1977.
3.10.4 Health Impact Assessment of Sandia Laboratories
The maximum dose to a hypothetical individual at the
Albuquerque site boundary from atmospheric releases of radioactivity
in 1977 was 2.0E-3 millirem from tritium releases from Technical
Area I, and l.OE-3 millirem from argon-41 released from Technical
Area V. These doses correspond to an estimated lifetime risk of
fatal cancer of 2.8E-8 and 1.4E-8, respectively.
The dose to the population within 80 kilometers from the 1977
airborne effluents from the Albuquerque site was 5.5E-2 person-rem,
primarily from argon-41. An estimated 1.1E-5 fatal cancers would
result from each year of operations at the site.
The radioactive atmospheric emissions for the Sandia
Laboratories, Livermore Site, are small when compared to the
emissions for the Lawrence Livermore Laboratory. They would not
contribute to any significant increase to that dose resulting from
Lawrence Livermore Laboratory operations which is reported in
Section 3.5 of this report.
-------
3.10-4
Table 3.10-2. Annual radiation doses from radioactive emissions at
Sandia Laboratories, 1977
Maximum individual
Source Site boundary Populations
(mrem/y) (Person-rern/y)
Albuquerque Site:
Technical Area I
Tritium 2.0E-3 2.2E-5
Technical Area V
Argon-41 l.OE-3 5.5E-2
aThe population is within 80 kilometers.
Table 3.10-3. Individual lifetime risks and number of fatal
cancers from radioactive emissions, Sandia Laboratories, 1977
Individual lifetime risksa Expected fatal cancers
Source Maximum Average individual per year of operationb
individual in the region (Fatal cancers)
Tritium 2.8E-8 1.2E-14 4.4E-9
Argon-41 1.4E-8 3.5E-11 1.1E-5
aThe maximum individual is at the site boundary.
population is within 80 kilometers.
-------
3.10-5
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report Nos. 02 and 05, Narrative Summary Data Base Master List and
Narrative Summary Data Base Master List Updated Reports for 1977.
(EIS 02 and EIS 05), (Computer Printout).
DOE78 Department of Energy, 1978, Draft Environmental Impact
Statement, Livermore Site, Livermore, California, DOE/EIS-0028-D.
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report (EIS 51), (Computer
Printout).
Si78 Simmons Theodore N., 1978, Environmental Monitoring Report
Sandia Laboratories, 1977, SAND78-0620.
-------
3.11-1
3.11 Nevada Test Site
3.11.1 General Description
The Nevada Test Site (NTS) is a part of the National weapons
research and development program. Nuclear weapons testing and
experiments on the site are performed in conjunction with weapons
systems developed at Lawrence Livermore Laboratory, the Los Alamos
Scientific Laboratory, and the Sandia Laboratories in response to
the Department of Defense requirements.
The Nevada Test Site is located in Nye County, Nevada, about
100 kilometers northwest of Las Vegas. It occupies about 349,000
hectares of Federally-owned land. The Nell is Air Force Base and
Tonopah Test Range border the Nevada Test Site on three sides, from
the northwest to the east, providing an additional one million
hectares of Federally-owned land as a buffer. With the exception of
Las Vegas, the region surrounding the NTS is rural and sparsely
populatedonly 4,500 people live within 80 kilometers of NTS.
Beatty, with a population of 500, is the largest town nearby.
3.11.2 Process Description
All nuclear weapons detonations at the NTS, have been
conducted underground since the Limited Test Ban Treaty in 1963. As
of January 1, 1977, there had been 289 announced underground nuclear
tests. Underground detonations contain the large amounts of
radioactive material created in the cavity formed by the explosion.
Since 1971, there have been no prompt ventings or inadvertent
releases of gaseous radioactivity from test explosions. After the
test detonations, one or more re-entry holes are drilled back into
the radioactive debris for samples to determine the performance of
the device.
During sample recovery operations, small quantities of
volatile radionuclides, primarily xenon-133, are brought to the
surface and released to the atmosphere. In order to minimize
releases, a system is used to either force the effluent gases back
into the drill hole or to pass the effluent through prefilters,
charcoal filters, and HEPA filters prior to release to the
atmosphere.
3.11.3 Emissions of Radionuclides
During 1977, re-entry drilling operations resulted in
occasional low-level releases of airborne radioactivity, primarily
radioxenon. There was also some small leakage of tritium to the
-------
3.11-2
atmosphere from the waste tritium storage area. Table 3.11-1
details the quantities of radionuclides released to the atmosphere
in 1977.
Continuous low-level releases of tritium and krypton-85 occur
at NTS. These radionuclides may seep to the surface from the sites
of underground testing. Tritium in drainage ponds is released by
evaporation; the amounts from seepage and evaporation are not
quantified, but are detected at on-site sampling stations and
sometimes at locations outside of the NTS.
Table 3.11-1. Atmospheric emissions of radionuclides,
Nevada Test Site, 1977 (DOE77b)
Emission
Radionuclide (Ci/y)
Tritium 4.1
Iodine-131 2.6E-6
Xenon-133 4.6E+1
Xenon-133m 6.2E-1
Xenon-135 8.5E-1
3.11.4 Health Impact Assessment of Nevada Test Site
The only radionuclide attributed to NTS operations was
xenon-133 which was detected at Beatty, Diablo, Hiko, Las Vegas and
Tonopah, Nevada. The highest levels detected resulted in a maximum
annual whole-body dose of 2.5 microrem at Beatty (table 3.11-2).
The maximum individual lifetime risk of fatal cancer is
estimated to be about 3.5E-8. There would be about 7.2E-5 fatal
cancers per year of operation to the populations listed in Table
3.11-2.
-------
3.11-3
Table 3.11-2. Estimated annual radiation doses from
xenon-133 concentrations, 1977 (EPA78)
Dose Population Population
Location Population equivalent dose commitment dose commitment3
(microrem/y) (person-rem/y) (person-rem/y)
Beatty, NV
Diablo, NV
Hiko, NV
Las Vegas,
NV
Tonopah, NV
Total
500
6
60
b370,500
2,000
2.5
1.2
1.1
1.0
1.4
1.3E-3
7.2E-6
6.6E-5
3.6E-1
2.8E-3
3.6E-1
1.3E-3
0.0
0.0
0.0
0.0
1.3E-3
the population within 80 kilometers of NTS.
Las Vegas and nearby communities within Clark County.
Table 3.11-3. Individual lifetime risks and number
of fatal cancers from Nevada Test Site operations, 1977
Individual lifetime risks
Radio- Nearby resident Average
nuclide in Beatty, NV individual3
Expected fatal cancers
per year of operation3
(Fatal cancers)
Xenon-133
3.5E-8
3.6E-8
7.2E-5
3To the population at the locations listed in table 3.11-2.
-------
3.11-4
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information
System Report Nos. 02 and 05, Narrative Summary Data Base Master
List and Narrative Summary Data Base Master List update Records
Report 1977, (EIS 02 and EIS 05).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, (EIS 51).
EPA78 Environmental Protection Agency, 1978, Off-Site Environmental
Monitoring Report for The Nevada Test Site and Other Test Areas
Used for Underground Nuclear Detonations January through December
1977.
ERDA77 Environmental Research and Development Administration,
1977, Final Environmental Impact Statement, Nevada Test Site, Nye
County, Nevada, ERDA-1551.
-------
3.12-1
3.12 Argonne National Laboratory
3.12.1 General Description
Argonne National Laboratory (ANL) is a multidisciplinary
research and development laboratory conducting a broad program of
research in the physical, biomedical and environmental sciences. It
is also an important center for nuclear and non-nuclear energy
research and development.
Argonne National Laboratory is located in Dupage County,
Illinois, about 43 kilometers southwest of downtown Chicago. The
laboratory facilities are located on the central portion of a
1513-hectare site. Approximately 8.1 million people live within 80
kilometers of ANL.
3.12.2 Process Description
The principal sources of atmospheric emissions of radioactive
materials from ANL are the CP-5 and Janus research reactors. Other
nuclear facilities of ANL include a critical assembly or zero power
reactor, the Argonne Thermal Source Reactor, a 12.5 GeV proton
accelerator, 60-inch cyclotron, several other particle accelerators,
chemical and metallurgical plutonium laboratories, and several hot
cells and laboratories designed for work with irradiated fuel
elements.
The CP-5 reactor is a 5 MW fully-enriched heavy water,
general purpose research reactor. It uses a helium cover gas system
to provide an inert atmosphere for the reactor and to maintain
isotopic purity of the D20 in the primary system. Neutron
activation of room cooling air and tritium leaks are the main
sources of effluents from the CP-5 reactor. These effluents are
released without treatment.
The Janus reactor is a 200 KW light water biological research
reactor which is fueled with fully enriched uranium. Neutron
activation of the atmosphere in the high dose room is the source of
radioactivity released in effluents from this facility. No waste
treatment system is in use for radioactive airborne effluents at
this facility.
The hot cells, where nuclear reactor material examinations
are conducted, are equipped with two stages of HEPA filters. These
HEPA filters are not effective in reducing radioactive gases that
may be in the effluent from these cells.
-------
3.12-2
At one building, where uranium grinding activities are
conducted, the airborne effluent stream is treated with a wet
scrubber and a electrostatic precipitator.
3.12.3 Emissions of Radionuclides
Argon-41 from reactor operations was the principal
radionuclide contributing to radioactive emissions to the atmosphere
from Argonne National Laboratory in 1977. Table 3.12-1 shows the
quantities released in 1977 from ANL.
Table 3.12-1. Atmospheric emissions of radionuclides
from Argonne National Laboratory, 1977 (DOE77b)
CP-5
Source
Reactor
Radionuclide
Argon-41
Tritium
Emissions
(Ci/y)
3.0E+4
8.5E+2
Janus Reactor Argon-41 3.4
Hot Cell
Various
s
laboratoriess
Krypton-85
Plutonium-239
Antimony- 125
Tr i t i urn
Krypton-85
Uranium-238
1.4E+1
1.3E-8
4.9E-5
6.5E-1
2.0E-1
1.9E-7
3.12.5 Health Impact Assessment of Argonne National Laboratory
The maximum doses from airborne emissions in 1977 occurred to
the northeast of Argonne National Laboratory. The maximum site
boundary doses from argon-41 and tritium were 5.6 millirem and 0.01
millirem; the dose to the nearest fulltime resident was 3.0
millirem. The effluents from the CP-5 Reactor resulted in a
population dose for 1977 of 178 person-rem.
-------
3.12-3
Table 3.12-2. Annual radiation doses from radioactive
emissions from the CP-5 reactor at ANL (Go78)
Radionuclide
Argon-41
Tritium
Site boundary
dose^
(mrem/y)
5.6
l.OE-2
Nearest
resident5
(mrem/y)
3.0
7.0E-3
Population
doseb
(person-rem/y)
177
1.1
Total 178
aThese doses are not additive since they occur at different
locations.
bWithin 80 kilometers.
Table 3.12-3 estimates the individual lifetime risks and the
number of fatal cancers resulting from these doses. The lifetime cancer
risks to the maximum individual at the site boundary and to the nearest
fulltime resident are estimated to be 7.8E-5 and 4.2E-5, respectively.
The estimated number of fatal cancers to the population within 80
kilometers per year of ANL operation is estimated to be about 3.6E-2.
Table 3.12-3. Individual lifetime risks and number of fatal cancers^
from radioactive emissions from the CP-5 reactor at ANL, 1977
Individual lifetime risks Expected fatal cancers
Site Nearest fulltime Average per year of operation
Source boundary resident individual (Fatal cancers)
CP-5
reactor 7.8E-5 4.2E-5 3.1E-7 3.6E-2
aTo the population within kilometers.
-------
3.12-4
REFERENCES
DOE77a U.S. Department of Energy, 1977, Effluent Information System
Report Nos. 02 and 05, Narrative Summary Data Base Master List and
Narrative Summary Update Records Report for Calendar Year 1977, (EIS
02 and EIS 05).
DOE775 U.S. Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year 1977,
(EIS 51).
Go78 Golchert, N.W., T.L. Duffy, and J. Sedlet, 1978, Environmental
Monitoring At Argonne National Laboratory, Annual Report for 1977,
ANL-78-26.
-------
3.13-1
3.13 Brookhaven National Laboratory
3.13.1 General Description
Brookhaven National Laboratory (BNL) is a multidisciplinary
scientific research center which supports a wide variety of
scientific, research, and development programs.
The laboratory facilities are located at Upton, in Suffolk
County, New York, about 113 kilometers east of New York City. The
BNL facilities occupy the central 405 hectares of a wooded
2130-hectare site. The principal population centers, located along
the shoreline of Long Island, are within 16 kilometers.
Approximately 5.1 million people live within 80 kilometers of
Brookhaven.
3.13.2 Process Description
Among the major facilities at Brookhaven National Laboratory
are the High Flux Beam Reactor (HFBR), the Tandem Van de Graaff
Accelerator, the Brookhaven Medical Research Reactor (BMRR), the 200
MeV Proton Linear Accelerator (Linac), which is operated in
conjunction with the Alternating Gradient Synchrotron (AGS) and the
Brookhaven Linac Isotope Producer Facility (BLIP).
The High Flux Beam Reactor (HFBR) is a 40 MW heavy water fully
enriched, uranium research reactor. It is used to provide intense
beams of neutrons for research. The HFBR has an inert cover gas
system to maintain purity of the heavy water coolant. At the end of
each operating cycle (10 times per year), the purging of the cover
gas releases tritium-bearing effluents from the 98-meter HFBR
stack. Tritium is also released from valve and pump seals from
evaporation during refueling. Airborne radioactivity produced
during routine operations is vented through HEPA and activated
charcoal filters before being discharged from the stack.
The hot area of the Hot Laboratory consists of five semihot
cells, three chemical processing hot cells and three high-level hot
cells for handling multicurie amounts of radioactive materials.
Each cell is equipped with its own exhaust air filter as well as a
backup HEPA filter in the exhaust line leading to the stack. The
process cells have a separate exhaust air system which use a NaOH
scrubber and charcoal filter to remove radioiodines. The hot cells
exhaust the airborne effluents to the HFBR stack.
-------
3.13-2
The Tandem Van de Graaff accelerator started operations in
1970. Each Van de Graaff electrostatic accelerator can accelerate
atomic particles up to 10 MeV. The maximum combined energy that can
be achieved is 30 MeV. Accelerator produced radiation in the
ventilation air includes trace quantities of the short-lived gases
carbon-11, nitrogen-13 and oxygen-15. The Tandem Van de Graaff
accelerator building air is vented to the HFBR stack.
Concentrations of radioactive gases from the accelerator are not
detectable by the stack gas monitor.
The Medical Research Reactor is a 5 MW tank-type reactor which
uses 12 weight percent fully-enriched uranium fuel. The
water-cooled reactor core is within an aluminum vessel surrounded by
an air-cooled graphite reflector and biological shield. The upper
surface of the primary cooling water in the reactor vessel is in
contact with the air cooling the reflector. The cooling air picks
up gaseous contaminants from the water. These gases along with the
neutron activated gas, argon-41, are exhausted through HEPA and
charcoal filters and released from a 46-meter stack.
The Alternating Gradient Synchrotron (AGS) accelerates protons
up to 33 GeV. This accelerator is used for ultra-high energy
particle physics research. Protons for the AGS system originate
from a Cockcroft-Walton generator which gives the protons an initial
energy of 750 KeV. These protons are then injected into a linear
accelerator (Linac) which accelerates them up to 200 MeV. The
proton beam is then injected into the 0.8 kilometer circular path of
the AGS vacuum chamber. The resultant beam can then be bent to
strike a target or deflected out of the ring into experimental
areas. Because the proton beam is highly focused and in a vacuum,
there is minimal activation of air in the surrounding tube.
Carbon-11, nitrogen-13 and oxygen-15 are the predominant nuclides
produced. Release of these short-lived activation gases is minimal
since most of the tunnel air is recirculated.
The 200 MeV Linac is also used in conjunction with the
Brookhaven Linac Isotope Producer Facility (BLIP) and the
infrequently used Chemistry Linac Irradiation Facility (CLIF).
Targets for radionuclide production at the BLIP facility are
irradiated at the bottom of a 10-meter, 2.4-meter diameter,
water-filled tank. The targets are sealed to prevent escape of
radioactivity during normal operations. Several radioactive gases
are induced by the incident protons in the target cooling water.
These include tritium, nitrogen-13, oxygen-14, oxygen-15 and
-------
3.13-3
nitrogen-16. The radioactive oxygens have larger release rates in
relation to production rates than do the other radioactive gases
because they are swept out with the absorbed oxygen in the cooling
water by the radiolytic formation of stable oxygen. Several
radiocarbons are formed as a result of proton interaction with the
water. However, release of these radiocarbons as gaseous carbon
dioxide is minimized by keeping the cooling water alkaline with
NaOH. The airborne effluents from the BLIP Facility undergo HEPA
filtration prior to monitoring and release from an 18-meter stack.
A number of lower energy accelerators are used for medium
energy physics experiments at BNL. These accelerators include a 3
MeV Dynamitron (electron accelerator), a 60-inch cyclotron (30 MeV),
the Febetron (1.9-MeV pulsed accelerator) and three Van de Graaff
accelerators which range in energy from 2 to 6 MeV. The principal
atmospheric emissions come from the physics research Van de Graaff
which is used to accelerate tritium to 3.5 MeV. Effluent from the
vacuum system is passed through the scrubber where tritium gas is
converted to the oxide which is then trapped in a desiccant. When
the system is first pumped down, the flow exceeds the capacity of
the recombiner and is routed to a bypass. Fifty percent of the
tritium used is trapped by the desiccant while the remainder is
released from an 18-meter stack. The other accelerators are
operated only intermittently and produce short-lived gases which are
released with room ventilation air at roof level.
In addition to the major facilities already mentioned, there
are some 482 laboratory hoods located within various buildings at
Brookhaven. These hoods discharge air through pipes to vents on the
roof. Those hoods, where toxic agents and millicurie amounts of
radionuclides are handled, are equipped with HEPA filters or
scrubbers to control emissions.
3.13.3 Emissions of Radionuclides
The principal releases from Brookhaven National Laboratory are
tritium and argon-41 from the reactors along with tritium and
short-lived activation gases from accelerator operations. Tritium
is the only airborne effluent detectable off-site. Table 3.13-1
shows the radioactive atmospheric releases from BNL in 1977.
-------
3.13-4
Table 3.13-1. Atmospheric emissions from radionuclides,
Brookhaven National Laboratory, 1977 (DOE77b)
Source
Radionuclide
Emissions
(Ci/y)
High Flux
Beam Reactor
Medical Research Center
Building 490
Medical Reactor
Research Van de Graaff
Accelerator
Linac Isotope
Producer Facility
(BLIP)
Chemistry
Building-555
Tr i t i urn
Unidentified
Beta & Gamma
Xenon-127
Tritium
Argon-41
Tr i t i urn
Tr i t i urn
Oxygen-15
Tritium
1.2E+2
1.1E-5
1.1
1.4
3.6E+2
l.OE+3
8.4E-2
6.7E+4
7.1E+1
3.13.4 Health Impact Assessment of Brookhaven National Laboratory
Oxygen-15 and argon-41 are released in significant quantities
from Brookhaven National Laboratory. However, because of their
short half-life, dilution with ambient air and the distance to the
site boundary, the concentration of activity in the air is reduced
to a level at which there was no detectable increase in dose
equivalent at the site boundary.
Concentrations of tritium in air above background levels
resulted in a maximum total-body dose at the site boundary of 0.16
millirem/y (table 3.13-2); this dose occurred about 2500 meters to
the southwest of the HFBR stack. The maximum dose corresponds to a
lifetime risk of fatal cancer of 2.2E-6 (table 3.13-3).
-------
3.13-5
The dose commitment to the population within 80 kilometers from
atmospheric emissions of tritium from BNL in 1977 was 19.2
person-rem. This would correspond to about 3.8E-3 fatal cancers per
year of operations at Brookhaven National Laboratory (table 3.5-3).
Table 3.13-2. Radiation doses from atmospheric emissions of
radionucTides from Brookhaven National Laboratory,
1977 (DOE77b)
Maximum individual
Crital organ & Site boundary Nearest resident Population^
Radionuclide (mrem) (mrem) (person-rem)
Total body
Tritium 0.16 NR 19.2
a~Tota1 body dose to the population within 80 kilometers.
NR Not reported.
Table 3.13-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Brookhaven National Laboratory, 1977
Individual lifetime risks Expected fatal cancers
Source Site Average individual per year of operation^
boundary Regiona (Fatal cancers)
BNL 2.2E-6 5.3E-8 3.8E-3
aThe region extends to 80 kilometers.
bTo the population within 80 kilometers.
-------
3.13-6
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
(EIS) Report No. 02, Narrative Summary Data Base Master List,
(EIS-02) (Computer Listing).
DOE77b Department of Energy, 1977, Effluent Information System
(EIS) Report No. 51, Release Point Analysis Report for Calendar
Year 1977, (EIS-51) (Computer Listing).
ERDA77 Energy Research and Development Administration 1977, Final
Environmental Impact Statement, ERDA-1540, Brookhaven National
Laboratory, Upton, New York.
Na78 Naidu, J.R. Ed., 1978, 1977 Environmental Monitoring Report,
Brookhaven National Laboratory, BNL-50813, UC-11.
-------
3.14-1
3.14 Oak Ridge Facilities
3.14.1 General Description
The three major facilities at the U.S. Department of Energy
Reservation at Oak Ridge, Tennessee, are the Oak Ridge National
Laboratory (ORNL), the Oak Ridge Gaseous Diffusion Plant (ORGDP),
and the Y-12 Plant. Two smaller facilities located on the
reservation are the Comparative Animal Research Laboratory and the
Oak Ridge Associated Universities.
The Oak Ridge Reservation is located in a valley between the
Cumberland and Great Smokey Mountains in eastern Tennessee and
consists of approximately 15,000 hectares of government-owned land.
The area is bordered to the south and west by the Tennessee Valley
Authority's (TVA) Melton -Hill and Watts Bar Reservoirs on the Clinch
River and to the north by the city of Oak Ridge (figure 3.14-1).
The surrounding area is rural with the largest population center,
Knoxville, Tennessee, about 24 kilometers to the east. About
689,000 people live within 80 kilometers of the Oak Ridge National
Laboratory.
3.14.2 Process Description
Oak Ridge National Laboratory
The Oak Ridge National Laboratory is a multidiscipline
research laboratory; its mission is the discovery of new knowledge
in all areas related to energy. Nuclear energy research facilities
consist of nuclear reactors, chemical plot plants, research
laboratories, radioisotope production laboratories and support
facilities.
The central radioactive gas disposal facilities release
tritium, iodine-131, and noble gases, krypton and xenon from
radioisotope separations, reactor operations, and handling
radioactive material in hot laboratories and chemistry
laboratories. The gases undergo HEPA filtration at their source
prior to discharge to the system. The stack is constantly monitored
and sampled.
The stack servicing the High Flux Isotope Reactor and the
Transuranic Processing Plant releases fission product gases
resulting from the chemical separation of curium and californium and
from reactor operations. Process effluent gases undergo HEPA
filtration.
-------
3.14-2
LAKE V
CITY V
-''NORTH
CAROLINA
MISSISSIPPI { ALABAMA I GEORGIA
OLIVER
SPRINGS
WHITE OAK CREEK
KINGSTON
>TENNESSEE RIVER
METL10N HILL DAM
WHITE OAK
LAKE
MELTON HILL
DAM
LENOIR CITY
Figure 3.14-1. The Department of Energy Reservation
at Oak Ridge, Tennessee
-------
3.14-3
Isotope separations and chemistry laboratory operations are
the principal source of effluents. Uranium and plutonium are
present in airborne effluent from the electromagnetic isotope
separations facility. There are 14 exhaust points from this
facility. All effluents are exhausted through one or two stages of
HEPA filtration. Oil traps are also used.
A tritium target fabrication building releases small amounts
of tritium from target preparation operations.
HEPA filters are used to reduce particulate activity from the
transuranic research and the metal and ceramics laboratories. The
effluents are monitored for alpha activity.
Oak Ridge Gaseous Diffusion Plant
The Oak Ridge Gaseous Diffusion Plant , a complex of
production, research, development and support facilities, has the
primary function to enrich uranium hexaflouride (UF5) in the
uranium-235 isotope.
The principal sources of release from OR6DP are the drum
dryers in the decontamination facilities, which are in the uranium
system, and the purging of light contaminants from the purge
cascade. During 1977 the old purge cascade which used sodium
fluoride and alumina traps to reduce emissions was replaced by a new
purge cascade vent which has a KOH gas scrubber in the emission
system.
Y-12 Plant
The Oak Ridge Y-12 Plant has four primary responsibilities:
(1) production of nuclear weapons components, (2) fabrication
support for weapons design, (3) support for the Oak Ridge National
Laboratory, and (4) support and assistance to other government
agencies. The Y-12 Plant conducts activities which include
production of lithium compounds, recovery of enriched uranium from
scrap material, and fabrication of uranium into finished parts and
assemblies. Fabrications operations include vacuum casting, arc
melting, powder compaction, rolling, forming, heat treating,
machining, inspection, and testing. Many of these procedures
release particulate activity into the room exhaust air. Laboratory
and room air exhaust systems are equipped with filtration systems
which may include prefilters, HEPA filters, or bag filters.
-------
3.14-4
Oak Ridge Associated Universities
The Oak Ridge Associated Universities conducts research in
areas such as biological chemistry, immunology, nuclear medicine,
and radiochemistry. Radionuclides are handled in encapsulated or
liquid form and the potential for producing gaseous effluents is
very small.
3.14.3 Emissions of Radionuclides
The principal radioactive atmospheric emissions are tritium,
krypton-85, and xenon-133 from the Oak Ridge National Laboratory.
Table 3.14-1 summarizes the radioactive airborne emissions from the
Oak Ridge Facilities for 1977.
-------
3.14-5
Table 3.14-1. Atmospheric emissions of radionuclides
Oak Ridge Reservation, 1977 (DOE77b)
Facility
Radionuclide
Emissions
(Ci/y)
Oak Ridge National Laboratory
Central Radioactive Gas
Disposal Ficilities
High Flux Isotope Reactor
and TRU Processing Plant
Electromagnetic Isotope
Separations Facility
Tritium Target Fabrication
Building
Transuranic Research and Metal
and Ceramics Laboratories
Oak Ridge Gaseous Diffusion Plant
Decontamination Facility
Old Purge Cascade
Tritium
Iodine-131
Krypton-85
Xenon-133
Iodine-131
Krypton-85
Xenon-133
Plutonium-239
Uranium-233
Tritium
Unidentified
Alpha
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Technetium-99
Uranium-234
Uranium-235
Uranium-236
Uranium-238
2.5E+3
1.2
6.8E+3
3.3E+4
1.8E-1
1.8E+3
8.7E+3
4.0E-6
l.OE-6
2.4E+1
2.0E-8
5.6E-4
2.0E-5
l.OE-5
3.3E-4
l.OE-6
4.3E-4
1.7E-5
7.4E-7
7.5E-5
-------
3.14-6
Table 3.14-1. Atmospheric emissions of radionuclides
Oak Ridge Reservation, 1977 (DOE77b)--continued
Facility Radionuclide Emissions
(Ci/y)
New Purge Cascade Technetium-99 l.OE-6
Uranium-234 1.4E-4
Uranium-235 5.6E-6
Uranium-236 2.5E-7
Uranium-238 2.5E-5
Y-12 Plant Uranium-234 5.8E-2
Oak Ridge Associated Universities
Carbon-14 2.5E-4
Tritium 1.4E-3
Mercury-203 5.0E-5
Iodine-131 4.6E-5
3.14.4 Health Impact Assessment of Oak Ridge Facilities
The 50-year dose commitment from 1977 atmospheric emissions to
an individual occupying the nearest residence at the site boundary
would result from inhalation. The maximum reported dose commitment
was 4.6 millirem to the lung of an Oak Ridge resident from
inhalation of uranium-234.
From milk samples taken in the immediate environs of the Oak
Ridge Reservation, a maximum food-chain pathway dose commitment was
reported to be 1.5 millirem to the thyroid and 5.5 millirem to the
bone. These doses were based on iodine-131 and strontium-90
concentrations found in milk samples at different locations.
The cumulative total body dose to the population within 80
kilometers of the Oak Ridge facilities from 1977 atmospheric
emissions was estimated to be 4.7 person-rem. This corresponds to
an estimated 9.4E-4 fatal cancers per year of operation of the Oak
Ridge facilities. Table 3.14-3 tabulates ' !ie individual lifetime
risks associated with the doses to the critical organs.
-------
3.14-7
Table 3.14-2. Radiation dosesa from atmospheric emissions of
radionucTides from Oak Ridge Facilities, 1977 (UC79)
Critical Organ
and
Radionuclide
Maximum individual
Milk stations Nearest resident
(mrem) (mrem)
Population^3
(person-rem)
Lung
Uranium-234
Bone
Strontium-90
Thyroid
Iodine-131
Total body
4.6
5.5 NR
1.5 NR
NR
NR
NR
4.7
aFifty-year dose commitment. Doses are not additive because
they are for different locations. The nearest resident is located
at the site boundary.
bSum of the total body doses to exposed individuals within
80 kilometers.
NR Not reported.
Table 3.14-3 Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Oak Ridge Facilities, 1977
Organ
Individual lifetime risks
Maximum individuals
Expected fatal cancers
per year of operation13
(Fatal cancers)
Lung
Bone
Thyroid
Total body
1.3E-5
1.2E-5
1.1E-7
NR
NR
NR
NR
9.4E-4
aRisks are not additive since they correspond to maximum
individual doses that occur at different locations.
^To the population within 80 kilometers.
NR Not reported.
-------
3.14-8
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report Nos. 02 and 05, Narrative Summary Data Base MAster List and
Narrative Summary Data Base Master list Updated Report for 1977,
(EIS 02 and EIS 05), (Computer Report).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51. Release Point Analysis Report for Calendar Year
1977, (EIS 51), (Computer Report).
UC79 Union Carbide Corporation-Nuclear Division, 1978, Environmental
Monitoring Report United States Department of Energy Oak Ridge
Facilities Calendar Year 1977, Y/UB-8.
-------
3.15-1
3.15 Portsmouth Gaseous Diffusion Plant
3.15.1 General Description
The Portsmouth Gaseous Diffusion Plant has been operated by
the Goodyear Atomic Corporation for the Department of Energy since
1954. The principal process at the plant is the enrichment of
uranium in the uranium-235 isotope by separation of uranium isotopes
through gaseous diffusion.
The Portsmouth Plant, located on 1619 hectares of Federally-
owned land in Pike County, Ohio, is situated in the south-central
part of the State about 1.6 kilometers east of the Scotia River
Valley and approximately 32 kilometers north of Portsmouth, Ohio.
The area around the plant is predominately rural with marginal farm
land and densely forested hills. Pike County is sparsely populated;
less than 600,000 people live within 80 kilometers of the plant site.
3.15.2 Process Description
The Portsmouth Gaseous Diffusion Plant cascade is the only
facility in the United States with the capability of producing a
very highly enriched uranium product (VHE = 97.65% uranium-235).
The cascade consists of 4020 isotopic stages that produce enriched
uranium-235 and 60 purge stages that separate and purge the light
gas contaminants that leak into the system. Three imposing
buildings (X-326, X-330, X-333) house the gaseous diffusion process
equipment. The principal points of release of radioactivity from
the process buildings are from the top purge vent in building X-326,
and the cold-recovery system vent in building X-330. Two other
support facilities are also sources of emissions at the site.
The purge facility separates the light contaminants (air,
N2, HF and coolant) from the process gas flow. The light
contaminant gases are passed through alumina traps to an air-jet
exhauster and discharged to the atmosphere from a roof-top vent.
Purge gases collected from locations throughout the plant are passed
through refrigerated cold traps which are a part of the X-300
cold-recovery system. These cold traps freeze out UFg from the
gases which are then passed through NaF traps for removal of
remaining traces of UFg before being released to the atmosphere
through a vent on a building roof by means of air-jet exhausters.
Building X-705, the decontamination facility, is centrally
located with respect to the three process buildings. This facility
provides space, special handling equipment, and fixtures for
-------
3.15-2
disassembly, decontamination and radiation monitoring of cascade
components that are removed for repair. The principal source of
airborne radioactive emissions is the process of converting uranium
oxide (UoOo) to uranium hexafluoride (UFg) from uranium
recovered from decontamination solutions. The UFg is then
re introduced as feed to the cascade. The UFg in the off-gas stream
is removed by a cold trap. Off-gases from the cold trap are passed
through sodium fluoride traps . Particulates in the gas stream are
removed prior to the cold trap by a sintered metal filter, and
volatile impurities are adsorbed on a magnesium flouride trap.
The remaining facility at the site that is an important
source of emissions is building X-744G which is used for receiving,
sampling, storaging, transferring and shipping of licensed and DOE-
owned uranium materials. Sampling and transferring operations are
sources of releases of UFg. The gases are passed through cold
traps and alumina traps before being vented to the atmosphere.
3.15.3 Emissions of Radionuclides
Table 3.15-1 summarizes the radioactive airborne emissions
from the four release points at the Portsmouth Gaseous Diffusion
Plant in 1977.
-------
3.15-3
Table 3.15-1. Atmospheric emissions of radionuclides from
Portsmouth Gaseous Diffusion Plant, 1977 (DOE77b)
Facility
Building X-326
Top Purge
Cascade Vent
Building X-330
Cold Recovery
System Vent
Building X-705
Building
Waste Gas Vent
Building X-744G
Oxide
Sampling Facility
Radionuclide
Protactinium-234m
Technetium-99
Thorium-234
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Protactinium-234m
Technetium-99
Thorium-234
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Protactinium-234m
Technetium-99
Thorium-234
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Emissions
(Ci/y)
4.0E-2
3.2
4.0E-2
9.5E-2
3.3E-3
1.1E-4
5.0E-4
5.9E-3
1.3
5.9E-3
2.8E-3
1.7E-4
8.3E-7
4.1E-3
9.4E-4
6.0E-5
3.1E-6
1.4E-5
1.6E-5
2.0E-3
1.6E-5
1.7E-4
1.1E-5
4.5E-7
1.1E-6
-------
3.15-4
3.15.4 Health Impact Assessment of Portsmouth Gaseous
Diffusion Plant
Table 3.15-2 summarizes the 50-year dose commitment from
radioactive airborne emissions from the Portsmouth Gaseous Diffusion
Plant. The individual doses reported were based on the maximum average
alpha and beta-gamma radioactivity concentrations obtained from air
sampler data. The maximum dose at the eastern site boundary and the
maximum off-site residence dose were considered the same since a few
families reside at the property boundary. The nearest community for
which these doses apply is the town of Piketon, about 8 kilometers
north of the plant.
Table 3.15-3 summarizes the estimated individual lifetime risk
of fatal cancer and the number of fatal cancers resulting from plant
operations.
Table 3.15-2. Radiation doses9 from atmospheric emissions of
radionuclides from Portsmouth Gaseous Diffusion Plant,
1977 (An78)
Maximum individual individual in the
Organ Site boundary nearest community^ Population0
(mrem) (mrem) (person-rem)
Lung
Bone
Kidney
G.I. tract
Total body
1.12
1.23
1.06
2.16
0.21
3.9E-2
4.9E-2
4.1E-2
8.5E-2
8.0E-3
NR
NR
NR
NR
0.19
aThe maximum individual and average individual doses are based
on measured airborne radioactivity concentrations. Doses are 50-year
dose commitments.
bpiketon is located approximately 8 kilometers from the plant.
cTotal body dose to the population within 80 kilometers.
NR Not reported.
-------
3.15-5
Table 3.15-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Portsmouth Gaseous Diffusion Plant, 1977
Individual lifetime risks
Maximum Nearest Average
Organ individual community^ individual
Region3
Expected fatal cancers
per year of operation0
(Fatal cancers)
Lung
Bone
Kidney
G.I. tract
Other soft
tissue
Total
3.1E-6
2.6E-6
7.4E-7
3.0E-6
7.4E-7
l.OE-5
1.1E-7
l.OE-7
2.9E-8
1-2E-7
2.8E-8
3.9E-7
NR
NR
NR
NR
NR
4.4E-9
NR
NR
NR
NR
NR
3.8E-5
TThe region extends to 80 kilometers.
bPiketon.
cTo the population within 80 kilometers.
NR Not reported.
-------
3.15-6
REFERENCES
An78 Anderson, Robert E., Bonnie J. Rumble, and Edgar R. Warner,
1978, Portsmouth Gaseous Diffusion Plant Environmental Monitoring
Report for Calendar Year 1977, GAT-955, Piketon, Ohio
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List, EIS 02,
(Computer Report).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, EIS 51, (Computer Listing)
ERDA77 Energy Research and Development Administration, 1977 Final
Environmental Impact Statement, Portsmouth Gaseous Diffusion Plant
Site, Piketon, Ohio, ERDA-1555, Washington, D.C.
-------
3.16-1
3.16 Paducah Gaseous Diffusion Plant
3.16.1 General Description
The Paducah Gaseous Diffusion Plant (P6DP) is a uranium
enrichment cascade plant with a uranium hexafluoride (1^5)
manufacturing plant and various other support facilities.
The Paducah Gaseous Diffusion Plant, located in McCracken
County, Kentucky, is 19 kilometers west of Paducah, Kentucky, and
about 6 kilometers south of the Ohio River. The plant is within a
300-hectare fenced area. The site is encircled by a buffer zone of
Government-owned land and beyond that, a wildlife management area.
The nearest incorporated town is Metropolis, Illinois, about 8
kilometers to the northeast.
3.16.2 Process Description
The primary plant is the Diffusion Cascade Plant. The
principal process is the separation of uranium isotopes through
gaseous diffusion. Uranium hexafluoride (UFs) gas is fed into the
system and pumped through up to 1812 stages in the enrichment
cascade. The product UF5 is enriched in the uranium-235 isotope.
The "tails" are withdrawn as UF6 depleted in uranium-235.
All the stages in the enrichment cascade are contained within
five buildings. The prime source of emissions is from the purge
cascade which is used for removal of light contaminants from the
process stream. These contaminants, which consist of isotopes of
uranium and technetium-99, are released from the diffusion cascade
building stack which is sampled regularly.
The manufacturing building or Feed Plant uses hydrogen,
anhydrous hydrogen fluoride (HF) and uranium oxide (1103) to
produce the UFs that is fed into the diffusion cascade. Gaseous
emissions, from fluorination operations of UF4 to UFe, are
passed through a series of waste treatment systems that include cold
traps, fluid bed absorbers and sintered metal filters. HEPA and bag
filters are also used to treat other emissions from the Feed Plant.
The Uranium Recovery and Chemical Processing Facility
conducts operations that involve pulverizing and screening of
uranium salts. Here bag filters are used to reduce airborne
emissions.
At the Metals Plant, depleted UF6 from the Cascade is
reacted with HF to convert it to UF4 which is more easily stored.
-------
3.16-2
Table 3.16-1. Atmospheric emissions of radionuclides
from Paducah Gaseous Diffusion Plant, 1977 (DOE77b)
Facility and Emissions
Radionuclide (Ci/y)
Diffusion Cascade
Technetium-99 6.8E-2
Uranium-234 9.0E-3
Uranium-235 3.8E-4
Uranium-236 5.0E-6
Uranium-238 3.1E-3
Feed Plant
Uranium-234 1.5E-1
Uranium-235 6.7E-3
Uranium-236 2.3E-4
Uranium-238 1.5E-1
Uranium Recovery and Chemical Processing
Uranium-234 1.6E-3
Uranium-235 1.4E-4
Uranium-236 5.1E-5
Uranium-238 1.1E-2
Metals Plant
Uranium-234 7.1E-4
Uranium-235 6.0E-5
Uranium-236 2.3E-5
Uranium-238 4.8E-3
Warehouse
Uranium-234 3.5E-2
Uranium-235 1.6E-3
Uranium-236 5.2E-5
Uranium-238 3.5E-2
-------
3.16-3
Bag filters are used for these operations to reduce emissions. Also
within the same facility UF4 can be reduced to metal by reaction
with magnesium. However, there were no such operations in 1977.
Ash which is collected at the bottom of the flame tower in
the process of fluorinating uranium oxides at the Feed Plant is
taken to a warehouse for storage. The ash, containing isotopes of
uranium, may become airborne and be released to the atmosphere.
There are no effluent control and exhaust air systems; however,
emissions are continuously sampled.
3.16.3 Emissions of Radionuclides
The radioactive airborne emissions from the Paducah Gaseous
Diffusion Plant consisted of four isotopes of uranium and
technetium-99. Table 3.16-1 summarizes the emissions for 1977.
3.16.4 Health Impact Assessment of Paducah Gaseous Diffusion Plant
Table 3.16-2 summarizes the 50-year dose commitment to the
bone and lung from inhalation of uranium emitted from Paducah
Gaseous Diffusion Plant in 1977. The maximum site boundary dose was
found to occur to the north of the site. The nearest resident is
approximately 2.2 kilometers east of the plant boundary. The
potential dose to the nearest resident would correspond to a
lifetime risk of fatal cancer of 2.8E-6. No population dose was
reported.
Table 3.16-2. Radiation dosesa from atmospheric emissions
of radioactive materials at
Paducah Gaseous Diffusion Plant, 1977 (UC78)
Organ
Lung
Bone
Maximum
site boundary
doseb
(mrem)
6.3
0.9
Nearest
resident
dosec
(mrem)
0.9
0.1
aFifty-year dose commitment.
b[_ocation--north of site.
cApproximately 2.2 kilometers east of the plant boundary.
-------
3.16-4
Table 3.16-3. Individual lifetime risks and number of fatal cancers
resulting from atmospheric emissions of radionuclides at
Paducah Gaseous Diffusion Plant, 1977
Source
Individual
Site
boundary
lifetime risk
Nearest
resident
Expected fatal cancers
per year of operation
(Fatal cancers)
Paducah Gaseous
Diffusion Plant 2.0E-4
2.8E-6
NR
Not reported.
-------
3.16-5
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System Re-
port No. 2, Narrative Summary Data Base Master List, (EIS 02),
(Computer Listing).
DOE775 Department of Energy, 1977, Effluent Information System Re-
port No. 51, Release Point Analysis Report for Calendar Year 1977,
(EIS 51), (Computer Listing).
UC78 Union Carbide Corporation, 1978, Environmental Monitoring Re-
port, United States Department of Energy, Paducah Gaseous Diffusion
Plant, Calendar Year 1977, Y/UB-9, Office of Health, Safety, and
Environmental Protect-ion, P. 0. Box Y, Oak Ridge, Tenn. 37830.
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3.17-1
3.17 Ames Laboratory
3.17.1 General Description
Ames Laboratory is operated by Iowa State University for the
Department of Energy. The principal facility is the Ames Laboratory
Research Reactor (ALRR) along with its associated radioactive waste
disposal/decontamination facility. Other facilities on campus
include offices, laboratories, warehouses and service buildings.
The reactor facility is located about 2.4 kilometers
northwest of the campus and about 4.8 kilometers west and north of
Ames on approximately 16.2 hectares of land. Ames is located in an
agricultural area of central Iowa. The population of Ames is
40,000, half of which is comprised of students. The total
population within 80 kilometers is about 590,500.
3.17.2 Process Description
The ALRR is fueled with uranium-235 and uses heavy water as a
coolant and moderator which is circulated in a closed system. The
principal uses of the 5 MW reactor, which operates on a continuous
schedule, are to produce fission products for research, to irradiate
crystals for neutron diffraction studies, to irradiate metals for
the study of radiation effects, to produce isotopes for tracer
studies, and to activate various elements for analytical studies.
Operations within the disposal/decontamination facility
include solid waste compaction, liquid storage in tanks with
filtration and ion exchange processing, evaporation and distilation,
and packaging of solid waste for shipment to a radioactive disposal
site.
Airborne emissions from reactor operations are filtered and
monitored before release.
3.17.3 Emissions of Radionuclides
The principal emissions from Ames Laboratory Research Reactor
are argon-41 and tritium. Table 3.17-1 summarizes the emissions
from Ames Laboratory for 1977.
-------
3.17-2
Table 3.17-1. Atmospheric emissions of radionuclides,
Ames Laboratory, 1977 (DOE77)
Emissions
Facility and Radionuclide (Ci/y)
Ames Laboratory Research Reactor (ALRR)
Argon-41 1.3E+4
Beryl!ium-7 9.0E-5
Cerium-141 2.9E-6
Cerium-144 1.5E-5
Cesium-137 9.1E-7
Tritium l.OE+3
Iodine-131 6.4E-7
Niobium-95 4.2E-5
Ruthenium-103 2.1E-5
Ruthenium-106 3.2E-6
Unidentified Alpha 8.3E-6
Unidentified Beta & Gamma 2.8E-5
Zirconium-95 1.3E-5
Radioactive Waste
Disposal Decontamination Facility
Tritium 9.1E-2
3.17.4 Health Impact Assessment of Ames Laboratory
The doses reported for Ames Laboratory were estimated by
applying principles of meteorological diffusion. Table 3.17.2
summarizes the dose to an average individual at the site boundary
(about 213 meters from the reactor) and to the population residing
within 80 kilometers.
An individual exposed to the total body dose reported in
table 3.17.2 would have an estimated lifetime risk of fatal cancer
of 7.7E-5. An estimated 4.9E-2 fatal cancers are estimated per year
of operation of ALRR to the population residing within 80 kilometers.
-------
3.17-3
Table 3.17-2. Annual radiation doses from atmospheric emissions of
radioanuclides from Ames Laboratory, 1977 (Vo78)
Average individual
Radionuclide Site boundary Population^
(mrem/y) (person-rem/y)
Argon-41
Tritium
5.4
0.1
4.6
240
Total 5.5 245
aTotal body dose to the population within 80 kilometers.
Table 3.17-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Ames Laboratory, 1977
Individual lifetime risks Expected fatal cancers
Organ Site Average individual per year of operation
boundary Regions (Fatal cancers)
Total body 7.7E-5 5.6E-6 4.9E-2
aThe region extends to 80 kilometers.
bTo the population within 80 kilometers.
-------
3.17-4
REFERENCES
DOE77 Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, (EIS 51), (Computer Report).
Vo78 Voss M. D., 1978, Environmental Monitoring Report at Ames
Laboratory, Calendar Year 1977, 15-4365, Ames, Iowa.
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3.18-1
3.18 Atomics International
3.18.1 General Description
Atomics International , a Division of Rockwell International
Corporation, is engaged in research and development of atomic
energy. Efforts include design, development fabrication and testing
of systems and components for central station power plants,
fabrication of nuclear fuel for test reactors, and decontamination
and disposition of facilities.
Atomics International operates two sites in California. At
the Canoga Park sitewhich is licensed by NRC and the State of
Californiaprograms utilize uranium fuel production facilities,
analytical chemistry laboratories, and a cobalt-60 irradiation
facility. The other site, the Santa Susana Field Laboratories
(SSFL), is approximately 46 kilometers northwest of Los Angeles,
California, on a 117-hectare site. Both DOE and Rockwell share this
site. The SSFL facilities, which are licensed by the NRC and the
State of California include a hot laboratory, a nuclear materials
development facility, a neutron radiography facility containing a
research reactor, and several X-radiography inspection facilities.
DOE contract activities are conducted within a 33-hectare area at
the SSFL site.
3.18.2 Process Description
DOE operations at SSFL which may release radioactive
materials into the atmosphere are conducted at the Radioactive
Material Disposal Facility (RMDF). The two buildings (021-022) that
comprise this facility are used for processing wastes generated by a
program for the decontamination and disposition of DOE facilities.
Liquid and dry radioactive wastes are processed, packaged and
temporarily stored for further disposal. Nuclear fuel material,
handled in encapsulated or encapsulated forms, contains the isotopes
uranium-234, uranium-235, uranium-236, uranium-238, cesium-137,
strontium-90 and promethium-147. Airborne emissions from this
facility are reduced by exhausting air through a HEPA filtration
system.
NRC and California State Licensed Activities include hot cell
operations conducted in Building 020. Here irradiated nuclear fuels
and reactor components are examined. Radioactive material handled
in unencapsulated form at this facility includes the following
radionuclides: thorium-232, uranium-233, uranium-234, uranium-235,
uranium-236, and uranium-238 as constituents in the various fuel
-------
3.18-2
materials; and cesium-137, strontium-90, krypton-85 and
promethium-147 as mixed fission products. Airborne emissions may
contain radioactive particulates and gases, depending on the
operations being performed.
Operations at Building 055, that are licensed by the NRC and
the State of California, involve the fabrication of plutonium and
plutonium-uranium fuel pins. The fuel may consist of depleted and
enriched uranium and plutonium materials.
3.18.3 Emissions of Radionuclides
Table 3.18-1 summarizes the radioactive emissions from the
Santa Susana Field Laboratories. The DOE emissions were less than
eight percent of the total emissions reported from all the
facilities at both Atomics International sites.
Table 3.18-1. Atmospheric emissions of radionuclides from
the Santa Susana Field Laboratories, 1977 (Mo78)
Emissions
Location/Radionuclide (Ci/y)
Buildings 021-022
Unidentified Alpha
-------
3.18-3
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List, (EIS 02).
(Computer Report).
DOE775 Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, (EIS 51), (Computer Report).
Mo78, More J. D., 1978, Atomics International Environmental
Monitoring and Facility Effluent Annual Report 1977, Canoga Park,
California.
-------
3.19-1
3.19 Battelle Columbus Laboratory
3.19.1 General Description
Battelle Columbus Laboratory (BCL) conducts various NRC-
licensed activities as well as activities under Department of Energy
contracts.
BCL operates two complexes in the Columbus, Ohio area. The
first site is the King Avenue site, which consists of four hectares
of land near a residential area in Columbus. The Ohio State
University intramural sports practice field borders the site to the
north.
The second site is the Nuclear Sciences Area of the West
Jefferson site, which is located about 27 kilometers west of the
King Avenue laboratories. This site occupies about four hectares on
a 405-hectare tract of land. There are approximately 623,400 people
living within 80 kilometers of the laboratory.
3.19.2 Process Description
The King Avenue site has a uranium-235 processing facility
located within Building 3. This building also houses the melting
facility and the powder metallurgy laboratory. The uranium
processing facility manages all transactions involving nuclear
material at the King Avenue site. However, handling of contract and
licensed material was very limited in 1977 and there have been no
reported airborne emissions since 1975.
At the West Jefferson site activities at the Nuclear Sciences
Area include: JN-1 hot cell operations where irradiated reactor
fuel elements are studied; JN-4 plutonium laboratory work, where
research is conducted on uranium-235/plutonium-239 nitride reactor
fuel; and materials accountability and storage operations, conducted
at the ON-2 vault.
Airborne radioactive emissions at the Battelle Columbus
Laboratory are first filtered at the points of operations i.e.,
glove boxes, hoods, test cells and then passed through one or two
stages of HEPA filters before release. The hot cell facility is
equipped with a charcoal bed so radioactive gases can be routed
through it when necessary.
-------
3.19-2
3.19.3 Emissions of Radionuclides
Table 3.19-1 summarizes the radioactive emissions from the
West Jefferson site in 1977. There were no reported emissions from
DOE contract activities from the King Avenue site.
3.19.4 Health Impact Assessment of Battelle Columbus Laboratory
Table 3.19.2 summarizes the estimated annual radiation doses
from radioactive emissions from the West Jefferson site of Battelle
Columbus Laboratory. The maximum dose at the site boundary is
considered coincident with the downwind position from the facility
where the highest annual concentrations of radionuclides will
occur. This point, for uncontrolled exposure, is outside the
security fence but still on BCL property. The table also shows the
maximum dose estimate for an individual in the nearest population
group, a distance of two kilometers.
Table 3.19-3 estimates the individual lifetime risks and
number of fatal cancers in the population resulting from these
doses. For each year of operation of BCL there would be an
estimated 2.1E-7 fatal cancers in the population within 80
kilometers.
-------
3.19-3
Table 3.19-1. Radioactive airborne emissions from Battelle Columbus
Laboratory, 1977 (Ev78)
Facility
JN-1 Hot Cell
JN-4 Plutonium
Laboratory
JN-2 Vault
Radionuclide
Barium- 133
Cesium-134
Cesium-137
Chromium-51
Cobalt-57
Cobalt-60
Europium- 152
Europium-154
Mercury-203
Iodine-131
Iridium-192
Krypton-85
L ant ban urn- 140
Manganese-54
Niobium- 95
Ruthenium-97
Antimony- 12
Antimony-125
Tin-113
Terbium- 160
Unidentified Alpha
Unidentified Beta
Uranium-235
Xenon-133m
Zinc-65
Zirconium-95
Plutonium-239
Plutonium-239
Emissions
(Ci/y)
6.0E-9
1.3E-7
1.3E-6
1.3E-8
3.3E-9
1.4E-4
7.9E-7
4.5E-7
4.2E-7
1.5E-7
4.0E-9
5.2E-6
2.9E-5
1.6E-7
4.0E-8
1.4E-7
2.6E-7
3.5E-5
6.0E-8
3.2E-7
1.8E-7
4.8E-6
2.0E-7
2.9E-7
9.0E-8
2.1E-7
7.2E-8
7.0E-9
-------
3.19-4
Table 3.19-2. Annual radiation doses from atmospheric emissions of
radioactive materials from Battelle Columbus Laboratory, 1977 (Ev78)
Critical
organ
Lung
Bone
K i dney
Thyroid
G. I. tract
Skin
Other soft
Maximum
Site boundary
(mrem/y)
1.1E-2
7.0E-2
1.2E-3
5.5E-5
3.4E-4
6.4E-7
tissue 1.3E-7
individual
Nearest resident
(mrem/y)
6.5E-06
4.1E-05
6.8E-07
3.2E-08
2.0E-07
3.8E-10
7.2E-11
Populationa
(person-rem/y)
1.6E-3
4.8E-3
7.9E-5
3.8E-6
2.3E-5
4.4E-8
8.4E-9
dTotal body dose to the population within 80 kilometers.
Table 3.19-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Battelle Columbus Laboratory, 1977
Critical
organ
Lung
Bone
Kidney
Thyroid
G. I. tract
Skin
Other soft
tissue
Total
Individual
Site
boundary
3.1E-08
1.5E-07
8.4E-10
3.9E-12
4.8E-10
2.2E-12
4.6E-13
1.8E-07
lifetime risks
Nearest
resident
1.8E-11
8.6E-11
4.8E-13
2.2E-14
2.8E-13
1.3E-15
2.5E-16
l.OE-10
Expected fatal cancers
per year of operation
(Fatal cancers)a
6.4E-08
1.4E-07
7.9E-10
3.8E-12
4.6E-10
2.2E-12
4.2E-13
2.1E-07
aTo the population within 80 kilometers.
-------
3.19-5
REFERENCES
Ev78 Evans R.G., 1978, Environmental Report for Calendar Year 1977
on Radiological and Non-Radiological Parameters to United States
Department of Energy Chicago Operations Office, Battelle Columbus
Laboratories, Columbus, Ohio.
-------
3.20-1
3.20 Bettis Atomic Power Laboratory
3.20.1 General Description
The Bettis Atomic Power Laboratory (BAPL) is operated by the
Westinghouse Electric Corporation for the Naval Reactors Division of
the Department of Energy. Operations involve design and development
of naval nuclear power reactors.
Bettis Atomic Power Laboratory occupies about 81 hectares of
land in West Mifflin, Pennsylvania, about 13 kilometers southwest of
Pittsburgh. The total population within 80 kilometers of the site
is about 3.3 million.
3.20.2 Process Description
Fuel development efforts and hot cell chemistry work are the
principal operations contributing to radioactive airborne emissions
from BAPL. HEPA filters are used to reduce particulates in airborne
effluents, and charcoal filters are used where necessary to reduce
gaseous emissions. Air exhausted from the laboratory's fume hoods
is treated by filtration or wet scrubbers to minimize emissions.
Exhaust stacks which discharge air from areas where radioactive
materials are handled are continiuosly monitored.
3.20.3 Emissions of Radionuclides
Table 3.20-1 summarizes the radioactive airborne emissions
from Bettis Atomic Power Laboratory for 1977.
3.20.4 Health Impact Assessment of Bettis Atomic Power Laboratory
Table 3.20-2 summarizes the annual doses from airborne
emissions from Bettis Atomic Power Laboratory in 1977. Bone was the
critical organ for particulate activity and the lung was the
critical organ for gaseous activity.
Table 3.20-3 estimates the individual lifetime risks of fatal
cancer to a hypothetical individual at the site boundary and to the
average individual in the region within 80 kilometers. The table
also estimates the number of fatal cancers that would occur per year
in the population within 80 kilometers for each year of operation of
Bettis Atomic Power Laboratory.
-------
3.20-2
Table 3.20-1. Atmospheric emissions of radioanuclides
from Bettis Atomic Power Laboratory, 1977 (DOE77b)
Facility and Emissions
Radionuclide (Ci/y)
Main Laboratory and NE Area
Tritium 4.8E-5
Iodine-131 1.1E-4
Krypton-85 8.2E-1
Antimony-125 1.4E-4
Unidentified Alpha 2.2E-5
Unidentified Beta
and Gamma 2.9E-4
Table 3.20-2. Annual radiation doses from atmospheric emissions of
radionuclides from Bettis Atomic Power Laboratory, 1977 (We78)
Maximum individual Average individual
Crital Site boundary Region Populationa
organ (mrem/y) (mrem/y) (person-rem/y)
Bone <1.0
Lung <1.0
Total body <0.1
3.7E-3
NR
6.1E-5
12.3
NR
0.2
°To the population wit rim 80 Kilometers.
NR Not reported.
-------
3.20-3
Table 3.20-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Bettis Atomic Power Laboratory, 1977
Individual lifetime risks Expected fatal cancers
Critical Maximum Average individual per year of operation
Organ individual Region^ (Fatal cancers)^
Bone 2.1E-6
Lung 2.8E-6
Total body 1.4E-6
7.8E-09
NR
8.5E-10
3.7E-4
NR
4.0E-5
°The region extends to 80 kilometers.
"To the population within 80 kilometers.
NR Not reported.
-------
3.20-4
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base List, EIS02. (Computer
Report).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, EIS 51, (Computer Report).
WE78 Westinghouse Electric Corporation, 1978, Effluent and
Environmental Monitoring Report for Calendar Year 1977,
WAPD-RSC(HE)-460, Bettis Atomic Power Laboratory, Pittsburgh,
PA.
-------
3.21-1
3.21 Feed Materials Production Center
3.21.1 General Description
The Feed Materials Production Center (FMPC) produces purified
uranium metal and compounds for use at other DOE sites.
The FMPC is located on a 425-hectare site about 32 kilometers
northwest of Cincinnati, Ohio. The area east of the site is in the
Miami River Flood plain which is primarily used for farming; the
area downriver is sparsely populated with small and scattered
industries. The total population within 80 kilometers of the FMPC
site is about 2.5 million.
3.21.2 Process Description
Uranium production may begin with ore concentrates, recycled
uranium from spent fuel, or with various uranium compounds. Impure
starting material is dissolved in nitric acid and the uranium is
extracted into an organic liquid and then back-extracted into dilute
nitric acid to yield a solution of uranyl nitrate.
Evaporation and heating convert the nitrate solution to
uranium trioxide (UO^) powder. This compound is reduced to
uranium dioxide (U02J with hydrogen and then converted to uranium
tetrafluoride (UF4) by reaction with anhydrous hydrogen fluoride.
Uranium metal is produced by reacting UF4 and magnesium metal in a
refractory-lined reduction vessel. This primary uranium metal is
then remelted with scrap uranium metal to yield a purified uranium
ingot which is extruded to form rods or tubes. Sections are then
cut and machined to final dimensions. These machined cores are
shipped to other DOE sites for canning and final assembly into
reactor fuel elements.
Periodically, small amounts of thorium are processed at the
Center. Thorium production steps, in general, are similar to those
followed in uranium production. Final products may be purified
thorium nitrate solution, solid thorium compounds, or metal (Bo78).
The eight buildings or plants at the FMPC which carry out the
operations just described are equipped with cloth type bag filters
to reduced atmospheric emissions.
-------
3.21-2
Table 3.21-1. Atmospheric emissions of radionucTides,
Feed Materials Production Center, 1977 (DOE77a, DOE77b)
Facility
Plant 1
Plant 2
Plant 4
Plant 5
Plant 6
Plant 8
Plant 9
Process or Operation
Material Sampling and
Grinding
Dumping Dry Feeds and
Feeding Digest Tanks
UF4 Production and
Repackaging
Metal Production and
Slag Grinding
Machining of
Uranium Metal
Dumping, Milling,
Production of Uranium Metal
Emissions
(Ci/y)
..
2.0E-3
4.0E-3
1.9E-2
~
1.7E-3
Pilot Plant
Remelting and Machining
Production of Thorium and
Uranium Compounds and Metal
Total
3.3E-3
3.0E-2
3.21.3 Emissions of Radionuclides
Table 3.21.1 summarizes the emissions of radioactivity into
the atmosphere in 1977 (DOE77b) from the FMPC and describes the
operations (DOE77a) associated with the emissions. The uranium-235
content of the uranium handled at the Feed Materials Production
Center may he depleted, normal, or slightly enriched. However, the
average content is close to that of natural uranium.
-------
3.21-3
3.21.4 Health Impact Assessment of the Feed Materials
Production Center
In 1977 the highest average concentration of airborne uranium
occurred at the air sampling station at the eastern boundary of the
site. Table 3.21-2 summarizes the doses resulting from 1977
emissions. The maximum dose to the hypothetical individual at the
site boundary is estimated to be 4.8 mrem to the lung, while the
nearest resident who lives near the sampling station is estimated to
receive a lung dose of 2.8 mrem. The airborne uranium would result
in a 50-year whole body dose commitment of 1.4 person-rem to the
population within 80 kilometers.
Table 3.21-3 summarizes the individual lifetime risks and
number of fatal cancers resulting from these doses.
The Town of Ross, the nearest community to the Feed Materials
Production Center, is approximately 4 kilometers to the northeast of
the production area. Table 3.21-4 summarizes the doses and lifetime
risks of cancer to an individual living in that community.
Table 3.21-2. Radiation dosesa from atmospheric emissions of
radionuclides from the Feed Materials Production Center, 1977 (Bo78)
Maximum individual
Crital Site boundary Nearest resident Population^
organ (mrem) (mrem) (person-rem)
Lung 4.8 2.8 NR
Total body NR NR 1.4
aFifty-year dose commitment.
"Total body dose to the population within 80 kilometers.
NR Not reported.
-------
3.21-4
Table 3.21-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from the Feed Materials Production Center, 1977
Individual lifetime risks Expected fatal cancers
Critical Site Nearest per year of operation
organ boundary resident (Fatal cancers)9
Lung 1.3E-5 7.8E-6
Total body NR NR
NR
2.8E-4
aTo the population within 80 kilometers,
NR Not reported.
Table 3.21-4. Radiation doses9 and individual lifetime risks
of fatal cancers to residents of the Town of Ross due to
radioactive atmospheric emissions from the
Feed Materials Production Center, 1977
Organ
Lung
Bone
K i dney
Other soft tissue
Average
individual
(mrem)
0.65
0.22
0.36
0.03
Individual lifetime ri
of fatal cancer to a
resident of Ross
1.8E-6
4.6E-7
2.5E-7
1.1E-7
sks
Total
2.6E-6
dFifty-year dose commitment.
-------
3.21-5
REFERENCES
Bo78 Boboak M. W., K. N. Ross, and D. A. Fuchs, 1978, Feed
Materials Production Center Environmental Monitoring Annual Report
for 1977, NLCO-1151, Cincinnati, Ohio.
DOE77a Department of Energy, 1972, Effluent Information System
Report Report No. 02. Narrative Summary Data Base Master List, EIS
02, (Computer Report).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, EIS 51, (Computer Report).
-------
3.22-1
3.22 Knolls Atomic Power Laboratory
3.22.1 General Description
The Knolls Atomic Power Laboratory (KAPL) consists of three
sites: the Knolls Site, the Kesselring Site, and the Windsor Site.
At the Knolls Site, the principal function is the development of
nuclear power plants, while at the Kesselring and Windsor sites the
principal function is training personnel in the operation of nuclear
reactors.
The Knolls Site is situated on 69 hectares of land along the
Mohawk River, about eight kilometers east of Schenectady, N. Y., in
a relatively low population, residential area. About 1.23 million
people live within 80 kilometers of Knolls Site.
The Kesselring Sitea 1579-hectare siteis located near
West Milton, NY., approximately 24.4 kilometers north of
Schenectady. The surrounding area is rural and sparsely populated;
about 1.08 million people live within 80 kilometers.
The Windsor Site consists of only 4 hectares of land near
Windsor, Connecticut, about eight kilometers north of the city of
Hartford. The area is a rural farming and industrial region along
the Farmington River. Approximately 3.1 million people live within
80 kilometers.
3.22.2 Process Description
The principal operations which are potential sources of
emissions of radioactive materials into the atmosphere at the Knolls
Site include: hot cell operations, the use of critical assemblies,
the handling of radioactive materials in chemistry and physics
laboratories, and the operation of a uranium metallurgy laboratory.
These operations, along with any other activities involving
radioactive materials, are serviced by controlled exhaust systems
which discharge through elevated stacks. Exhaust air is passed
through HEPA and carbon filters and continuously sampled prior to
release.
The Kesselring site has three pressurized water nuclear
reactor plants and associated support facilities. Particulate and
gaseous activity contained in the primary coolant may become
airborne from reactor coolant discharges, sampling operations and
during laboratory operations. Exhaust air from these operations is
passed through HEPA filters, monitored, and released from elevated
stacks.
-------
3.22-2
The Windsor Site contains one pressurized water nuclear
reactor plant used for training. As at the Kesselring Site, the
exhaust air from operations which have the potential for release of
airborne radioactivity is passed through HEPA filters, then
monitored, and released through elevated stacks.
3.22.3 Emissions of Radionuclides
Table 3.22-1 summarizes the emission of radioactivity to the
atmosphere from the three Knolls Atomic Power Laboratory sites in
1977.
3.22.4 Health Impact Assessment of Knolls Atomic Power Laboratory
Tables 3.22-2 summarizes the doses from emissions of
radioactivity from Knolls Atomic Power Laboratory operations. The
estimated doses reported in this table include the doses from both
airborne and liquid radioactive emissions. The monitoring report
did not describe the nearest population group nor indicate where the
maximum individual dose occurred.
The lifetime risks of fatal cancer to the maximum exposed
individual at each site and to the average individual in the region
are summarized in table 3.22-3. The total number of fatal cancers
per year of operation of all Knolls Atomic Power Laboratory sites is
estimated to be 3.1E-5. Table 3.22-3 also summarizes the fatal
cancers estimated to occur in the population within 80 kilometers of
each site.
-------
3.22-3
Table 3.22-1. Atmospheric emissions of radionuclides from
Knolls Atomic Power Laboratory, 1977 (DOE77b)
Facility
Knolls Site
Kesselring Site
Windsor Site
Radionucl ide
Argon-41
Iodine-131
Krypton-85
Krypton-85m
Krypton-87
Krypton-88
Krypton-89
Mixed Fission Products
Plutonium- 239
Antimony-125
Thorium-232
Uranium-235
Xenon-131m
Xenon-133
Xenon-135
Xenon-138
Argon-41
Carbon-14
Cobalt-60
Krypton-83m
Krypton-85
Krypton-85m
Krypton-87
Krypton-88
Xenon-131m
Xenon-133
Xenon-133m
Xenon-135
Argon-41
Carbon-14
Cobalt-60
Krypton-83m
Krypton-85
Krypton-85m
Krypton-87
Krypton-88
Xenon-131m
Xenon-133
Xenon-133m
Xenon-135
Emissions
(Ci/y)
4.1
1.7E-5
4.3E-1
3.2E-3
6.6E-3
1.1E-2
6.2E-4
1.2E-4
1.3E-8
3.6E-4
1.4E-6
2.9E-6
2.3E-4
1.2E-3
9.7E-3
2.8E-3
8.5
l.OE-1
l.OE-5
4.2E-3
1.3E-5
1.2E-2
1.3E-2
2.5E-2
5.7E-4
1.1E-1
3.6E-3
l.OE-1
6.8E-1
6.3E-3
3.8E-6
9.4E-4
2.9E-6
2.8E-3
3.0E-3
5.9E-3
1.3E-4
2.4E-2
8.4E-4
2.5E-2
-------
3.22-4
Table 3.22-2. Annual radiation dosesa from atmospheric emissions
of radionuclides from Knolls Atomic Power Laboratory,
1977 (GE78)
Maximum individual
Site Site boundary Nearest population Population
(mrem/y) (mrem/y) (person-rem/y)
Knolls Site l.OE-2 2.0E-4
Kesselring Site 4.0E-2 4.0E-4
Windsor Site l.OE-3 2.0E-5
3.7E-2
1.1E-1
8.0E-3
aTotal body dose from airborne and liquid emissions.
Table 3.22-3. Individual lifetime risks and number of fatal
cancers due to radioactive atmospheric emissions
from Knolls Atomic Power Laboratory, 1977
Individual lifetime risks Expected fatal cancers
Site Maximum Average individual per year of operation
individual Regions (Fatal cancers)
Knolls Site 1.4E-7 4.2E-10 7.4E-6
Kesselring Site 5.6E-7 1.4E-09 2.2E-5
Windsor Site 1.4E-8 3.6E-11 1.6E-6
aThe region extends to 80 kilometers.
the population within 80 kilometers.
-------
3.22-5
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List, EIS 02,
(Computer Listing).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, EIS 51, (Computer Listing).
GE78 General Electric Company, 1978, Knolls Atomic Power Laboratory
Annual Environmental Monitoring Report Calendar Year, 1977,
KAPL-M-7537, Schenectady, N.Y.
-------
3.23-1
3.23 Shippingport Atomic Power Station
3.23.1 General Description
The Shippingport Atomic Power Station, operated by the
Duquesne Light Company for the Department of Energy, was the first
large scale central station nuclear reactor in the United States.
Initial power generation was achieved in December 1957. In 1977 the
station was shut down for the installation of the light water
breeder reactor (LWBR) core; initial criticality was achieved in
August 1977 and full power in September 1977.
The Shippingport Atomic Power Station is located on the same
site as the Beaver Valley Power Station, also operated by the
Duquesne Light Company. The site is a 197-hectare tract of land
located along the Ohio River in the Borough of Shippingport, Beaver
County, Pennsylvania. The site is approximately 40 kilometers
northwest of Pittsburgh. Beaver County, Pennsylvania, is considered
an integral part of the greater Pittsburgh industrial complex.
There are approximately 3.8 million people living within 80
kilometers of the site.
3.23.1 Process Description
The nuclear reactor at Shippingport Atomic Power Station is a
pressurized water reactor (PWR); however, it has the LWBR core which
operates on the basis of the thorium fuel cycle. Typical
characteristics of the plant and LWBR core are summarized in table
3.23-1. The reactor fuel is in the form of ceramic fuel pellets
with uranium-233 as the fissile material and thorium-232 the fertile
material. The major difference in operation of the LWBR core from
previous PWR cores, other than the type of fuel, is that the
reactivity behavior of the LWBR core is controlled by movable seed
or fissile fuel elements rather than by traditional control rods.
The potential source of radioactive airborne emissions is the
reactor coolant system which contains activated corrosion and wear
products, activated impurities, and small quantities of fission
products. The radioactivity can be released and become airborne
from coolant leaks, sampling operations, and maintenance and
overhaul operations.
-------
3.23-2
Gaseous wastes stripped from the reactor coolant are
circulated through a hydrogen analyzer and catalytic hydrogen burner
system where the hydrogen is removed. The gases are initially
stored in a vent gas surge drum, sampled, and subsequently
compressed and transfered to one of four gas storage drums. After a
long decay period, the decayed gases are sampled again prior to
release. In addition, the exhaust from the containment is equipped
with high efficiency filters to prevent release of radioactive
particulates. Protective devices are utilized in the event of high
airborne activity to automatically seal off the primary containment
to prevent an inadvertent release of radioactivity. Reactor plant
exhausts from the Decontamination Room, Sample Preparation Room,
Laundry Room, Radiochemistry Laboratory, Gaseous Waste System, and
Compacting Station are also equipped with high efficiency filters
and are continuously monitored for radioactive particulate by the
use of fixed filter monitors.
2.23.3 Emissions of Radionuclides
Table 3.23-2 summarizes the emissions from Shippingport
Atomic Power station in 1977. Since the plant was not in operation
the entire year, table 3.23-3 is provided to summarize the estimated
emissions to the atmosphere for the year.
-------
3.23-3
Table 3.23-1. Major core and plant parameters for the LWBR core at
Shippingport Atomic Power Station (ERDA76)
Parameter Value
POWER PLANT
Gross electrical output, MW(e) a62
Net station output, MW(e) £59
Net station heat rate, Btu/kW-hr 13,450
Steam pressure
Full load at generator, psia 744
No load at generator, psia 895
Number of loops 4
Reactor pressure drop, psi 69.2
Coolant piping, 00, in. 18
Coolant piping, ID, in. 15
Coolant velocity, main piping, ft/sec 35
REACTOR CORE
Type Pressurized light
water cooled and
moderated seed and
blanket
Total reactor heat output, MW(t) a204
Total coolant flow rate, 106 lb/hr 30.6
Reactor coolant inlet temperature
at 236.6 MW(t), OF 520
Reactor coolant outlet temperature
at 236.6 MW(t), °F 542
Average coolant temperature, nominal , op 531
Primary system pressure, nominal, psia 2000
Nominal core height, including
Th02 reflector, ft 10.0
Mean core diameter, ft 7.5
Fuel loading (thorium and uranium), metric tons ^42
Lifetime, EFPH 15,000a
aSee footnotes at end of table.
-------
3.23-4
Table 3.23-1. Major core and plant parameters for the LWBR core at
Shippingport Atomic Power Stationcontinued
Parameter Value
Fuel material
Movable Seed 233U02-Th02;
with Th02
end reflectors
Stationary Blanket 233LI02-Th02;
with Th02 end
reflectors
Reflector Blanket Th02
Fuel cladding material
Seed, blanket, and reflector Zircaloy-4, low
Hafnium
aThese are the minimum expected performance values for the LWBR
operation at Shippingport. To assure that environmental impacts
have been conservatively evaluated, the following parameters have
been analyzed in the environmental statement:
Gross electrical output, MW(e) 72
Net Station output, MW(e) 60
Total reactor heat output, MW(t) 236.6
Lifetime, EFPH 18,000
Table 3.23-2. Emissions of radionuclides to the atmosphere,
Shippingport Atomic Power Station, 1977 (DOE77b)
Emissions
Radionuclide (Ci/y)
Cobalt-60 2.5E-6
Manganese-54 1.6E-8
Xenon-133 1.8E-4
-------
3.23-5
Table 3.23-3. Estimated annual emissions of radionuclides
from LWBR operations (ERDA76)
Emissions
Radionuclide (Ci/y)
Argon-41 2.4
Krypton-83m 1.1E-2
Krypton-85m 1.8E-2
Krypton-85 . 2.5E-6
Krypton-87 3.3E-2
Krypton-88 5.7E-2
Iodine-130 7.0E-7
Iodine-131 6.4E-5
Iodine-132 4.6E-3
Iodine-133 l.OE-3
Iodine-134 1.1E-2
Iodine-135 2.4E-3
Xenon-131m 3.0E-8
Xenon-133m 6.1E-4
Xenon-133 6.0E-2
Xenon-135 1.9E-1
Xenon-135m 4.9E-2
Xenon-137 1.8E-1
Xenon-138 1.5E-2
3.23.4 Health Impact Assessment ofShippingport Atomic Power Station
The maximum exposure to a hypothetical individual residing at
the site boundary would be less than 0.5 mrem from airborne
radioactive emissions in 1977. These same emissions result in a
population dose of less than 1.0 person-rem to the population living
within 80 kilometers. Table 3.23-4 and table 3.23-5 summarize the
doses and risks from radioactive airborne emissions from the
Shippingport Atomic Power Station in 1977.
Since Shippingport Atomic Power Station was not in full
operation for the entire year of 1977 and because it shares the same
site as the Beaver Valley nuclear power plants, table 3.23-6 and
table 3.23-7 are provided for comparison.
-------
3.23-6
Table 3.23-4. Annual radiation doses from radioactive airborne
emissions from Shippingport Atomic Power Station, 1977 (DLC78)
Maximum individual dose Population dose
at the site boundary within 80 km
Organ (mrem/y) (Person-mrem/y)
Total Body 0.5 0.1
Table 3.23-5. Individual lifetime risks and number of fatal cancers
from radioactive emissions, Shippingport Atomic Power Station, 1977
Individual lifetime risks Expected fatal cancers
Source Site Average individual per year of operation
boundary Region (Fatal Cancers)
Shippingport 7.0E-6 3.7E-9 2.0E-4
Atomic Power Station
Table 3.23-6. Estimated annual radiation doses to the maximum
individual at the site boundary from radioactive airborne
emissions, Shippingport Atomic Power Station (ERDA76)
Organ
Bone
G.I. Tract
Thyroid
Skin
Total Body
From
Shippingport
only
(mrem/y)
1.8E-2
1.8E-2
3.7E-2
2.7E-2
1.8E-2
Combined effect
Shippingport and
Valley Units 1 &
(mrem/y)
4.7E-1
4.7E-1
7.7
8.3E-1
4.7E-1
from
Beaver
2
-------
3.23-7
Table 3.23-7. Estimated population doses from Shippingport and the
combined effects of Shippingport and Beaver Valley
Units 1 and 2 (ERDA76)
Facility
Shippingport
Shippingport and
Beaver Valley 1 & 2
Average
individual
(mrem/y)
8.0E-4
l.OE-3
Populationa
(person-rem/y)
3.2
4.0
aTo the population within 80 km.
-------
3.23-8
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List, EIS 02,
(Computer Listing).
DOE77b Department of Energy, 1977, Effluent Information System
Report NO. 51, Release Point Analysis report for Calendar Year
1977, EIS 51, (Computer Listing)
DLC78 Duquesne Light Company, 1978, 1977 Environmental Report,
Radiological - Volume #2, Duquesne Light Company, Beaver Valley
Power Station and Shippingport Atomic Power Station.
ERDA76 Energy Research and Development Agency, 1976, Final
Environmental Statement, Light Water Breeder Reactor Program,
ERDA-1541, Washington, D.C.
-------
3.24-1
3.24 Reactive Metals, Inc., Company (RMI)
3.24.1 General Description
The Reactive Metals, Inc., Company (RMI) operates a uranium
extrusion plant for the formation of rod or tubing from uranium
ingots for use in reactor fuel elements.
The RMI Company plant is located in Ashtabula, Ohio, in the
northeastern corner of the State.
3.24.2 Process Description
There are four stacks that emit airborne radioactivity from
the extrusion processes. The operations that are associated with
these release points are: Stack 1 exhausts from the extrusion press
tunnel where ingots are converted to rod or tubing; Stack 2 exhausts
from the extrusion exit roundout table for the press; Stack 4 emits
air from the abrasive saw where the extrusions are sectioned; and
Stack 5 exhausts where the pyrophoric scraps are converted to
oxide. Stack 5 is the only stack equipped with a waste treatment
system. Exhausted gases are passed through a Roto-Clone Type N Air
Scrubber before release to the atmosphere. All stacks are sampled
on a regular basis.
3.24.3 Emissions of Radionuclides
Table 3.24.1 summarizes atmospheric emissions of radio-
nuclides from the RMI Company in 1977. Natural uranium is the only
radionuclide reported.
3.24.4 Health Impact Assessment of the RMI Company
There were no reported doses for this facility.
-------
3.24-2
Table 3.24-1. Atmospheric emissions of radioriuclides,
RMI Company, 1977
Stack
number
1
2
4
5
Total
Radionuclide
Uranium- 238
Uranium-238
Uranium-238
Uranium-238
Emissions
(Ci/y)
6.7E-4
1.1E-4
2.0E-2
5.9E-4
2.11-2
-------
3.24-3
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List, EIS 02,
(Computer Listing).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis report for Calendar Year
1977, EIS 51, (Computer Listing).
-------
3.25-1
3.25 Lawrence Berkeley Laboratory
3.25.1 General Description
Lawrence Berkeley Laboratory (LBL) is a large multi-
disciplinary research institute. The laboratory carries out a wide
range of programs of research in the fields of physical and
biological sciences. The facilities include a number of large
accelerators, and various physics, chemistry, biology and medical
research laboratories.
The Lawrence Berkeley Laboratory (LBL) is part of the
University of California and is located on the western slope of the
hills parallel to the eastern side of San Francisco Bay. Populated
residential areas of the cities of Berkeley and Oakland enclose the
site to the north and south. The Berkeley Campus of the University
of California is on the west of LBL, and an uninhabited regional
park is to the east of the site. There are about 4.5 million people
residing within 80 kilometers of LBL.
3.25.2 Process Description
There are four large high energy particle accelerators at
LBL. The Bevatron, the SuperHilac and the 88-Inch Cyclotron are in
almost continuous operation while the 184-inch Synchrocyclotron is
used for short periods for biomedical studies.
The Bevatron is a large proton synchrotron, used for physics
research requiring energies up to 6.2 GeV. The SuperHilac is a
linear accelerator capable of accelerating natural elements,
including uranium, up to energies of about 8 MeV per nucleon or a
maximum of nearly 2 GeV per particle. The SuperHilac is used to
study transuranic elements and as an injector to the Bevatron.
When used as an injector to the Bevatron, a hybrid
accelerator is formed called the Bevalac. With this instrument
heavy ions may be accelerated to several GeV per nucleon and applied
to research in high energy physics, nuclear chemistry, radiobiology
and radiotherapy. The 88-inch sector-focused cyclotron accelerates
light and medium mass nuclei to energies intermediate between the
SuperHilac and the Bevalac. It is used for studies of nuclear
structure and radioisotope production.
The use of radionuclides in various research laboratories is
the principal potential source of leakage of radionuclides into the
-------
3.25-2
environment. There are over 100 such exhaust points, located on a
number of different buildings throughout the site. Most of these
consist of chemical laboratory room exhausts. Each laboratory room
has its own locally controlled exhaust. Handling of significant
quantities of radioactive materials is conducted in glove boxes
which are equipped with HEPA filters.
3.25.3 Emissions of Radioanuclides
The total quantities of radionuclides emitted into the
atmosphere are summarized in table 3.25-1. These emissions were
from chemical laboratory research operations and not from
accelerator operations.
Table 3.25-1. Atmospheric emissions of radionuclides
from Lawrence Berkeley Laboratory, 1977 (DOE775, LBL78)
Radionuclide Emissions
(Ci/y)
Tritium 7.8E+1
Carbon-14 2.5E-1
Gallium-67 1.3E-3
Iodine-125 4.6E-4
Unidentified Alpha <1.0E-6
Unidentified Beta & Gamma 4.1E-5
3.25.4 Health Impact Assessment of Lawrence Berkeley Laboratory
Maximum individual doses from radioactive airborne emissions
were not reported.
Airborne emissions resulted in a dose of 2.88 person-rem to
the population within 80 kilometers. Table 3.25-2 summarizes the
individual lifetime risks and number of fatal cancers to the
population from these doses.
-------
3.25-3
Table 3.25-2. Population dose, individual lifetime risks,
and number of fatal cancers due to radioactive atmospheric emissions
from Lawrence Berkeley Laboratory, 1977
Radio-
nuclides
Population
dose
(person-rem/y)
Individual
lifetime risks
Average individual
Region3
Unknown Beta
and Gamma .03
Unknown Alpha .3
Total 2.88
9.3E-11
9.3E-10
9.0E-09
Expected fatal cancers
per year of operation
(Fatal cancers)
Tr i t i urn
Carbon-14
Iodine-125
Gallium-67
2.4
.02
.09
.04
7.5E-09
6.2E-11
2.8E-10
1.2E-10
4.8E-4
4.0E-6
1.8E-5
8.0E-6
6.0E-6
6.0E-5
5.8E-4
aThe region extends to 80 kilometers.
°To the population within 80 kilometers.
-------
3.25-4
REFERENCES
DOE77a Department of Energy, 1977, Effluent Information System
Report No. 02, Narrative Summary Data Base Master List, EIS 02,
(Computer Listing).
DOE77b Department of Energy, 1977, Effluent Information System
Report No. 51, Release Point Analysis Report for Calendar Year
1977, EIS 51, (Computer Listing).
ERDA Energy Research and Development Administration, 1976,
Environmental Monitoring at Major U.S. Energy Research &
Development Administration Contractor sites, Calendar Year 1975,
ERDA-76-104, Washington, D.C.
LBL78 Lawrence Berkeley Laboratory, 1978, Annual Environmental
Monitoring Report of the Lawrence Berkeley Laboratory 1977, LBL
7570, UC,-41, Berkeley, California.
-------
3.26-1
3.26 Fermi National Accelerator Laboratory
3.26.1 General Description
Fermi National Accelerator Laboratory (Fermilab) is a proton
synchrotron that can operate at energies up to 500 GeV. Operations
at 400 GeV are now routine. The primary purpose of the installation
is fundamental research in high energy physics.
Fermilab is located in the greater Chicago area on a
2750-hectare tract of land east of Batavia, Illinois. The area
around the site is changing from farming to residential use with
many municipalities in the vicinity. About 8 million people live
within 80 kilometers of the site.
3.26.2 Process Description
The proton beam is extracted from the 2 km diameter main
accelerator and is taken to three different experimental areas.
Radioactivity is produced as a result of the interaction of the
accelerated protons with matter. Airborne radioactivity is produced
by radioactivation of air when the proton beam or the spray of
secondary particles resulting from its interactions with matter
passes through air. Most proton beam lines travel inside evacuated
pipes; thus, radioactivation of the air is usually caused by
secondary particles. No effluent treatment system is used, but
monitoring of such activation is carried out to control occupational
exposure.
3.26.3 Emissions of Radionuclides
Radioactive gas, primarily carbon-11, is produced by the
interaction of secondary particles with air. Small amounts of
radioactivity were released from stacks in the Neutrino Area during
1977. Tritiated helium was also released from the Meson Area Target
Box. Table 3.26-1 summarizes the emissions from Fermilab in 1977.
3.26.4 Health Impact Assessment of Fermilab
The maximum dose at the site boundary from radioactive
airborne emissions was 0.3 mrem. Airborne emissions resulted in an
off-site exposure of 0.5 person-rem to the population within 80
kilometers. Tables 3.26-2 and 3.26-3 summarize the doses and risks
resulting from Fermilab operations.
-------
3.26-2
Table 3.26-1. Atmospheric emissions of radionuclides from
Fermi National Accelerator Laboratory, 1977
Source Radionuclide Emissions
(Ci/y)
Neutrino Area Carbon-11 l.OE+4
Meson Area
Target Box Tritium 5.5E-1
Table 3.26-2 Annual radiation doses from radioactive airborne
emissions from Fermi National Accelerator Laboratory, 1977
Maximum individual Population
Organ Site boundary within 80 km
(mrem/y) (person-rem/y)
Total body 0.3 0.5
Table 3.26-3. Individual lifetime risks and number
of fatal cancers due to radioactive emissions
from Fermi National Accelerator Laboratory
Expected fatal
Individual lifetime risks cancers per year
Source Maximum individual Average individual of operation
Site boundary Regiona (Fatal Cancers)
Fermilab 4.2E-6 8.8E-10 l.OE-4
aTo the population within 80 km.
-------
3.26-3
REFERENCES
Ba78 Baker, Samuel I., 1978, Environmental Monitoring Report for
Calendar Year 1977, Fermilab-78/27, 1104.100, Fermi National
Accelerator Laboratorey, Batavia, Illinois, .
-------
3.27-1
3.27 Stanford Linear Accelerator Center
3.27.1 General Description
Stanford Linear Accelerator Center (SLAC) is a Targe research
laboratory devoted to theoretical and experimental research in high
energy physics and to the development of new techniques in high
accelerator particle detectors.
SLAC is located about 3 kilometers west of the Stanford
campus in San Mateo County, California. The total length of the
accelerator and the experimental area is approximately 4.8
kilometers, oriented almost east-west. The accelerator center
occupies about 170 hectares of Stanford University land in the
foothills of the Santa Cruz Mountains on the San Francisco
peninsula. The site is halfway between San Francisco and San Jose.
There are about 4.2 million people living in the six county area of
the San Francisco Bay Area.
3.27.2 Process Description
The linear accelerator at Stanford is 2 miles long. It is
capable of accelerating beams of electrons with energies up to 22
billion electron volts (GeV), and positrons up to 15 GeV. The east
end of the 2-mile accelerator contains the research area. Included
in this area are the beam switchyard, end stations A and B, counting
house, data, assembly, and cyrogenics buildings and several utility
buildings.
The accelerator and beam switchyard are vented at a location
slightly above roof level after the electron beam is shut off. The
area is vented for 10 minuits before entry.
3.27.3 Emissions of Radionuclides
Release of airborne radioactivity is infrequent and only for
brief periods of time, usually 30-60 minutes. During 1977 only 1.7
curies of short-lived gaseous radioactivity were released to the
atmosphere from SLAC. The isotopes emitted have half-lives ranging
from 2 minutes to 1.8 hours. The isotopes are oxygen-15,
nitrogen-13, carbon-11, and argon-41.
3.27.4 Health Impact Assessment of Stanford Linear Accelerator
Facility
The maximum individual dose from airborne radioactive
emissions from SLAC was less than 0.03 mrem per year. This may be
-------
3.27-2
compared to a maximum annual dose at the site boundary of 8.2 mrems
from penetrating neutron radiation. No population dose was reported
from airborne emissions from Stanford Linear Accelerator
operations. Tables 3.27-1 and 3.27-2 summarize the dose and
lifetime risks associated with airborne emissions from SLAC.
Table 3.27-1. Annual radiation doses from radioactive airborne
emissions from Stanford Linear Accelerator Center, 1977
Maximum individual
Source Site boundary Population
(mrem/y) (person-rem/y)
SLAC <0.03 NR
~NRNot reported.
Table 3.27-2. Individual lifetime risks and number of
fatal cancers due to radioactive emissions from
Stanford Linear Accelerator Center
Individual lifetime risks Expected fatal cancers
Source Maximum individual per year of operation
Site boundary (Fatal Cancers)
SLAC <4.2E-7 NR
-------
3.27-3
REFERENCES
ERDA76 Energy Research and Develoment Administration, 1976, Final
Environmental Statement, Positron-Electron Storage Ring Project,
Stanford Linear Accelerator Center, Stanford, California, ERDA-
1546, Washington, D. C.
SLAC78 Stanford Linear Accelerator Center, 1978, Annual Environ-
mental Monitoring Report, January-December 1977, Stanford,
California.
-------
CHAPTER 4
SOURCES OF EMISSIONS OF NATURALLY OCCURRING RADIONUCLIDES
This chapter deals with sources which emit naturally
occurring radioactive materials into the atmosphere. The source
categories include: uranium mining, uranium milling, phosphate
mining and processing, coal-fired steam electric generation, metal
and non-metal mining (other than uranium and phosphate), and
underground water sources. These are the source categories, which
at the present time, have been identified as having the greatest
potential for release of naturally occurring radionuclides to the
air. Future reports will include the assessments of other source
categories now being evaluated. The uranium mining and milling
source categories could have been included in Chapter 2 under the
uranium fuel cycle; however, they are presented in this chapter
because the types of activities and radionuclides involved with
these categories are similar to those treated in the other sections
of this chapter.
Naturally occurring radionuclides fall into two general
categories: primordial and cosmogenic. The more important of
these, in terms of air emissions, are the primordial radionuclides
and their daughter products. These would include the uranium-238,
uranium-235, and thorium-232 decay series. The soils and rocks
which make up the earth's crust contain these radionuclides and
their daughter products in widely varying amounts. Average values
for uranium-238 and thorium-232 in soils have been reported to be
about 1.8 ppm (0.6 pCi/g) and 9 ppm (1 pCi/g), respectively (NCRP75),
Because of the presence of these radionuclides in the earth's
crust, almost all of man's activities which involve the removal and
processing of materials from the earth's surface, or the removal of
gases, vapors, or liquids from below the earth's surface can result
in the release of some of these radioactive materials to the
atmosphere. These releases can become potentially important when
(1) the activity involves the handling of materials containing
concentrations of these radionuclides (specific activities)
significantly above the average concentrations in soil, (2) these
radionuclides are concentrated during processing to a level
significantly above the average concentrations in soil, or (3) the
radioactive material is redistributed from its place in nature into
a pathway where man can be exposed. Each of the source categories
covered in this chapter involves one or more of the above
considerations.
-------
4.0-2
REFERENCES
NCRP75 National Council on Radiation Protection and Measurements,
1975, Natural Background Radiation in the United States, NCRP Re-
port No. 45, Washington, D. C.
-------
4.1-1
4.1 Uranium Mining
4.1.1 General Description
Uranium mining operations involve the removal from
underground of large quantities of ore containing uranium and its
daughter products in concentrations up to 1000 times the
concentrations of these radionuclides in the natural terrestrial
environment. The concentration of uranium in currently mined ores
ranges from 0.1 to 0.2 percent U^QQ or 280 to 560 microcuries of
uranium-238 per metric ton of ore. Since the uranium-238 in these
ores is usually present in secular equilibrium with its daughter
products, these ores also contain an equal amount of each of the
members of the uranium decay series.
After mining, the ores are shipped to a uranium mill (see
section 4.2) for separation of uranium for subsequent use in light
water nuclear power reactors. Emissions from uranium mines consist
of airborne radioactive dusts and radon-222 gas.
Uranium mining is generally carried out either by open pit or
underground mining methods. When ore deposits are near the surface,
open pit methods are used. For deep ore deposits underground
methods are preferred. In 1977 there were 251 underground and 36
open pit uranium mines in operation in the United States (table
4.1-1). These mines accounted for about 96 percent of the uranium
produced with each mining method accounting for approximately the
same amount of uranium (DOE78). In recent years in-situ solution
mining has become more widely used and the amount of uranium mined
by this method is expected to increase in future years. However,
during 1977 this method accounted for only a few percent of the
uranium mined in the United States.
All of the present uranium mining takes place in Western
States. In general these mines are located in relatively remote low
population areas. Table 4.1-2 shows the production of uranium in
ore by State. These data show that 77 percent of the uranium
production takes place in the States of New Mexico and Wyoming.
Projections of future requirements of uranium by the nuclear
power industry indicate that an annual production of about 75,000
metric tons of U308 in ore will be needed in the year 2000
(NRC79). This is about a five-fold increase in uranium in ore
production over 1977 and indicates a substantial increase in uranium
mining in the next two decades.
-------
4.1-2
Table 4.1-1. Distribution of 1977 UqOg production in ore
by mining method (DOE/8)a
Source
Underground mines
Open pit mines
Others: heap leach,
mine water, solution
mining, low-grade
stockpiles
Total
Number
251
36
27
314
Tons 1)303
8,300
7,600
800
16,700
% of total
50
46
4
100
aShort tons
Table 4.1-2. Distribution of 1977 UoQs production in
ore by State (DOE78)a
State
New Mexico
Wyomi ng
Others: Arizona,
Colorado, Texas,
Utah, & Washing-
ton
Total
Tons of ore
4,209,000
3,834,000
2,967,000
11,012,000
Tons 0303
7,600
5,200
3,900
16,700
% of total
UaOa
46
31
23
100
aShort tons
-------
4.1-3
4.1.2 Process Description
Underground Mining
Underground uranium mining is usually carried out using a
modified room and pillar method. In this method, a large diameter
main entry shaft is drilled to a level below the ore body. A
haulage way is then established underneath the ore body. Vertical
raises are then driven up from the haulage way to the ore body.
Development drifts are driven along the base of the ore body
connecting with the vertical raises. Mined ore is hauled along the
development drift to the vertical raises and gravity fed to the
haulage way for transport to the main shaft for hoisting to the
surface.
Figure 4.1-1 is an example of an underground mining
operation. Ventilation shafts are installed at appropriate
distances along the ore body. Typical ventilation flow rates are
on the order of 6,000 m^/min. The principal radioactive effluent
in the mine ventilation air is radon-222 which is released during
mining operations.
Surface Mining
Open pit mining usually is carried out by excavating a
series of pits in sequence. The mining procedure followed is to
remove the topsoil and overburden from above the ore zone and to
stockpile these materials in separate piles for use in future
reclamation operations. The uranium ore is removed from the
exposed ore zone and stockpiled for transport to a uranium mill.
Ore stockpiles range in size up to several hundred thousand metric
tons of ore. During the mining of the uranium ore, low grade waste
rock is also removed from the pits and stored in a waste stockpile
for possible future use.
Figure 4.1-2 is an example of an open pit mining operation.
As the mining progresses, mining and reclamation operations take
place simultaneouslypits are mined in sequence and the mined-out
pits are reclaimed by backfilling with overburden and topsoil. In
some cases the last of the open pits in a mining operation are not
backfilled but are allowed to fill with water, forming a lake.
Radioactive emissions from open pit mining operations are
radioactive fugitive dust and radon-222 gas.
-------
4.1-4
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-------
4.1-5
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-------
4.1-6
4.1.3 Emissions of Radionuclides
Underground Mines
Only limited information on radioactive emissions from
underground uranium mines is currently available. In general
information is limited to radon-222 emissions because this
radionuclide represents the major airborne radioactive effluent from
underground uranium mines. The data originates primarily from
measurements made at seven Kerr-McGee mines in the Ambrosia Lake
area of New Mexico. Radon-222 emissions from these mines measured
in 1976 by Kerr-McGee and in 1978 by the Pacific Northwest
Laboratory (PNL79a) are summarized in table 4.1-3.
The Pacific Northwest Laboratories are also conducting
measurements of radon-222 emissions from a series of other
underground uranium mines. These data should be available in the
near future.
Particulate emissions from underground mines are believed to
be much less significant than the radon-222 emissions.
Open Pit Mines
Radon-222 is also believed to be the major airborne
radioactive effluent from open pit uranium mining. However, because
of the diffuse nature of these mining operations, no direct
measurement of radon-222 emissions can be made. Therefore, emission
rates of radon-222 from open pit mines must be estimated using
calculational methods. These methods involve calculating radon-222
emissions either (1) from measured or estimated fluxs of radon-222
exhaled from the various surfaces of the mine, or (2) from measured
radon-222 concentrations in surface air downwind of the mining
operation.
The only information currently available on radon-222
emissions from open pit uranium mines are estimates calculated by
Pacific Northwest Laboratories (PNL79b). An annual release of 2000
curies per year was calculated for a model open pit mine based on an
estimated flux of radon-222 exhaled from the mine surface. The
model mine parameters used in these calculations were obtained from
a survey of eight open pit uranium mines in Wyoming.
-------
4.1-7
Table 4.1-3. Radon-222 emissions from Kerr-McGee underground
uranium mines at Ambrosia Lake, New Mexico
Mine
section
17
19
24
30
30W
35
36
Mine
section
17
19
24
30
30W
35
36
Number
of vents
10
4
6
11
5
4
4
Number
of vents
10
4
6
11
5
4
4
Ventilation
rate (m3/min)
5,900
4,200
3,900
10,400
8,100
9,900
5,300
Ventilation
rate (m3/min)
5,600
4,400
4,800
11,400
8,100
11,000
6,400
a!976
Ore process
rate (t/d)
327
428
208
548
863
867
308
b!978
Ore process
rate (t/d)
328
420
208
548
865
870
202
Radon-222
Ci/d
13.1
4.3
12.2
12.1
4.3
3.7
5.1
Radon-222
Ci/d
22
4
40
23
12
12
16
released
Ci/t
0.040
0.010
0.059
0.022
0.005
0.004
0.017
released
Ci/t
0.067
0.010
0.19
0.043
0.014
0.014
0.080
aLetter from W.J. Shelley of Kerr-McGee to Ralph M. Wilde of the
Nuclear Regulatory Commission, December 9, 1976.
bSource: PNL79a.
-------
4.1-8
4.1.4 Model Facilities
In order to estimate radioactive emissions and health impacts
from uranium mining operations, model underground and open pit mines
were developed by assigning appropriate values for the various
parameters that are important in estimating emissions . These
parameters which are taken directly from work reported by Pacific
Northwest Laboratories (PNL79a and PNL79b) are listed in tables
4.1-4 and 4.1-5.
Mine Emissions
The atmospheric emissions of radioactive materials from the
model mines are listed in tables 4.1-6 and 4.1-7. These data were
obtained from PNL reports (PNL79a and PNL79b). The data presented
are limited to radon-222 emissions, the only data presently
available and the radionuclide representing the major airborne
effluent from uranium mines.
Table 4.1-4.' Model underground uranium mine
Parameter Valuea
Ore mining rate 500 t per day
Mine lifetime 20 years
Ore grade 0.2% U$QQ
Number of vents 5
Surface Ore Storageb 3500 t
Waste storage pilesb
Area 11 hectares
Height 1.2 meters
Grade 0.025% U308
aSource: PNL79a.
^Average values during mine lifetime.
-------
4.1-9
Table 4.1-5. Model open-pit uranium mine
Parameter
Valuea
Ore mining rate
Mine lifetime
Average ore grade
Average subore grade
Average overburden gradeb
Number of pitsc
Pit depth
Overburden depth
Ore zone depth
Overburden to ore ration
Ore to subore ratio
Total volume of pit
Total volume of ore
Area of ore storage pile
Ore stockpile residence time
Area of subore pile d
1600 t per day
17 years
0.11% U^QQ
0.025% 0303
0.002% U^OQ
7
77 meters
65 meters
12 meters
77:1
1:1
5.5xlo7_m3
7.1xl05_m3
1.1 ha
41 days
10 hectares
aSource: PNL79b.
"Overburden initially contains 4 ppm U^QQ but reaches
20 ppm through relocation and mixing.
cThe model mine consists of a series of 7 pits in various stages
of excavation, mining and reclamation. A detailed description of
the model mine, and the various sources contributing to radon-222
emissions is presented in PNL79b.
^Area of subore pile at midpoint of mine lifetime.
-------
4.1-10
Table 4.1-6. Atmospheric emissions of radionucTides
from the model underground uranium mine
aRadon-222
Source of Emission (Ci/y)
Mine vents& 6500
Ore storage 12
Waste storage piles 217
Total 6729
aSource: PNL79a.
b!300 Ci/y per vent.
Table 4.1-7. Atmospheric emissions of radionuclides
from the model open pit uranium mine
aRadon-222
Source of Emission (Ci/y)
Active open pit 894
New pit being excavated 148
Ore stockpile 103
Subore waste pile 163
Overburden waste pile 148
Refilled pits 391
Increased land area 15
Truck loading and dumping 99
Total b!961
aSource: PNL79b.
^Average annual radon-222 emission over the lifetime of mine.
-------
4.1-11
4.1.4 Health Impact Assessment of Model Uranium Mines
The estimated working level exposures that would result from
radon-222 emissions from the model uranium mines are listed in table
4.1-8. These are estimates for a low population density, generic
uranium mining and milling site in the Western United States (Site
E, Appendix A). The model underground mine consists of 5 vents each
releasing an equal amount of radon-222. The working level exposure
to the highest group of individuals was calculated for a location
500 meters from one of the vents in the predominent wind direction.
The model underground mine was treated as an area source of 1000
hectares and the working level exposure to the highest exposed group
of individuals was calculated at a location 500 meters from the edge
of the mining area in the predominant wind direction.
In addition to the working level exposures from the
inhalation of short-lived radon-222 daughter products listed in
table 4.1-8, radiation doses from radon-222 emissions also occur to
body organs from the inhalation and ingestion of lead-210 formed
from the decay of the released radon-222. Data on the doses to the
population of the United States from lead-210 from radon-222
released from uranium milling operations is presented in detail in
ORNL79. Our preliminary evaluation of these data indicated that the
health impact resulting from the organ doses from lead-210 is
smaller than the health impact from the inhalation of the
short-lived radon-222 daughters. The health risk data presented in
this report does not include these small additional incremental
risks from lead-210.
Estimates of the individual lifetime risks and number of
fatal cancers resulting from these working level exposures are given
in table 4.1-9. The risks from the model underground mine are
greater than the open pit mine because of the larger quantity of
radon-222 released.
For the model underground mine the lifetime risk to the
highest exposed group of individuals is estimated to be about 1 x
10-2. The lifetime fatal cancer risk to the average individual in
the regional and the United States populations is estimated to be
about 5 x lO-5 and 3 x 10'8 respectively. The number of fatal
cancers per year of mine operation is estimated to be 0.03 in the
regional population and 0.08 in the population of the United States.
-------
4.1-12
Table 4.1-8. Working level exposures from radon-222
emissions from model uranium mines
Source
Underground mine
Open pit mine
Maximum
individual
(ML)
b6.0E-3
C8.4E-4
Regional
population
(person-WL)
1.3
3.8E-1
United States
population^
(person-WL)
3.8
1.1
Calculated from data in ORNL79Table 2.1 where a one kCi
radon-222 release is estimated to result in an exposure of 8.0E+4
person-pCi/m3 to the population of the United States. This is
equivalent to an exposure to 0.56 person-working levels based on an
assumption of a 70 percent equilibrium of the radon-222 daughter
products (100 pCi/L radon-222 = 0.7 WL) which is considered to be
representative of indoor exposure conditions (Ge78).
^Exposure to an individual living 500 meters from a mine
vent in predominant wind direction
cExposure to an individual living 500 meters from edge of mining
area which has been treated as an area source of 1000 ha.
-------
4.1-13
Table 4.1-9 Individual lifetime risks and number of fatal
cancers in the population from radon-222 emissions
from model uranium mines.
Individual lifetime risks
Source Maximum Average individual
Individual Region United States
Underground mine
Open pit mine
9.0E-3
1.3E-3
5.3E-5
1.6E-5
2.6E-8
7.4E-9
Expected fatal cancers per year of operation
Source Region United States Total
(Fatal cancers) (Fatal cancers) (Fatal cancers)
Underground mine 2.7E-2 8.0E-2 1.1E-1
Open pit mine 8.0E-3 2.3E-2 3.1E-2
-------
4.1-14
REFERENCES
DOE78 Department of Energy, 1978, Statistical Data of the Uranium
Industry, GJO-100(78), Washington, D.C.
6e78 George A.C., and Breslin, A.J., 1978, The Distribution of Ambi
ent Radon and Radon Daughters in Residential Buildings in the New
JerseyNew York Area, Presented at Symposium on the Natural Radia-
tion Environment III, Houston, Texas.
NRC76 Nuclear Regulatory Commission, 1976, Final Generic
Environmental Statement on the Use of Recycle Plutonium in Mixed
Oxide Fuel in Light Water Cooled Reactors (GESMO), NUREG-002,
Vol 3, Washington, D.C.
NRC79 Nuclear Regulatory Commission, 1979, Draft Generic
Environmental Impact Statement on Uranium Milling, NUREG-0511
Washington, D.C
PNL79a Jackson, P., et al., 1979, Radon-222 Emissions in Ventilation
Air Exhausted from Underground Uranium Mines, PNL-2888,
NUREG-CR-0627, Richland, Washington.
PNL79b Neilson K., 1979, Prediction of Net Radon Emission from a
Model Open Pit Uranium Mine, PNL-2889, NUREG-CR-0628,
Richland, Washington.
ORNL79 Travis C.C., et al., 1979, A Radiological Assessment of
Radon-222 Released from Uranium Mills and other Natural and
Technologically Enhanced Sources, (NUREG/CR-0573),
ORNL/NUREG-55 Oak Ridge, Tennessee.
-------
4.2-1
4.2 Uranium Mills
4.2.1 General Description
Uranium milling operations involve the handling and
processing of large quantities of ore containing uranium and its
daughter products in concentrations up to 1000 times the
concentrations of these radionuclides in the natural terrestrial
environment.
The concentration of uranium in the ores currently being
processed ranges from about 0.1 to 0.2 percent 1)303 or 280 to
560 microcuries of uranium-238 per metric ton of ore. Since the
uranium in these ores is usually present in secular equilibrium
with its daughter products, these ores also contain an equal amount
of each of the members of the uranium-238 decay series.
The function of a uranium mill is to extract uranium in
concentrated form from naturally occurring ore deposits. The
product is a semirefined uranium compound called yellowcake which
is shipped to a conversion plant for a further purification step in
preparing the uranium for use in light-water nuclear power
reactors. Liquid and solid wastes are impounded near the mill in a
tailings pond or pile. Emissions consist of airborne radioactive
dusts of ore, tailings, and yellowcake and radon-222 gas.
As of January 1979 there were 20 active (conventional)
uranium mills in the United States (GJ079) located in low
population density areas in Western States. Table 4.2-1 lists
these mills, their locations, and normal processing capacities.
In 1977 uranium mills processed about 10 million tons of
ore with an average ore grade of 0.15 percent U30g containing
about 4000 curies of each of the members of the uranium-238 decay
series. The operation of these uranium mills has resulted in the
accumulation of large quantities of waste tailings.
There are currently over 100 million tons of tailings
stored at these active uranium mill sites containing in excess of
50,000 curies each of thorium-230, and radium-226 and its decay
products. These tailings represent the major source of radon-222
emissions at a uranium mill site.
Projections of future uranium requirements (NRC79a) indicate
that by the year 2000 about 75,000 metric tons of U30o will be
needed to support the nuclear power industry. To produce this
quantity of uranium would require an additional 43 uranium mills
with 1800 t/day capacities. Based on this projection, a large
increase in the uranium milling industry can be expected over the
next two decades.
-------
4.2-2
Table 4.2-1.
Location and capacity of uranium mills in the
'United States (GJ079)
State & Company
Location
Nominal capacity
ore/day
(metric tons)
New Mexico
Anaconda Company
Kerr-McGee Nuclear Corporation
Sohio Natural Resources Co.
United Nuclear Corporation
United Nuclear-Homestake Partners
Total
Wyoming
Exxon, U.S.A.
Federal-American Partners
Pathfinder Mines Corporation
Pathfinder Mines Corporation
Petrotomics
Bear Creek Uranium Company
Union Carbide Corporation
Western Nuclear, Inc.
Total
Utah
Atlas Corporation
Rio Algom Corporation
Total
Colorado
Cotter Corporation
Union Carbide Corporation
Total
Texas
Conoco & Pioneer Nuclear, Inc.
Washington
Dawn Mining Company
Western Nuclear
Total
Grants
Grants
Cebolleta
Church Rock
Grants
Powder River Basin
Gas Hills
Gas Hills
Shirley Basin
Shirley Basin
Powder River Basin
Natrona County
Jeffrey City
Moab
La Sal
Canon City
Uravan
Falls City
Ford
Wellpinit
5,450
6,350
1,450
2,720
3.090
2,720
860
1,500
1,630
1,360
1,270
1,090
1,540
11,970
1,270
680
410
1,180
2,910
410
1.820
Grand total
39,710
-------
4.2-3
4.2.2 Process Description
Ore Storage
When ore is delivered to the mill site, it is stored on an ore
storage pad with the amount stored ranging from a 10-day to a 6-month
supply. Therefore, at some mill sites as much as several hundred
thousand tons of ore may be stored prior to milling. Ore as delivered
to the storage pads generally contains a relatively high moisture
content; however, significant drying out can take place during
storage. Radioactive materials are released from these storage piles
through diffusion of radon-222 gas, dusting from wind erosion, and
heavy equipment operation. The ore is transferred from the storage
pads to the mill crushing unit using front-end loaders or bulldozers.
Milling
The process of uranium extraction involves the following
steps: crushing, grinding, chemical leaching, separation of the
uranium from the leach solution, precipitation, drying and packaging
of yellowcake. Mill processes fall into three general types:
acid-leach solvent extraction, acid-leach ion-exchange, and alkaline
leach. Most mills today utilize an acid-leach solvent extraction
process (figure 4.2-1).
The steps in the milling process which generate the major
radioactive emissions are the front-end crushing operations and the
drying and packaging of yellowcake. The other operations in the
milling process are carried out in a wet state and therefore do not
result in the generation of any significant airborne dust emissions.
Tailings Impoundment
Uranium mill wastes are usually stored in a tailings
impoundment located on the mill site. The tailings pile is usually
located in a gently sloping natural drainage area and covers an area
of about 30 to 60 hectares. Tailings are discharged to the
impoundment area in slurry form, at about 50 percent solids. The
liquids are partially recycled to the mill or undergo natural
evaporation. In the past a starter dam was constructed of native soil
materials and the remainder of the dam was built from sand tailings.
However, present practice is to construct the dam entirely from
imported materials using an impermeable clay core. Tailings dams may
be as high as 30 meters. The tailings are comprised of two fractions,
sands and slimes. The sand fraction (>200 mesh) makes up about 70
per cent of the tailings and the slime fraction (<200 mesh), the
remaining 30 percent. However, it has been estimated (ORNL75) that
the slime fraction contains about 85 percent of the radionuclide
content of the tailings.
-------
4.2-4
RAFFINATE
FROM
ORE
CRUSHING
WATER
WET GRINDING
SULFURIC
ACID
1 SODIUM
CHLORATE
It
LEACHING
COUNTERCURRENT
DECANTATION
(CCD)
AMINE.
KEROSENE.
ALCOHOL
TAILINGS- SAND,
SLIME, LIQUID
WASTES TO
TAILINGS POND
SOLVENT
EXTRACTION
RAFFINATE
RECYCLED
TO LEACHING
STRIPPING
AMMONIA
1
PRECIPITATION
FILTRATION
DRYING
YELLOW CAKE
PACKAGING
PRODUCT
Figure 4.2-1. Flow diagram for the acid-leach process (NRC79a)
-------
4.2-5
The tailings impoundment is made up of a pond and dry beach
area with the size of each component dependent on the amount of
water recycled and the rate of evaporation. In areas of high
evaporation, large, dry beach areas are exposed. Radioactive
airborne emissions from these dry beach areas take place as a result
of wind erosion of the tailings and the diffusion of radon-222 gas.
Essentially no airborne emissions originate from pond areas.
Control Technology
Methods used for controlling airborne dust emissions at
uranium mills consist of wet scrubbers, impingement scrubbers, or
bag filters. No attempt is made to control radon-222 emissions from
these mills. Dust control of crushing and fine-bin storage areas is
accomplished by passing air through a wet scrubber prior to
exhausting it through a roof vent. However, a number of mills do
not use exhaust ventilation in these areasdusting is not a problem
because of the high moisture content of the ore processed or the use
of a semi-autogenous grinding operation. All mills utilize some
method of dust control in yellowcake areas. Air cleaning systems on
the yellowcake exhausts are largely wet scrubbers, usually orifice
or impingement scrubbers, with Venturi scrubbers used occasionally.
Bag filters are used sometimes on the packaging exhaust stack, but
are not suitable for use on the dryer stack because of the
temperature and moisture content of the dryer exhaust air.
4.2.3 Emissions of Radionuclides
Table 4.2-2 summarizes estimates of airborne radioactive
emissions reported in the most recent environmental impact
statements for uranium mills. These emission estimates vary
considerably and cover a very wide range of values because of
differences in the milling operations and uncertainties about actual
emissions, which is due to a lack of actual measurement data on
emission rates. Only a few such measurements are currently
available and these represent relatively short time periods.
Emission rates for the milling operations (table 4.2-2) fall
in these approximate ranges: 30-120 mCi/y for uranium-238; 30-120
mCi/y for uranium-234; 5-50 mCi/y for thorium-230, radium-226 and
lead-210; and 25-170 Ci/y for radon-222. The range of emission
rates from the ore storage and crushing activities is much wider
than the range for the yellowcake drying and packaging activities.
However, the more recent estimates for crushing and storage
activities (USFS78, NRC79a, NRC79b) show a much smaller rangeabout
1-4 mCi/y for each of the radionuclides emitted in particulate
form. The data indicate that the major source of atmospheric dust
emissions from a uranium mill originate from the yellowcake drying
and packaging operations.
-------
4.2-6
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4.2-7
Emission rates from the tailings disposal area (table 4.2-2)
show the following approximate ranges: 0.2-14 mCi/y of uranium-238;
0.2-14 mCi/y of uranium-234; 3-200 mCi/y of thorium-230, radium-226,
lead-210; and 14-8500 Ci/y of radon-222. The amount of airborne
emissions from tailings disposal areas depends upon the size of dry
tailings beach areas which are subject to wind erosion and radon-222
diffusion. Tailings impoundment areas almost completely covered by
water will have low radionuclide emissions (NRC79b).
As stated, actual measurement data on emissions rates are
relatively scarse. However, a number of special studies are now in
progress to develop information on radioactive emissions rates from
uranium mills. These studies are being carried out by EPA-Las Vegas
and by Argonne National Laboratory and Battelle Northwest Laboratory
under contract to the Nuclear Regulatory Commission (NRC). In
addition, newly developed monitoring requirements for uranium mills
by the NRC (NRC77d) will result in the generation of emission data
by the industry in the near future.
4.2.4 Model Facility
In order to estimate radioactive emissions and health impacts
from uranium milling operations, a model uranium mill and tailings
pile were developed by assigning values for the various parameters
that are important in assessing impacts (table 4.2-3). The mill
parameters were taken primarily from an EPA report (EPA73) and the
parameters for the tailings pile are those used by Magno (Ma78).
Mill Emissions
The atmospheric emissions of radioactive material from the
model uranium mill are listed in table 4.2-4. Emissions are
presented separately for the milling operations and the tailings
disposal area. The emission rates for the milling operations
(except for radon-222) come from an EPA report (EPA76) which was
based on data reported by Oak Ridge National Laboratory (ORNL75).
The emission rates for the tailings disposal area were derived from
information reported by Magno (Ma78) and Oak Ridge (ORNL75). The
footnotes to table 4.2-4 indicate the source and basis for these
emission estimates.
-------
4.2-8
Table 4.2-3. Model uranium mill
Parameter
Value
Type of process
Ore process rate
Operating days per year
Mill lifetime
Ore grade
Uranium recovery
Ore activity
Ore storage area
Ore storage time
Effective stack height
Area of tailings impoundment
Dry beach
Pond and wet beach
Average depth of tailings
Acid-leach solvent extraction
2000 metric tons per day
300 days
20 years
0.2% U308
95%
560 pCi/g, Uranium-238 and
daughter products in secular
equilibrium
1 hectare
10 days
15 meters
60 hectares
15 hectares
45 hectares
12 meters
-------
4.2-9
Table 4.2-4.
Radionuclide
Atmospheric emissions of radioactive materials
from the model uranium mill
Emissions
(Ci/y)
Milling
operations9
Tailings disposal area^
(0-10 ym) (10-80 ym)
Uranium-238
Uranium-234
Thorium-230
Radium-226
Lead-210
Polonium-210
Radon-222
9.0E-2
9.0E-2
l.OE-2
5.0E-3
5.0E-3
5.0E-3
c!.2E+2
6.0E-4
6.0E-4
1.2E-2
1.2E-2
1.2E-2
1.2E-2
d2.7E+3
1.5E-3
1.5E-3
3.0E-2
3.0E-2
3.0E-2
3.0E-2
Reference: EPA76table 5.0-1.
"Estimated from data in ORNL75 as follows: Dust emissions
for 4.5 m/s average wind speed=1.7E-2 g/ha-s for 0-10 ym
particles and 4.0E-2 g/ha-s for 10-80 ym particles (table 7.4).
Radionuclide concentrations of dust particles were 1610 pCi/g
for 230Th, 226Ra, 210Po, 210Pb and 80 pCi/g for 234U
and 238u (table 4.12).
GBased on values listed in table 4.2-1 of this report.
^Estimated from data in Ma78 as follows: Radon-222 release
from dry tailings area = 180 Ci/ha (based on 560 pCi/g 22°Ra).
-------
4.2-10
4.2.5 Health Impact Assessment of Model Uranium Mill
Tables 4.2-5 and 4.2-6 estimate annual radiation doses and
working level exposures resulting from radioactive emissions from
the model uranium mill. The estimates are for a low population
density, generic uranium mining and milling site in the Western
United States (Site E, Appendix A).
In addition to the working level exposures from the
inhalation of short-lived radon-222 daughter products listed in
table 4.2-6, radiation doses from radon-222 emissions also occur to
body organs from the inhalation and ingestion of lead-210 formed
from the decay of the released radon-222. Data on the doses to the
population of the United States from lead-210 from radon-222
released from uranium milling operations is presented in detail in
ORNL79. Our preliminary evaluation of these data indicated that
the health impact resulting from the organ doses from lead-210 is
smaller than the health impact from the inhalation of the
short-lived radon-222 daughters. The health risk data presented in
this report does not include these small additional incremental
risks from lead-210.
Table 4.2-7 estimates the individual lifetime risks and
number of fatal cancers to the population resulting from these
doses and working level exposures. The lifetime cancer risk to the
highest exposed group of individuals is estimated to be about
1x10-2 an(j resL|its primarily from emissions from the tailings
disposal area. The lifetime cancer risks to the average individual
in the regional and United States populations are estimated to be
about 10-5 and 10~8, respectively.
The number of cancers per year of plant operation is
estimated to,be 0.01 to the population living in the region around
the plant and 0.03 to the population of the United States. These
result primarily from radon-222 emission from the tailings disposal
area.
-------
4.2-11
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-------
4.2-12
Table 4.2-6. Working level exposures from radon-222
emissions from model uranium mill
Maximum Regional
Mill area individual population
(WL) (person-WL)
Milling operations 2.3E-4 2.2E-2
Tailings disposal area 4.7E-3 5.2E-1
Total 4.9E-3 5.4E-1
National
population3
(person-WL)
6.7E-2
1.5
1.6
aCalculated from data in ORNL79--Table 2.1 where a one kCi
radon-222 release is estimated to result in an exposure of 8.0E+4
person-pCi/m3 to the population of the United States. This is
equivalent to an exposure to 0.56 person-working levels based on an
assumption of a 70 percent equilibrium of the radon-222 daughter
products (100 pCi/L radon-222 = 0.7 WL) which is considered to be
representative of indoor exposure conditions (Ge78).
-------
4.2-13
Table 4.2-7. Individual lifetime risks and number of fatal cancers
due to radioactive emissions from the model uranium mill
Mill area
Uranium milling operations
Particulates
Radon-222
Tailings disposal area
Particulates
Radon-222
Total
Individual
Maximum
Individual
8.9E-4
3.4E-4
1.7E-3
7.0E-3
9.9E-3
lifetime risks
average individual
Region United States
2.3E-7
9.0E-7 4.5E-10
5.8E-7
2.2E-5 1.1E-08
2.4E-5 1.1E-08
Mill area
Expected fatal cancers
per year of operation
Region United States
(Fatal cancers)(Fatal cancers)
Uranium milling operations
Particulates
Radon-222
Tailings disposal area
Particulates
Radon-222
Total
1.2E-4
4.6E-4
3.0E-4
1.1E-2
1.2E-2
1.4E-3
3.3E-2
3.4E-2
-------
4.2-14
REFERENCES
EPA73 Environmental Protection Agency, 1973, Environmental Analysis
of Uranium Fuel Cycle, Part I, Fuel Supply, EPA-520/9-73-003-B,
Washington, D.C.
EPA76 Environmental Protection Agency, 1976, Environmental Analysis
of the Uranium Fuel Cycle, Part IV, Supplementary Analysis-1976,
EPA-520/4-76-017, Washington, D.C.
Ge78 George A.C., and Breslin, A.J., 1978, The Distribution of
Ambient Radon and Radon Daughters in Residential Buildings in the
New JerseyNew York Area, Presented at Symposium on the Natural
Radiation Environment III, Houston, Texas.
GJ079 Grand Junction Office, 1979, Statistical Data of the Uranium
Industry, GJO-100(79), Department of Energy, Grand Junction, Colo.
Ma78 Magno, P., 1978, Radon-222 Releases from Milling Operations,
Testimony before the Atomic Safety and Licensing Board in the
Matter of Perkins Nuclear Station, May 16, 1978.
NRC77a Nuclear Regulatory Commission, 1977, Final Environmental
Statement Related to Operation of Bear Creek Project, NUREG-
0129, Washington, D.C.
NRC77b Nuclear Regulatory Commission, 1977, Final Environmental
Statement Related to Operation of Lucky McGas Hills Uranium
Mill, NUREG-0357, Washington, D.C.
NRC77c Nuclear Regulatory Commission, 1977, Draft Environmental
Statement Related to Operation of Moab Uranium Mill, NUREG-0341,
Washington, D.C.
NRC77d Nuclear Regulatory Commission, 1977, Effluent Monitoring Re-
quirements for Uranium Mills, Regulatory Guide 4.14, Washington,
D.C.
NRC78 Nuclear Regulatory Commission, 1978, Draft Environmental
Statement Related to Operation of Morton Ranch Uranium Mill,
NUREG-0439, Washington, D.C.
NRC79a Nuclear Regulatory Commission, 1979, Generic Environmental
Impact Statement on Uranium Milling, NUREG-0511, Washington,
D.C.
-------
4.2-15
REFERENCEScontinued
ORNL79 Travis C.C., et al., 1979, A Radiological Assessment of
Radon-222 Released from Uranium Mills and other Natural and
Technologically Enhanced Sources, (NUREG/CR-0573),
ORNL/NUREG-55 Oak Ridge, Tennessee.
ORNL75 Sears M.B., et al., 1975, Correlation of Radioactive Waste
Treatment Costs and the Environmental Impact of Waste Effluents
in the Nuclear Fuel Cycle for Use in Establishing "as Low as
Practicable" GuidesMilling of Uranium Ores, ORNL-TM-4903,
Vol. 1, Oak Ridge, Tennessee.
NRC79b Nuclear Regulatory Commission, 1979, Final Environmental
Statement Related to Operation of the Sweetwater Uranium Project,
NUREG-0403, Washington, D.C.
USFS78 U.S. Forest Service, 1978, Draft Environmental Statement for
Homestakes Mining Company's Pitch Project, Washington, D.C.
-------
4.3-1
4.3 Phosphate Industry
Phosphate rock is the starting material for the production of
all other phosphate products. The basic operations of the phosphate
industry are mining the ore which contains the phosphate rock and
processing it to produce phosphoric acid and elemental phosphorus.
These two products are then combined with various other chemicals to
produce fertilizers, detergents, animal feeds, food products, and
other phosphorus-derived chemicals. The most important use of
phosphate rock is in the production of fertilizers, accounting for
approximately 80 percent of the United States production of the
phosphate rock.
Phosphate deposits contain appreciable quantities of natural
radioactivity, principally uranium-238 and members of its decay
series. Uranium concentrations in phosphate deposits range from 10
to 100 times the concentration of uranium in the natural terrestrial
environment. Uranium concentrations in mined phosphate ores range
up to 10 percent of the concentration in ores currently mined for
the recovery of uranium. The mining and processing of phosphate
ores therefore result in the atmospheric releases of radioac-
tivity and in the production of large quantities of waste materials
which also are a potential source of atmospheric emissions.
Because of the diversity of the operations in the phosphate
industry which may be performed at widely separated locations, the
radiological impact of the industry has been assessed by grouping
the operations that are usually carried out at the same location:
mining and beneficiation, ore drying and grinding, phosphoric acid
production by the wet process, and elemental phosphorus production
by the thermal process.
4.3A Mining and Beneficiation
4.3A.1 General Description
About 120 million metric tons of phosphate ore are mined in
the United States each year. There are about 35 phosphate mines in
the United States, 17 in Florida, 7 in Tennessee, 1 in North
Carolina, 4 Idaho, 3 in Montana, 1 in Wyoming, and 2 in Utah (TVA
74). Florida produces the largest amount of phosphate rock; of the
ore mined in 1973 about 91 percent was mined in Florida. Table
4.3A-1 lists the current phosphate mines in the Bone Valley deposit
area of Florida.
-------
4.3-2
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4.3-3
The United States production of phosphate rock will increase
to about 60 million metric tons "per year by 1995 when it will level
off and then decline after the year 2000 (EPA78).
The uranium concentration in phosphate ore ranges up to 10
percent of the uranium concentration in the ore currently mined for
uranium recovery, but because of the large amount of phosphate ore
mined, the amounts of radioactivity involved, including the wastes,
are approximately the same for phosphate ore processing and uranium
ore processing (Gu75).
Ore used in the production of phosphoric acid usually
requires beneficiation, the process where the phosphate rock is
separated from the matrix by washing, screening, and flotation.
This process results in large quantities of wastes consisting of
equal quantities of sand tailings and slimes.
The data indicate that the radionuclides of the uranium,
actinium, and probably, thorium decay series are in equilibrium and
that the beneficiation process does not significantly alter this
state (Gu75), although beneficiation does produce a redistribution
of the radionuclide concentrations among the phosphate rock, slimes,
and sand tailings. Phosphate rock, which is one-third of the mined
ore, contains 40 percent of the total activity of the mined ore.
The rest of the radioactivity is accounted for in the slimes (48
percent) and in the sand tailings (12 percent).
4.3A.2 Process Description
Mining
The profile of a typical phosphate deposit in the Bone Valley
region of Florida is shown in figure 4.3A-1.
Almost all phosphate ore is mined by strip mining
techniques. A typical mining operation (Ho76) in Florida has one or
more large excavating machines called draglines which mine the ore
in strips measuring 99 m wide by 10 m deep over a length of 60 m to
1000 m. During the initial box-cut the overburden is placed on
virgin ground. As the mining progresses, the dragline returns the
overburden into the adjacent mined-out area. The dragline casts
matrix at the rate of 1100 m3 per hour into a well that is
-------
4.3-4
:1P' J0.5.Q':;
;SAND (OVERBURDEN);..;x.';>::';.'; ;:;:.:.:x:
X-6' 'TO'fo'1
:.PHOSPHATE PART 1C ES
.
(MATRIX
'^.QUARTZ SAND.;-.:-'-.vi:
).\CLAY ^Y:'-V-V/NV.'^
5; TO 50 '
;ro
Figure 4.3A-1. Profile of a typical phosphate deposit
excavated on unmined ground. A dragline production rate of 910
cubic meters per hour is required to feed a plant producing 400
metric tons of finished product per hour. Three high pressure water
guns jet water into the matrix, breaking it into a slurry that
slumps to the lower part of the well where the slurry at 35-40
percent solids is then pumped to the beneficiation plant. Pumping
of matrix slurry (normally over a distance of 3-10 km and in one
case, 16 km) is the standard method of transporting from pit to
plant in Florida.
-------
4.3-5
All mines in Central Florida operate approximately 24 hours a
day since they operate 20 to 21 shifts per week.
The cross section of a mining area is shown in figure
4.3A-2. Due to the topography, the surface area of a mined strip is
about 12 hectares.
About 160 hectares of land are mined per year at a typical
Florida mine removing 10 million cubic meters of overburden and
mining 7 million cubic meters of matrix (EPA78).
Four categories of reclamation are currently in use (Ro77):
two deal with the reclamation of mined areas and involve balancing
the available materials (i.e., overburden and sand tailings) with
proposed reclamation contours; the third method treats special
problems encountered in reclaiming slime ponds; the fourth method
involves mixing sands and clays. The latter method is presently
used by only one company and accounted for only two parcels of
reclaimed land as of 1977.
Beneficiation
In the washing and beneficiation process (Pa78), phosphate
rock is separated from sand tailings and clay slimes through a
series of screening and flotation steps (figure 4.3A-3).
Sand tailings are used for dam construction, slime ponds, or
land reclamation. Slime disposal involves large settling areas (30
to 60 percent of mined lands), and reclamation to support
agricultural uses requires 5-20 years (EPA78, Ro77).
4.3A.3 Emissions of Radionuclides
The only significant radioactive emission from the mining and
beneficiation operations is the release of radon-222. Particulate
emissions do not seem to be a problem. No measurement studies have
been made of the actual releases of radon-222 during these
operations. So for this report it will be assumed that all the
available (20 percent of the total) radon-222 in the ore is
released, one-half of the radon-222 being released during the mining
operation and one-half being released during the beneficiation
process.
4.3A.4 Model Facility
This preliminary assessment of the phosphate industry
estimates the radioactive emissions and the consequent health
-------
4.3-6
Figure 4.3A-2. Cross section of a typical mined-out.
strip of a phosphate mine (Ro77)
-------
4.3-7
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4.3-8
impacts by developing model facilities and assigning appropriate
values for parameters which affect releases of radioactivity to the
atmosphere at each model facility (table 4.3A-2).
Production data are based on current mining practices at
Florida mines (EPA78, Ho76). Data on the radionuclide content of
materials in Central Florida were taken from reports by Roessler and
Guimond (Ro77, Gu75). Radon-222 flux data were obtained from two
reports by Roessler (Ro77, Ro78).
Emissions
Emissions from the model facility, table 4.3A-3, were
calculated by using the data on radon-222 fluxes and areas listed in
table 4.3A-2. Slime ponds, because they have a high moisture
content, are assumed to be negligible sources of radon-222. Other
assumptions are:
1) All of the radon-222 available (20 percent of the total)
is released during the mining and beneficiation process.
2) The amount of radon-222 ingrowth between the time of
mining and the end of the beneficiation process is negligible.
3) Radon-222 emission is the only significant source term.
4.3A.5 Health Impact Assessment of a Model Phosphate Mine
and Beneficiation Plant
Table 4.3A-4 lists estimates of the working level exposures
from radon-222 emissions from the model phosphate mine and
beneficiation plant. These estimates are for a moderate population
density, generic site in Central Florida (Site C, Appendix A).
Estimates of the individual lifetime risks and fatal cancers
in the population resulting from these working level exposures are
given in table 4.3A-5. The number of fatal cancers resulting from
radon-222 emissions per year of plant operation is estimated to be
0.1 for the population living in the region around the plant.
-------
4.3-9
Table 4.3A-2. Model phosphate mine and beneficiation plant
Parameter Value
Total area of mine 4000 ha
Active mining area 480 ha
Undisturbed area 2000 ha
Reclaimed land 520 ha
Slime pond area 1000 ha
Annual production
Ore 9.0E+6 t
7.0E+6 m
Phosphate rock 3.0E+6 t
Overburden removed 1.4E+7 t
Leached zone material removed 5.0E+6 t
Mine lifetime 24 years
Production yeara 12th year
Radionuclide contentb_-(pCi/g)
Overburden (3 m depth)
Uranium-238 3.0
Thorium-232c 3.0E-2
Leached zone material (1 m depth)
Uranium-238 1.9E+1
Thorium-232 l.OE-1
Ore (4 m seam)
Uranium-238 3.8E+1
Thorium-232c 3.0E-1
Radon-222 flux for land types (pCi/m2-s):
Undisturbed area 3.0E-1
Tailings 1.6
Overburden 4.4
Production and emission data are based on the 12th year of
plant operations.
bMembers of the ^^Sy ancj 232j^
-------
4.3-10
Table 4.3A-3. Radon-222 emissions from the model
phosphate mine and beneficiation plant
Source Radon-222
(Ci/y)
Ore removal and beneficiation 7.0E+1
Overburden removal 8.0
Leached zone material removed 1.9E+1
Mined-out area (before reclamation) 5.9E+2
Reclaimed land 6.4E+2
Total 1.3E+3
Table 4.3A-4. Working level exposures from radon-222
emissions from the model phosphate mine and beneficiation plant
Maximum Regional
Source individual population
(WL) (person-WL)
Mine and
beneficiation plant 2.2E-4 4.9
Table 4.3A-5. Individual lifetime risks and the number
of fatal cancers due to radioactive emissions from
the model phosphate mine and beneficiation plant
Individual lifetime risks Expected fatal cancers
Maximum Average per year of operation
Source individual individual (Fatal cancers)
Mine and bene-
ficiation plant 3.3E-4 4.7E-6 1.06-1
-------
4.3-11
4.3B Phosphate Ore Drying and Grinding Facility
4.3B.1 General Description
After phosphate rock has been beneficiated, it is usually
dried and ground to a uniform particle size to process it
efficiently. Ore drying and grinding operations release significant
amounts of particulate material (phosphate rock dust) and are a
source of radon-222 emissions to the atmosphere. Dry grinding is
currently the general practice, but wet grinding methods, which
would reduce dust releases, are being studied.
There are approximately 20 ore drying and grinding facilities
in the United States which process about 50 million metric tons of
phosphate rock annually. A typical facility processes about 2 to 3
million metric tons of rock each year. They are either separate
facilities or associated with phosphoric acid or elemental
phosphorus plants.
The growth of the ore drying operation keeps pace with
phosphate mining operations. The industry is expected to grow from
about 50 million metric tons per year of phosphate rock in the
1970's to about 60 million metric tons per year through the year
2010 (EPA78).
4.3B.2 Process Description
Following the washing process the phosphate rock is
transferred to the drying and storage area where the wet rock is
dried in large rotating drums or on a fluid bed. Wet phosphate rock
entering the ore drying operation weighs 1.4 g/cc with a 6 to 10
percent moisture content and is dried to a 2 to 3 percent moisture
content (Ho76b). After drying the rock is separated according to
size and grade, and then stored. In some cases, material from the
dryers is ground using ball mills before storage.
The predominant airborne emissions are fine rock dust from
drying and grinding operations. Phosphate rock dryers are usually
equipped with dry cyclones followed by a fine particulate
collector. Control devices, such as baghouse collectors (Pa78),
electrostatic precipitators, and wet-type collectors (Sp67),
indicate 98 plus percent collection efficiency.
4.3B.3 Emissions of Radionuclides
Phosphate rock dust is a source of particulate radioactivity
in the atmosphere because the dust particles have approximately the
-------
4.3-12
same specific activity (pCi/g) of uranium-238 and its decay products
as the phosphate rock (Pa78). The radioactivity of the dust
particles varies according to the type of phosphate ore mined. In
Central Florida the concentration of radium-226 in phosphate rock
ranges from 26 to 97 pCi/g (Ro78). The amount of particulate
radioactivity released per year may be calculated by multiplying the
amount of rock dust released per year by the concentrations of the
radionuclides in the phosphate rock. For example, if a facility
releases 1 x 108 grams of rock dust per year and the concentration
of radium-226 in phosphate rock is 40 pCi/g, the facility then
released approximately 4 mCi/y of radium-226 to the atmosphere.
Since there are no data on the amount of radon-222 released
during the drying and grinding operations, it has been assumed that
radon-222 is in equilibrium with radium-226 and that all the
available (20 percent of the total) radon-222 is released during the
drying process.
4.3B.4 Model Facility
In order to estimate radioactive emissions and health impacts
from ore drying and grinding operations, a model facility was
developed by assigning the various parameters that are important in
assessing impacts (table 4.3B-1). These parameters were based on
reports by Partridge (Pa78) and Guimond (Gu75).
Emissions
The atmospheric emissions of radionuclides from the model ore
drying and grinding facility are listed in table 4.3B-2. The
particulate emission rates were based on the amount of rock dust
released per year and the radionuclide concentrations measured in
the dust effluents (Pa78). The radon-222 emission rate was
estimated from the amount of phosphate rock processed per year, the
radium concentration in the rock, and the assumption that the
radon-222 equilibrium with radium-226 was reestablished after
beneficiation and that all the available radon-222 (20 percent of
the total) was released during the operations.
4.3B.5 Health Impact Assessment of a Model Phosphate Ore Drying
and Grinding Facility
The estimated annual radiation doses and working level
exposures resulting from radioactive emissions from the model ore
drying and grinding facility are listed in tables 4.3B-3 and
-------
4.3-13
4.3B-4. These estimates are made for a moderate population density,
generic site in Central Florida (Site C, Appendix A).
Estimates of the individual lifetime risks and the number of
fatal cancers in the population resulting from these doses and
working level exposures are given in table 4.3B-5. The lifetime
cancer risk to the highest exposed group of individuals is estimated
to be about 5 x 10-4 and results primarily from particulate
emissions from the facility. The lifetime cancer risk to the
average individual in the region is estimated to be about 2 x 10-7.
The number of fatal cancers per year of plant operation is
estimated to be 3.6 x 10-3 ^0 the population living in the region
around the plant. Particulate releases and radon-222 emissions from
the facility contribute about equally to this number.
Table 4.3B-1. Model phosphate ore drying and grinding facility
Parameter
Value
Annual process rate (wet rock)
Moisture Content
Wet rock
Dry rock
Plant lifetime
Plant stacks
Stack height
Stack diameter
Particulate releases
Radionuclide Content--(pCi/g)a
Phosphate rock
Uranium-238
Thorium-232
2.7E+6 t
8%
3%
20 years
10 meters
1 meter
1.1E+8 g/y
4.2E+1
4.4E-1
aMembers of the 238|j an(j 232^ decay series are in
equilibrium.
-------
4.3-14
Table 4.3B-2. Emissions of radionucTides from the model
phosphate ore drying and grinding facility
Emissions
Source (Ci/y)
Participates
decay series (each member) 5.0E-3
decay series (each member) 2.0E-4
Radon-222 2.0E+1
Table 4.3B-3. Annual radiation doses from radioactive
particulate emissions from the model ore drying and
grinding facility
Maximum individual Population
Organ (mrem/y) (person-rem/y)
Lung
Bone
Kidney
Liver
Thyroid
G.I. tract
Other soft tissue
5.4E+1
7.9E+1
2.7E+1
1.5E+1
2.5E+1
l.OE+1
2.5E+1
1.7E+1
1.8E+1
9.5
5.0
7.4
3.5
7.5
-------
4.3-15
Table 4.3B-4. Working level exposures from radon-222
emissions from the model ore drying and grinding facility
Maximum
individual
(WL)
Regional
population
(person-WL)
Ore drying and
grinding
5.3E-5
8.4E-2
Table 4.3B-5. Individual lifetime risks and the number
of fatal cancers due to radioactive emissions
from the model ore drying and grinding facility
Individual lifetime risks Expected fatal cancers
Maximum Average. per year of operation
Source individual individual (Fatal cancers)
Particulates 4.5E-4 8.5E-8
Radon-222 8.0E-5 8.5E-8
1.8E-3
1.8E-3
Total
5.3E-4
1.7E-7
3.6E-3
-------
4.3-16
4.3C Phosphoric Acid Plant
4.3C.1 General Description
Phosphoric acid, used mainly in the manufacture of
high-analysis fertilizers, is produced by the wet process method.
At a typical phosphate plant complex, the phosphoric acid is the
starting material for ammonium phosphate and triple superphosphate
fertilizers which are manufactured in other units of the complex.
The total phosphoric acid production in the United States was
about 9 million metric tons of the acid (as phosphorus pentoxide) in
1976, produced from more than 50 million metric tons of phosphate
rock. The waste generated, in the form of gypsum, is five times the
weight of the phosphoric acid (as phosphorus pentoxide) produced.
This means that in producing 9 million tons of phosphoric acid, 45
million metric tons of gypsum per year are produced(EPA74). The
disposal area for gypsum required per year is approximately 1200 cubic
meters for each daily metric ton of phosphorus pentoxide produced. A
large plant produces 1,000 metric tons of acid (as phosphorus
pentoxide) per day. Therefore, 35 plants would require 35 million
cubic meters of storage area per year for the gypsum produced (S168).
4.3C.2 Process Description
Phosphoric acid, which is usually manufactured by the wet
process method, is the basic building block from which essentially all
mixed fertilizer used in the United States is made. The raw materials
used in this process are ground phosphate rock, 93 percent sulfuric
acid and water. Phosphate rock is mixed with sulfuric acid after the
acid has first been diluted with water to a 55 to 70 percent sulfuric
acid concentration. A diagram of the process is shown in figure
4.3C-1. The simplified overall reaction is represented by the
following equation:
3Ca3(P04)2 + 9H2S04 + 18H20
= 6H3P04 + 9CaS04 2H20 (1)
Phosphate rock is not the pure compound indicated above, but a
fluoroappitite material containing minor quantities of fluorine, iron,
aluminum, silica and uranium. Following the reaction in the digester,
the mixture of phosphoric acid and gypsum is pumped to a filter
-------
4.3-17
21-0-0
18-46-0
16-20-0 11-54-0 10-34-0
0-52-0
Figure 4.3C-1. Flow diagram of the wet process
-------
4.3-18
which mechanically separates the participate gypsum from the
phosphoric acid (approximately 30 percent phosphorus pentoxide
concentration). An enormous amount of the byproduct gypsum is
producedeach metric ton of phosphorus pentoxide, as phosphoric
acid, produces approximately 5 metric tons of gypsum. Normally the
gypsum is sluiced with process water from the plant to the disposal
area. The phosphoric acid separated from the gypsum is collected
for further processing (EPA74).
Fertilizer Production
The next step in the process (figure 4.3C-1) is the
production of fertilizer, usually triple superphosphate. The raw
materials used in its manufacture are ground phosphate rock and
phosphoric acid. The basic chemical reaction is shown by the
following equation:
Ca3(po4)2 + 4H3P04 + 3H20
= 3Ca(H2P04)2 2H20 (2)
The two primary raw materials used to produce the ammonium
phosphate fertilizers are ammonia and wet process phosphoric acid.
The chemical reactions involved are indicated by the following
equations:
H3P04 + NH3 = NH4H2P04 (3)
(monoammonium phosphate, MAP)
2NH3= (NH4)2HP04 (4)
(diammonium phosphate, DAP)
The steps involved in producing the ammoniated phosphate fertilizer
are shown in figure 4.3C-1 (EPA74a).
Control Technology
Wet scrubbers are used on all plant stacks to control process
effluents and dust emissions from ore and product handling areas.
Radon-222 emissions from plant operations and gypsum piles are not
controlled.
-------
4.3-19
4.3C.3 Emissions of Radionuclides
Atmospheric emissions from the manufacture of phosphoric acid
are phosphate rock dust from ore handling operations, particulates
of the ammonium phosphate and triple superphosphate products, and
radon-222. In addition, radon-222 is released in significant
amounts from the gypsum pile, since most of the radium-226 contained
in the phosphate rock is found in the gypsum byproduct.
The radionuclide equilibrium that exists in the phosphate
rock is disrupted during the chemical process with approximately one
percent of radium-226, 60 to 80 percent of thorium-230, and 80
percent of the uranium being dissolved in the acid (Gu75). The
dissolved radionuclides are accounted for finally in the fertilizer
products.
Some wet process plants (in Western United States) include a
calcining process to remove organic material from the phosphate
rock. Therefore, this is an additional source of particulates which
include polonium-210 and lead-210 (ORP78).
4.3C.4 Model Facility
The parameters of a model facility needed to estimate the
emissions and the health impacts from phosphoric acid production are
listed in table 4.3C-1. Production data and most other parameters
are based on reports on wet process plants in Florida (Pa78) and in
Idaho (ORP78). The exhalation rate of radon-222 from a gypsum pile
was obtained from a draft report by Morton (Ho79).
Emissions
Emissions from the model facility are listed in table
4.3C-2. Particulate emissions are based on a report by Partridge
(Pa78). Radon-222 emissions from the gypsum pile were determined
from the exhalation rate and area given in table 4.3C-1. The
radon-222 emissions from the plant are based on the amount of ore
processed per year and the assumption that all the radon-222 in
equilibrium with radium-226 contained in the ore is released during
the chemical process.
-------
4.3-20
Table 4.3C-1. Model phosphoric acid plant
Parameter
Value
Annual process rate
Annual production^
Phosphoric acid
Diammonium phosphate
Triple superphosphate (TSP)
Plant lifetime
Plant stacks
Height
Diameter
Flow rate
Gypsum pile
Effective emission height
Area
Amount stored
Radon-222 exhalation rate
Annual particulate releases (grams)
Diammonium phosphate (DAP)
Triple superphosphate (TSP)
Product storage area (DAP+TSP)
Phosphate rock
Radionuclide content--(pCi/g)
Phosphate rockb
Uranium-238
Thorium-232
Phosphoric acid (52%)
Radium-226
Uranium-238
Triple superphosphate
Radium-226
Uranium-238
Uranium-234
Thorium-230
Polonium-210
2.0E+9 kg
4.3E+8 kg
2.0E+8 kg
9.1E+7 kg
20 years
10 meters
1 meter
2400 m3/min
10 meters
4.9E+5 m2
2.0E+9 kg/y
1.6E+3 pCi/m2_min
1.5E+8
1.4E+8
2.1E+8
5.1E+7
4.2E+1
4.4E-1
8.4E+2
5.1E+4
2.1E+1
5.8E+1
5.8E+1
4.8E+1
3.2E+1
See footnotes at end of table.
-------
4.3-21
Table 4.3C-1. Model phosphoric acid plantcontinued
Parameter Value
Radionuclide content--(pCi/g)-continued
Ammonium phosphate
Radium-226 5.6
Uranium-238 6.3E+1
Uranium-234 6.3E+1
Thorium-230 6.5E+1
Thorium-232 4.0E-1
Thorium-238 8.0E-1
Gypsum
Radium-226 3.3E+1
Uranium-238 6.0
Uranium-234 6.2
Thorium-230 1.3E+1
Thorium-232 2.7E-1
Thorium-238 1.4
aWeights given are expressed as weight of
^Members of the 238y ancj 232jn decay series are in
equilibrium.
-------
4.3-22
Table 4.3C-2.
Emissions of radionucTides
phosphoric acid plant
from the model
Source
Emissions
(Ci/y)
Participates
Uranium-238
Uranium-234
Thorium-228
Thorium-230
Thorium-232
Radium-226
Polonium-210
Lead-210
Radium-228
Radon-222
Plant operations
Gypsum pile
2.3E-2
2.3E-2
4.8E-4
2.0E-2
4.8E-4
1.1E-2
a!.2E-2
b!.2E-2
4.8E-4
8.0E+1
4.0E+2
aEmission rate is calculated from particulate releases and the
mpnt.rat inn nf 210pr, maacnvarl in T^P anrl nhncnha-f-a vnrl/
concentration of ^lupn measured in TSP and phosphate rock.
^Measurements of 210pb Were not made; but releases, a
approximation, are assumed to be at the same rate as 210
s a first
Po.
4.3C.5 Health Impact Assessment of a Model Phosphoric Acid
Plant
The estimated annual doses and working level exposures
resulting from radioactive emissions from the model phosphoric acid
plant are listed in tables 4.3C-3 and 4.3C-4. These estimates are
for a moderate population density around a generic site in central
Florida (Site C, Appendix A). The highest exposed group of
individuals is assumed to be located 700 m from the center of the
model plant.
Table 4.3C-5 lists the estimates of the individual lifetime
risks and fatal cancers in the population resulting from these doses
and working level exposures. The lifetime cancer risk to the
-------
4.3-23
highest exposed group of individuals is estimated to be about 1.6 x
10-3. YniS number results from the release of particulates and
radon-222 from the plant and radon-222 emissions from the gypsum
pile. The lifetime cancer risk to the average individual in the
region is estimated to be about 2 x 10-6. The number of fatal
cancers per year of plant operation is estimated to be about 0.05 to
the population living in the region around the plant. This number
is primarily due to radon-222 emissions from the gypsum pile.
Table 4.3C-3. Annual radiation doses from radioactive
particulate emissions from the model phosphoric acid plant
Organ
Lung
Bone
Kidney
Liver
Thyroid
G.I. tract
Other soft tissue
Maximum individual
(mrem)
85
110
37
19
30
13
31
Population
(person-rem)
46
45
21
11
16
7.8
16
Table 4.3C-4. Working level exposures from radon-222
emissions from the model phosphoric acid plant
Plant operations
Gypsum pile
Total
Maximum
individual
(WL)
1.3E-4
5.3E-4
6.6E-4
Regional
population
(person-WL)
3.4E-1
1.7
2.0
-------
4.3-24
Table 4.3C-5. Individual lifetime risks and number of fatal
cancers due to radioactive emissions from
the model phosphoric acid plant
Source
Individual lifetime risks
Maximum Average
individual individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Phosphoric acid
plant
Particulates
Radon-222
Gypsum pile
Radon-222
Total
6.4E-4
2.0E-4
8.0E-4
1.6E-3
2.1E-7
3.3E-7
1.7E-6
2.3E-6
4.5E-3
7.1E-3
3.6E-2
4.8E-2
-------
4.3-25
4.3D Elemental Phosphorus Plant
4.3D.1 General Description
Elemental phosphorus is the starting material for the
nonfertilizer part of the phosphorus industry. Over 87 percent of the
elemental phosphorus produced is used in the manufacture of the high
grade phosphoric acid by the furnace or "dry" process. The rest of
the elemental phosphorus is either used directly or converted to other
chemicals for use by the organic chemicals industry (EPA74b).
There are 9 operating elemental phosphorus plants in the United
States located in Tennessee (3), Alabama (1), Idaho (2), Montana (1),
and Florida (2).
Approximately 500 thousand metric tons of elemental phosphorus
are produced each year (EPA76) from 5 million metric tons of phosphate
rock. About 4.4 million metric tons of slag, a waste product, are
also produced. Slag contains most of the radioactivity originally
present in phosphate rock.
4.3D.2 Process Description
The phosphate rock is crushed, screened, and usually
briquetted; the rock is then fed into calciners where it is heated to
13000Q. Calcining serves two purposes: (1) to burn out organic
material and (2) to heat-harden agglomerates so they will withstand
further processing steps without disintegrating. The calcined nodules
pass through a proportioning building where sized coke and silica are
blended into the material now called the "burden" which then moves
into the electric furnace. The high temperature reaction in the
furnace (14000c~4400°C) drives off two gases, phosphorus and
carbon monoxide. It leaves two molten residues, slag, and
ferrophosphorus (a mixture of iron, of vanadium, chromium, etc.). A
simplified chemical equation for the electric furnace reaction is:
2Ca3(P04)2 + 6Si02 + IOC
= P4 + 10CO + GCaSiOs (5)
Off-gases are treated in electrostatic precipitators for dust
removal, then in a waterspray cooler where the gaseous elemental
phosphorus is condensed, collected in a sump, and pumped to
-------
4.3-26
INPUT
PROCESS
PRODUCTS
& BY-PRODUCTS
'PHOSPHATE\
ROCK
\
1
p
CALCINER
i '
COKE
SILICA
V
CARBON \
MONOXIDE
Recycled
CALCINED
BRIQUETTE
ELECTRIC
FURNACE
PRECIPITATOR
(gaseous)
CONDENSERS
STACK VENT EXHAUST
FERROPHOSPHORUS
ELEMENTAL
PHOSPHORUS SALES
CARBON MONOXIDE
FLARE STACK
Figure 4.3D-1. Flow diagram of the thermal process for
production of elemental phosphorus
-------
4.3-27
storage. Most of the carbon monoxide is used in the calciner but
some is burned to become carbon dioxide, which is released to the
atmosphere via the flare stack. A flow diagram of the process is
shown in figure 4.3D-1.
Control Technology
Wet scrubbers are generally used at elemental phosphorus
plants to control the amount of particulate material and soluble
gases released from plant stacks. There are no controls applied to
the emissions from slag material.
4.3D.3 Emissions of Radionuclides
Particulates and radon-222 are released to the atmosphere
during plant operations. Because most of the radioactivity from the
phosphate rock is transferred to the slag, this waste product is a
source of radon-222. Slag is a molten material as it comes the
furnace and, when crushed after solidification, breaks into chunks
rather than small particles. Thus, the resuspension of particles in
the air is probably not a problem.
Phosphate rock is subjected to a high temperature (13000^)
in the calciner which causes polonium-210 and lead-210 to
volatilize. These radionuclides contribute a significant portion of
the radioactivity being emitted from the elemental phosphorus plants.
The only source of data on atmospheric emissions of
radioactivity from an elemental phosphorus plant is an EPA study
(ORP77). However, this study did not include estimates of radon-222
emissions from the the plant or slag.
4.3D.4 Model Facility
Various parameters were assigned to a model elemental
phosphorus plant (table 4.3D-1) to estimate the emissions of
radioactivity to the atmosphere and the consequent health impacts.
The information in this table was obtained from general phosphate
industry data (EPA74b) and a report of a study at an elemental
phosphorus plant (ORP77). The estimate of the area of the slag pile
was based on the amount of slag produced per year, the assumption
that the plant is 10 years old, and that the height of the pile is
10 meters.
The exhalation rate of radon-222 from the slag pile is a
rough estimate based on a calculation (Section C.2, Appendix C)
-------
4.3-28
which assumes: an emanation power of 0.1 of radon-222, a radium-226
concentration of 32 pCi/g, the fraction of void spaces is 0.4, and
the radon-222 diffusion coefficient is 0.02 cm2/s.
Emissions
The atmospheric emissions of radioactive material from the
model facility are listed in table 4.3D-2. The data on particulate
emissions were taken from a report on an elemental phosphorus plant
(ORP77). The most significant amount of particulate radioactivity
released annually from the plant is 7.4 curies of polonium-210 which
amount to approximately 20 percent of the polonium-210 entering the
process (ORP77). Estimates of radon-222 emissions from the slag
pile were based on the exhalation rate and the slag pile area listed
in table 4.3D-1.
The amount of radon-222 released during the plant operations
was estimated by assuming that the radon-222 is in equilibrium with
the radium-226 in the phosphate rock and that all the radon-222 was
released during the chemical process. Then, knowing the amount of
phosphate rock processed per year and the concentration of
radium-226 in the rock, a rough estimate of the amount of radon-222
released was made.
4.3D.5 Health Impact Assessment of the Model Elemental
Phosphorus Plant
The estimated annual radiation doses and working level
exposures that result from radioactive emissions from the model
elemental phosphorus plant are listed in tables 4.3D-3 and 4.3D-4.
These estimates are for a moderate population density generic site
in Central Florida (Site C, Appendix A). The highest exposed group
of individuals is assumed to be located 1000 m from the center of
the model plant.
Estimates of the individual lifetime risks and fatal cancers
to the population resulting from these doses and working level
exposures are shown in table 4.3D-5. The lifetime cancer risk to
the highest exposed group of individuals is estimated to be about 6
x 10-3 and results primarily from emissions of polonium-210 from
the calciner stacks at the plant. The lifetime cancer risk to the
average individual in the region is estimated to be 5 x 10-6.
The number of fatal cancers per year of plant operation is
estimated to be about 0.1, primarily due to emissions of
polonium-210 from the plant, but the contribution of radon-222 from
the slag pile is also significant.
-------
4.3-29
Table 4.3D-1. Model elemental phosphorus plant
Parameter
Value
Annual process rate
Phosphate
Coke
Sili ca
Annual production
Elemental Phosphorus
Slag
Ferrophosphorus
Fluid-bed prills
Plant lifetime
Plant stacks
Number of stacks
Height
Diameter
Flow rate
Slag pile
Height
Area
Amount stored
Radon-222 exhalation rate
Radionuclide content--(pCi/g)
Raw materials
Phosphate rocka
Coke
Silica
Products and wastes
Slag
Elemental phosphorus
1.6E+9 kg
1.7E+8 kg
1.1E+8 kg
1.1E+8 kg
1.6E+9 kg
2.0E+7 kg
1.6E+7 kg
20 years
10 meters
1.9 meters
3.0E+3
10 meters
100 hectares
1.6E+9 kg/y
15
26 pCi/g 238U
1 PCi/g 238H
1 pCi/g 238y
32 pCi/g 226n,
0.02 pCi/g 226Ra
0.2 pCi/g 210Po
^Members of the uranium-238 decay series are in equilibrium.
-------
4.3-30
Table 4.3D-2. Emissions of radionuclides from
the model elemental phosphorus plant
Source Emissions
(Ci/y)
Particulates
Uranium-238 4.0E-2
Uranium-234 4.0E-2
Thorium-230 4.0E-2.
Radium-226 7.0E-3
Lead-210 2.0E-2
Polonium-210 7.4
Thorium-232 l.OE-3
Radium-228 l.OE-3
Thorium-228 l.OE-3
Radon-222
Plant stacks 4.2E+1
Slag pile 4.5E+2
Table 4.3D-3. Annual radiation doses from radioactive
particulate emissions from model elemental phosphorus plant
Maximum individual Population
Organ (mrem/y) (person-rem/y)
Lung 740 770
Bone 570 440
Kidney 1800 1400
Liver 320 260
Thyroid 120 99
G.I. tract 30 25
Other soft tissue 120 99
-------
4.3-31
Table 4.3D-4. Working level exposures from radon-222
emissions from model elemental phosphorus plant
Maximum
individual
(WL)
Regional
population
(person-WL)
Plant operations 3.4E-5
Slag pile 3.2E-4
Total 3.5E-4
1.7E-1
1.8
2.0
Table 4.3D-5. Individual lifetime risks and the number
of fatal cancers due to radioactive emissions from
the model elemental phosphorus plant
Source
Individual lifetime risks
Maximum Average
individual individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Elemental
phosphorus plant
Particulates
Radon-222
Slag pile
5.2E-3
5.1E-5
3.1E-6
1.7E-7
6.6E-2
3.6E-3
Radon-222
Total
4.8E-4
5.8E-3
1.8E-6
5.1E-6
3.8E-2
1.1E-1
4.3E.1 Summary
Phosphate ore deposits contain a significant amount of uranium
and its decay products. Because of the large amount of phosphate ore
mined each year, the amount of uranium removed along with the
phosphate ore roughly equals that mined each year in the uranium
mining industry.
-------
4.3-32
Phosphate ore processsing results in emissions of radioactive
participates and radon-222 from the processing plants and waste piles.
Table 4.3E-1 summarizes the lifetime risks and the number of fatal
cancers resulting from radioactive emissions from the model phosphate
facilities.
In this preliminary assessment of the phosphate industry, the
most significant particulate emissions come from the model elemental
phosphorus plant which releases polonium-210 from the calcining
operation. The release of about 7 curies per year of this radionuclide
is primarily responsible for the estimated lifetime cancer risk of 6 x
10-3 to tne highest exposed group of individuals and the fatal cancer
rate of 0.1 per year in the population living around the plant.
Emissions of radon-222 from the mining of phosphate ore are also
significantresulting in an estimated 0.1 fatal cancers per year in
the population around the mine.
Table 4.3E-1. Summary of individual lifetime risks
and number of fatal cancers due to radioactive emissions from
the model facilities of the phosphate industry
Individual lifetime risks
Source Maximum Average
individual individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Mine and beneficiation
plant 3.3E-4
Ore drying and grinding
plant 5.3E-4
Phosphoric acid
plant
1.6E-3
Elemental phosphorus
plant 5.8E-3
4.7E-6
1.7E-7
2.3E-6
5.1E-6
l.OE-1
3.6E-3
4.8E-2
1.1E-1
-------
4.3-33
REFERENCES
EPA74a Environmental Protection Agency, 1974, Basic Fertilizer
Chemicals, EPA-440/l-74-011a, Washington, D.C.
EPA74b Environmental Protection Agency, 1974, Phosphorus Derived
Chemicals, EPA-440/l-74-006-a, Washington, D.C.
EPA78 Environmental Protection Agency, 1978, Draft Areawide
Environmental Impact Statement, Central Florida Phosphate
Industry, EPA 904/9-78-006, Atlanta, Georgia.
Gu75 Guimond R. J., Windham S. T., 1975, Radioactivity Distri-
bution in Phosphate Products, By-Products, Effluents, and
Wastes, Technical Note, ORP/CSD-75-3, August 1975, Washington,
D. C.
Hi68 Hill L., 1968, Solving Air Pollution Problems in Fertilizer
Manufacturing Plants, Crop!ife, March 1968.
Ho76a Hoppe R. W., 1976, Phosphates are Vital to Agriculture
--and Florida Mines for One-Third of the World, Engineering
and Mining Journal, May 1976.
Ho76b Hoppe R. W., 1976, From Matrix to Fertilizers: Florida's
Phosphate Industry Girds to Produce Over 50 Million TPY,
Engineering and Mining Journal, September 1976.
Ho79 Horton T. R., 1979, A Preliminary Radiological Assessment
of Radon Exhalation from Phosphate Gypsum Piles and Inactive
Uranium Mill Tailings Piles, Draft, Eastern Environmental
Radiation Facility, Montgomery, Alabama, April, 1979.
ORP77 Environmental Protection Agency, 1977, Radiological
Surveys of Idaho Phosphate Ore Processingthe Thermal Process
Plant, Technical Note, ORP/LV-77-3, November 1977.
ORP78 Environmental Protection Agency, 1978, Radiological Surveys
of Idaho Phosphate Ore Processingthe Wet Process Plant, Techni-
cal Note, ORP/LV-78-1, April 1978.
Pa78 Partridge J. E., Horton T. R., 1978, Sensintaffer E. L.,
Boysen G. A., Radiation Dose Estimates due to Air Particulate
Emissions from Selected Phosphate Industry Operations, Technical
Note, ORP/EERF-78-1, June 1978.
-------
4.3-34
REFERENCEScontinued
RHS78 Radiological Health Services, Florida Department of Health
and Rehabilitative Services, 1978, Inventory of Mining Activity
in Florida and Its Association with Naturally Occurring Radio-
activity, 1978.
Ro77 Roessler C. E., Wethington J. A. Jr., Bolch W. E., 1977,
Natural Radiation Exposure AssessmentRadioactivity of Lands
and Associated Structures, Florida Phosphate Council, Lake-
land, Florida, August 1977.
Ro78 Roessler C. E., Kautz R., Bolch W. E. Jr., Wethington J. A.,
Jr.,1978, The Effects of Mining and Land Reclamation on the
Radiological Characteristics of the Terrestrial Environment of
Florida's Phosphate Regions, Presented to the Symposium:
The Natural Radiation Environment III, Houston, Texas, April 23-
28, 1978.
S168 Slack A.V., Editor, 1968, Phosphoric Acid, Volume 1, Part II,
Marcel Dekker, Inc., New York, 1968.
Sp67 Specht R. C., Calaceto R. R., 1967, Gaseous Fluoride Emission
from Stationary Sources, Chemical Engineering Progress, 63:78,
May 1967.
TVA74 Tennessee Valley Authority, 1974, Fertilizer Trends-1973,
Bulletin Y-77, National Fertilizer Development Center,^ Muscle
Shoals, Ala., June 1974.
USDI74 United States Department of the Interior, Bureau of Land
Management, 1974, Final Environmental Impact Statement, Phos-
Phate Leasing on the Osceola National Forest in Florida, USDI
Int. FES 74-37, Washington, D.C.
-------
4.4-1
4.4 Steam-electric coal-fired generating stations
4.4.1 General Description
Coal is black, soft rock containing at least 50 weight
percent carbon. It also contains sulfur, iron, moisture, and trace
quantitities of naturally occurring radioactive materials. All
radionuclides of each of the three natural series and potassium-40
are found in coal. When coal is burned, the mineral content of coal
is converted to ash and slag. These waste products contain the
radionuclides originally present in the coal. A fraction of the ash
is released to the atmosphere; the quantity released depends upon
the efficiency of the particulate control system, mineral matter
content of coal, furnace design, and applicable emission control
standards. Retained ash may be stored on the station site. These
waste piles are sources of fugitive emissions.
Half of our coal is mined west of the Mississippi River.
Kentucky is the leading coal producing State while Ohio is the
leading coal consuming State.
In 1977 there were 1200 coal-fired units at 395 sites in the
United States. The total installed electric generating capacity of
these units was 226 GW or about 42 percent of the nation's
capacity. Most of these coal-fired units are relatively
smallabout 50 percent (600) have a generating capacity of less
than 100 MW. The remaining units range from 100 MW to 1.0 GW with
about one percent exceeding 1.0 GW capacity. During 1976 electric
utilities burned 480 M metric tons of coal containing about 290
curies of uranium-238 and 250 curies of thorium-232.
At least 230 large units are projected to come on line in the
interval 1980 to 2001. Many of the older and smaller currently
operating units will be retired during this period. The Department
of Energy estimates more than 500 GW of coal-fired capacity will be
on line by 2000. Table 4.4-1 presents the estimated distribution of
new stations.
In addition to coal-fired utility boilers, there are 24,000
fossil-fueled industrial boilers at over 3,000 sites, 750 oil-fired
utility boilers, 975 gas-fired utility boilers, 950 combustion
turbines, 285 refineries, 625 steel mill furnaces, and over 13,000
coal-fed coke ovens.
-------
4.4-2
Table 4.4-1. Proposed new coal fired power plants, 1979-200ia
Fuel Number of unitsb Total capacity (GW)C
Eastern Bituminous
Western Bituminous
Lignite
95
97
18
52
53
10
Total 210 115
aSource: Ri78.
bThe number of units burning Eastern and Western bituminous
coal is based on an arbitrary but quasi-equal division among the
total number of units burning this rank of coal.
cThe capacity estimate is based on 550 MW per unit.
4.4.2 Process Description
Combustion
The technology for producing electricity from the combustion
of coal has been well established for many years. The basic steps
involved are the combustion of the coal in a furnace, the capture of
the combustion heat by a boiler that produces high-temperature steam
under pressure, the expansion of the steam through a turbine that
drives a generator, and the condensation of the steam exhausted by
the turbine (Le77). The overall thermal efficiency of this process
has increased to an average of 35 percent in recent years .
Additional components required for electric power stations
include heat-dissipation devices, such as evaporative cooling towers
and cooling ponds; stack gas cleanup equipment, such as
electrostatic precipitators and scrubbers; and coal preparation
equipment (Le77).
The mineral matter in the coal forms an inorganic
solidtermed ash. A portion of the mineral matter forms bottom ash
or slag. The remaining portion forms fly ash which enters the flue
-------
4.4-3
gas stream. The radioactive materials are analogous to a solute
dispersed in a solventthe flue gas. Particulate emission control
devices (including SOX scrubbers) strip (i.e., clean) the
particulates from the flue gas prior to atmospheric release. The
quantity of particulates released is set by the applicable
environmental standard or New Source Performance Standards.
Particulate release rates (for new sources) may be less than 1.0
weight percent of the particulates in the flue gas stream.
Waste Piles
Solid waste piles (fly ash, bottom ash, slag, scrubber
sludges, etc.) at coal-fired plants range in area from 80 to 100
hectares for a single 550 MW unit. In 1977 about 50 M metric tons
of ash were generated by coal-fired electric generating plants in
the United States. Some of the ash is stored near or on the station
site; some is returned to a coal mine for disposal; and some can, be
used (about 10 to 20 percent of the ash.)
Control Technology for Airborne Emissions
Control technology includes cleaning the coal before it is
used, process controls, and emission controls. Control technology
for new plants is summarized in table 4.4-2. The control
methodology is based on existing technology, current operating
practices, and DOE's fuel developmental programs (EPA72, EPA77,
Bo78, Si77, B&W72).
New power plants can use furnace design features minimizing
fly ash formation. The options for existing stations include those
for new units as well as: (1) early retirement; (2) capacity factor
limitations; (3) use of improved quality coal and (4) back-
fitting with particulate control.
4.4.3 Emissions of Radionuclides
Calculated and measured emission rates are discussed below.
The calculated emission rates are those developed for model
facilities. The measured values were obtained by radioassays of
samples collected by EPA's sampling program.
Calculated Emissions
Models have been developed for existing coal-fired power
plants (units) and stations (RCM-10) and power plants and stations
in the design stage (RCM-1) (table 4.4-3). The RCM-10 and RCM-1
-------
4.4-4
Table 4.4-2 Airborne emissions control strategies for new
coal-fired power plants
Fuel Block
Use high heating value fuels3.
Avoid selection of coal containing high concentrations of
uranium and thorium.
Use low mineral matter and moisture content bituminous.
Clean coal thoroughly (for example, to levels D, E, or F).
(Refer to EPRI78)
Power Generation Block
Minimize unit net heat rate.
Avoid specifying overfeed stokers or dry ash, pulverized
coal units.
Cyclone firing is the preferred mode of firingb.
Slag bottom, pulverized units may be considered.
Air Block
Particulate emission control efficiency should be 99 percent
or better.
If a significant degree of scrubber reheat is specified,
and inline reheaters are not provided, then emission
analysis is indicated to determine if a particulate control
device is required in series with and following the scrubber.
aCoal oil mixtures (COM) are classified as high heating value
fuels.
^The use of a cyclone furnace may be contraindicated by the re-
quirement to limit NOX emissions.
-------
4.4-5
models are based on a survey of data for operating units and units
under construction. Eight variables relating to fuel properties,
furnace design, and emission control methodology determine
radionuclide release rates.
Radionuclide concentrations in the released particulates may
be enriched relative to those in the mineral content of the fuel as
a result of the combustion and emission control processes. Some
typical enrichment factors for fossil fuel power plants are shown in
table 4.4-4. The emission rates calculated for new and existing
model stations incorporate enrichment factors suggested by Beck*
(Be78).
The radon emission rate was determined by assuming all radon
in the coal is released from the stack. Based on a literature
search (Be78, Ka76), four reference coals were developed (table
4.4-5).
The RCM-10 model applies to about two-thirds of the existing
units (about 800 in number--250 stations). Seventy-fire percent of
these stations are believed to be located at urban and suburban
sites. The model is based on a survey of Federal Power Commission
data for generating stations operating as of January 1, 1970
(FPC73). The base includes data for over 300 units; Eastern
bituminous is the model coal.
The RCM-1 model is representative of 300 stations built,
commissioned, or planned sinced 1970. The model was developed by
identifying factors controlling emission rates (table 4.4-6) and
then surveying the literature for new designs (over 700)
commissioned since 1959 with emphasis given to the more recent
designs (306) which include environmental controls (Po75-78, NUS78,
EW77). Western bituminous coal is used in this model.
Annual emission rates per operating year for model coal-fired
stations (new and existing) based on the RCM-10 and RCM-1 models are
listed in tables 4.4-7 and 4.4-8. The radionuclide release rate for
the existing station model exceeds that for the new station model by
almost a factor of three, although the existing station capacity is
*For a new plant (RCM-1), the radionuclide enrichment factor
is 1.0, except for uranium, radium, lead, and polonium, which have
enrichment factors of 2.0, 1.5, 5.0, and 5.0, respectively. For an
existing plant (RCM-10), the radionuclide enrichment factor is 1.0
except for lead and polonium, which each have an enrichment factor
of 2.0.
-------
4.4-6
less than that for the new station. The difference in release rates
is due to participate control efficiency and plant net heat rate.
Measured Emissions
Samples of coal, supplementary fuels, scrubber feedstocks,
slag, bottom ash, retained fly ash, scrubber sludge, and stack
emissions were collected by EPA at about one percent of the nation's
1200 utility boilers located at 13 sites in 9 States. Data defining
operating conditions during the stack sampling period were also
obtained at each station. Factors controlling emission rates (table
4.4-9) were taken into consideration in selecting the stations to be
sampled.
The initial results from this EPA sampling program are
tabulated in tables 4.4-10, 11, and 12. A comparison of calculated
and measured uranium-238 emission rates for several existing units
is shown in table 4.4-13. The technique used to calculate the
emission rates for these units was the same procedure used in the
model facilities.
Analyses of fugitive emission data for coal storage and ash
piles indicate the radon "exhalation rate" is less than that for
soil, as reported by Beck (Be78, p.24).
4.4.4 Model Facilities
Models for existing coal-fired units and stations and new
coal-fired units and stations were discussed in Section 4.4.3 and
are described in table 4.4-3. The model stations consist of three
RCM-10 units for existing stations and a pair of RCM-1 units for new
stations.
Emissions
The radioactive emissions from the model of an existing coal
fired station are shown in table 4.4-7; for the new station, in
table 4.4-8.
-------
4.4-7
Table 4.4-3. Model coal-fired power stations
Parameter
Units
Model
New Existing
(RCM-1) (RCM-10)
Generating Unit Model
Type of unit
Unit capacity
Capacity factor9
Net heat rate
Furnace design
Partition coefficient0
Particulate control
Particulate control efficiency
Stack gas flow rate
Stack height
Stack lip diameter
Stack gas average temperature
Scrubber reheater duty
Plume rise
Rank of coal
Coal source region
Heating value
Mineral matter content6
Sulfur content
Uranium concentration
Thorium concentration
Annual generation^
Annual energy requirement
Annual coal consumption
Annual total ash formations
Annual uranium input
Annual thorium input
NA
MW
%
MJ/kWH
NA
wt%/wt%
NA
wt%
m3/s
m
m
°K
TJ/day
m
NA
NA
GJ/tond
wt%
wt%
ppm
ppm
TWH/y
PJ/y
M ton/y
k ton/y
ton/y
ton/y
SE
550
65
10.3
(b)
20/80
LIS
99.0
440
185
4.4
355
3.3
50
Bituminous
Western US
21.46
12.0
1.0
1.9
5.0
3.13
32.2
1.50
164
2.85
7.5
SE
135
54
13.35
(b)
20/80
ESP
80.0
ND
97
ND
ND
NA
50
Bituminous
Eastern US
26.5
11.0
3.0
1.9
5.0
1.92
25.6
0.968
106
1.84
4.84
See footnotes at end of table.
-------
4.4-8
Table 4.4-3. Model coal fired power stationscontinued
Parameter
Model
Units
New
(RCM-1)
Existing
(RCM-10)
Generating Station Model
Type of generation
Number of units
Station capacity
NA
NA
MW
Base load
2
1100
Base load
3
405
NA Not applicable.
SE Steam electric.
ND Not determined.
LIS Limestone scrubber.
ESP Electrostatic precipitator.
aCapacity factor is the ratio of actual or planned generation to
maximum theoretical generation. The RCM-1 value is an average
value over a 30-year plant life. The RCM-10 value is a single year
value.
^Pulverized coal, dry ash bottom.
cThe partition coefficient is a function of furnace design.
This coefficient represents the quantity of ash going into bottom
ash (or slag) and the flue gas.
dthe unit ton, wherever it appears in this table, means metric
ton.
elnorganic material in coal, often referred to as ash content.
^The annual value for RCM-1 is based on calculating the 30-year
plant life value and dividing this value by 30.
9Ash weight is based on the assumption that the weight of the
mineral matter in coal equals the weight of ash formed.
-------
4.4-9
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-------
4.4-10
Table 4.4-5. Characteristics3 of the reference coals by rank
Coal rank
Anthracite
Eastern
Bituminous
Western
Bituminous
Lignite
Heating
value
(GJ/ton)
30.16
26.50
21.46
19.23
Mineral
content
(wt%)
16.7
11.0
12.0
11.0
Radionucl
Coal
ide
concentration
Ash
Thorium- Uranium- Thorium-
232 238 232
5.4
5.0
5.0
6.3
1.5
1.9
1.9
2.3
32
99
42
57
(ppm)
Uranium-
238
9
38
16
22
GJ/ton giga joules per metric ton.
wt% weight percent.
aThe properties of coal are highly variable. The above values are
possible ones selected from a range of values defining model coals.
-------
4.4-11
Table 4.4-6. Factors which affect emissions of radionuclides
from coal fired power plants
Fuels
Types: Coal, crude oil and fuel oils.
Ranks of coal:
Anthracite
Bituminous
Subbituminous
Lignite
Coal source regions:
Eastern United States
Central United States
Western United States
Furnace Designs
Spreader stoker
Cyclone
Slag-bottom, pulverized coal
Dry ash, pulverized coal
Air Handling Block
Electrostatic Precipitators (ESPs)
Baghouse
Wet venturi scrubber
Wet limestone scrubber
Wet lime scrubber
Mechanical precipitators
-------
4.4-12
Table 4.4-7. Atmospheric emissions of radionucTides from
the model existing coal-fired station (3 RCM-10 units)
Radionuclide
Emissions
(Ci/y)
Uranium Series
Uranium-238
Thorium-234
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
9.7E-2
9.7E-2
9.7E-2
9.7E-2
9.7E-2
6.7E-1
9.7E-2
9.7E-2
9.7E-2
9.7E-2
1.9E-1
1.9E-1
1.9E-1
Actinium Series
Uranium-235
Thorium-231
Protactinium-231
Actinium-227
Thorium-227
Radium-223
Radon-219
Polonium-215
Lead-211
Bismuth-211
Thallium-207
Thorium-232
Radium-228
Actinium-228
Thorium-228
Thorium Series
4.6E-3
4.6E-3
4.6E-3
4.6E-3
4.6E-3
4.6E-3
2.8E-2
4.6E-3
4.6E-3
4.6E-3
4.6E-3
8.4E-2
8.4E-2
8.4E-2
8.4E-2
-------
4.4-13
Table 4.4-7. Atmospheric emissions of radionucTides from
the model existing coal-fired station (3 RCM-10 units)continued
Radionuclide Emissions
(Ci/y)
Radium-224 8.4E-2
Radon-220 5.3E-1
Polonium-216 8.4E-2
Lead-212 8.4E-2
Bismuth-212 8.4E-2
Polonium-212 8.4E-2
Thallium-208 8.4E-2
-------
4.4-14
Table 4.4-8.
the model
Atmospheric emissions of radionuclldesa from
of a new coal-fired station (2 RCM-1 units)
Radionuclide
Emissions
(Ci/y)
Uranium Series
Uranium-238
Thorium-234
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
3.0E-2
1.5E-2
3.0E-2
1.5E-2
2.3E-2
1.9
1.5E-2
3.0E-2
1.5E-2
1.5E-2
7.6E-2
1.5E-2
7.6E-2
Actinium Series
Uranium-235
Thorium-231
Protactinium-231
Actinium-227
Thorium-227
Radium-223
Radon-219
Polonium-215
Lead-211
Bismuth-211
Thallium-207
Thorium-232
Radium-228
Actinium-228
Thorium-228
Thorium Series
5E-3
3E-4
3E-4
3E-4
3E-4
9.9E-4
8.8E-2
7.3E-4
7.3E-4
7.3E-4
7.3E-4
1.3E-2
2.0E-2
1.3E-2
1.3E-2
aSee footnote at end of table.
-------
4.4-15
Table 4.4-8. Atmospheric emissions of radionuc1idesa from
the model of a new coal-fired station (2 RCM-1 units)--continued
Radionuclide Emissions
(Ci/y)
Radium-224 2.0E-2
Radon-220 1.6
Polonium-216 1.3E-2
Lead-212 1.3E-2
Bismuth-212 1.3E-2
Polonium-212 1.3E-2
Thallium-208 1.3E-2
aThese emission rates do not include contributions from any
radionuclides in scrubber reactants.
-------
4.4-16
Table 4.4-9. Parameters controlling emissions from
coal-fired electric power stations
Fuel Block
Type of fossil fuel (coal, oil, gas, SRC, COM)
Rank and grade of coal
Energy content (heating value) of fuel
Mineral matter content (ash content) of fuel
Moisture content of coal3
Sulfur content^
Uranium and Thorium concentration
Energy Production Block
Plant net heat rate3
Furnace design0
Air Handling Block
Particulate removal efficiency
Scrubber reactants (if any)
3A11 fossil fuels are burned as vapors. Before coal is combusted,
the moisture content must be driven off. This step requires energy,
thereby reducing boiler efficiency and in turn lowering overall
efficiency.
bSulfur content relates to the degree of sulfur and SOX removal
required. The quantity of lime or limestone needed by scrubbers is
directly proportional to the sulfur content of the coal as fired.
GStoker, cyclone, or pulverized coal unit designs.
SRC Solvent refined coal.
COM Coal oil mixture.
-------
4.4-17
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4.4-19
Table 4.4-12. Radionuclide emission rates (pCi/s) from stacks
at selected coal-fired steam electric generating stations
Stations3
Radionuclide M-l
Uranium-238
Uranium-234
Thorium-230
Radium-226
Lead-210
Polonium-210
Uranium-235
Thorium- 227
Thorium-232
Thorium- 238
1180
1175
73.8
257
1380
3326
60
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39.3
35.2
M-2
278
350
199
202
734
698
19.2
20.6
73.2
84.7
aM-l West North Central Station
M-2 East North Central Station
M-3 South Atlantic Station (125
M-4 Mountain Station (12.5 MW).
M-3
36.9
39.3
14.3
10.4
65.9
54.6
1.59
1.2
9.9
14.8
(510 MW).
(450 MW).
MW).
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1.64
1.66
1.05
0.73
6.5
6.3
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0.16
0.56
0.79
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3.17
2.96
0.34
2.42
1.38
0.20
0.05
1.77
1.69
NR Not reported.
-------
4.4-20
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4.4-21
4.4.5 Health Impact Assessment of Model Facilities
Estimates of working level exposures and annual radiation
doses resulting from the radioactive emissions from the model
facilities are shown in tables 4.4-14, 4.4-15, and 4.4-16. Because
these coal fired stations are located on sites exhibiting a wide
range of characteristics, estimates are presented for the model
station located at an urban, suburban, rural, and remote site and
the population distribution within 80 kilometers of these sites is
shown in table 4.4-17.
Tables 4.4-18 and 4.4-19 present estimates of the individual
lifetime risks and numbers of fatal cancers to the population
resulting from particulate doses at each of the generic sites for
each model station. The urban site is a conservative selection and
estimates for this site represent an upper limit of the potential
health impacts to a regional population.
The risks from radon-222 are not significant in comparison to
the risks from radionuclides in particulate materials.
For the new coal fired station, the lifetime risk of a fatal
cancer to the ost exposed group of individuals ranged fom about 1 x
10-5 to about 6 x 10"5. The number of fatal cancers per year of
station operation in the regional population ranged from 8 x 10-5
to 0.2.
For the existing coal fired station, the lifetime risk to the
most exposed group of individuals ranged from about 6 x 10-5 to 7
x 10-4. The number of fatal cancers per year of station operation
in the regional population ranged from 4 x 10-4 ^Q ^5^
Table 4.4-14. Working level exposures from radon-222
emissions from the model coal fired stations
New Stations
Site
Urban
Suburban
Rural
Remote
Maximum
Individual
(WL)
5.4E-9
6.8E-9
9.7E-9
6.7E-9
Regional
Population
(person-WL)
2.4E-2
1.9E-3
5.0E-4
4.8E-6
Existing
Maximum
Individual
(WL)
4.0E-9
4.4E-9
6.3E-9
5.2E-9
Stations
Regional
Population
(person-WL)
1.3E-2
7.9E-4
2.2E-4
2.1E-6
-------
4.4-22
Table 4.4-15. Annual radiation doses from radioactive participate
emissions from the model "existing" coal-fired station3
Urban site
Organ Maximum
Individual Population
(mrem/y) (person-rem/y)
Suburban site
Maximum
Individual Population
(mrem/y) (person-rem/y)
Lung
Bone
Kidney
Liver
Thyro i d
G.I. tract
Other soft
tissue
6.7
8.3
4.7
4.7
5.5
4.2
5.6
19000
11000
5200
4600
5500
3700
5700
9.2
11.1
5.9
5.8
6.8
5.1
6.9
1050
940
460
310
360
220
370
Rural site
Organ
Lung
Bone
Kidney
Liver
Thyroid
G.I. tract
Other soft
tissue
Maximum
Individual
(mrem/y)
15.0
62.0
19.0
13.0
17.0
8.6
17.0
Population
(person-rem/y)
300
290
150
97
110
67
120
Remote site
Maximum
Individual
(mrem/y)
8.1
12.0
3.5
2.2
2.9
1.4
2.9
Population
(person-rem/y)
2.7
7.1
2.0
1.0
1.4
0.4
1.4
aThis station contains three RCM-10 model units.
-------
4.4-23
Table 4.4-16. Annual radiation doses from radioactive participate
emissions from the model "new" coal-fired stationa
Urban site
Organ Maximum
Individual Population
(mrem/y) (person-rem/y)
Suburban site
Maximum
Individual Population
(mrem/y) (person-rem/y)
Lung
Bone
Kidney
Liver
Thyroid
G.I. tract
Other soft
tissue
0.88
1.6
0.79
0.77
0.9
0.67
0.9
2600
1300
650
580
650
470
680
1.1
2.1
1.0
0.9
1.1
0.8
1.1
190
160
95
43
49
31
51
Rural site
Organ Maximum
Individual Population
(mrem/y) (person-rem/y)
Remote site
Maximum
Individual Population
(mrem/y) (person-rem/y)
Lung
Bone
Kidney
Liver
Thyroid
G.I. Tract
Other soft
tissue
2.1
16.0
4.6
2.4
2.7
1.4
2.8
51
50
31
16
15
10
16
0.79
2.1
0.61
0.31
0.36
0.18
0.37
0.47
1.5
.43
.17
.18
.063
.18
aThis station consists of two RCM-1 units.
-------
4.4-24
Table 4.4-17. Population distribution of
model coal fired power station sites
Sitea Population15
Urban (Site A) 17,100,000
Suburban (Site B) 2,490,000
Rural (Site D) 592,000
Remote (Site F) 11,900
aSites are described in Appendix A.
bWithin a radius of 80 km.
Table 4.4-18. Individual lifetime risks and number of
fatal cancers due to radioactive particulate emissions from the
model "existing" coal fired plant
Individual lifetime risks Expected fatal cancers
Site Maximum Average per year of operation
individual individual (Fatal cancers)
Urban
Suburban
Rural
Remote
7.0E-5
9.2E-5
7.1E-4
6.3E-5
6.2E-6
2.8E-6
3.5E-6
2.5E-6
1.5
l.OE-1
3.0E-3
4.3E-4
-------
4.4-25
Table 4.4-19. Individual lifetime risks and number
of fatal cancers due to radioactive particulate
emissions from the model "new" coal-fired station
Site
Urban
Suburban
Rural
Remote
Individual lifetime risks Expected fatal cancers
Maximum Average per year of operation
individual individual (Fatal cancers)
1.1E-5
1.4E-5
5.6E-5
8.9E-6
8.6E-7
4.8E-7
5.7E-7
4.7E-7
2.1E-1
1.7E-2
4.8E-3
8.0E-5
-------
4.4-26
REFERENCES
Be78 Beck H.L., Gogolak C.V., Miller K.M., and Lowder W.M., 1978,
Perturbations on the Natural Radiation Environment due to the
Utilization of Coal as an Energy Source, DOE/UT Symposium
Proceedings, Natural Radiation Environment III, Houston, Texas,
1978.
Bo78 Borer T.C. and Karr A.W., 1978, Multimedia Environmental
Control Engineering Handbook: Methodology and Sample Summary
Sheets, EPA-600/7-78-187, Office of Research and Development,
Environmental Protection Agency, Washington, D.C. 20460.
B&W72 Babcock & Wilcox, Inc., 1972, Steam, 38th Rev. Ed., Lynchburg,
Va.
EPA72 Environmental Protection Agency, 1972, Air Pollution
Engineering Handbook, AP-40, U.S. Environmental Protection
Agency, Washington, D.C., 1972.
EPA77 Environmental Protection Agency, 1977, Compilation of Air
Pollution Emission Factors, 3rd ed., AP-42, Part B, OAWM, OAQPS,
EPA, Research Triangle Park, N.C., August 1977.
EPRI78 Electric Power Research Institute, 1978, Coal Preparation
for Combustion and Conversion, Final Report, EPRI AF-791,
Phillips P.O., Principal Investigator for Gibbs & Hill, Inc.,
New York, N.Y. Published by EPRIPalo Alto, California.
EW77 Electric World Utility Handbook, 1977, McGraw-Hill, New
York, N.Y.
FPC73 Federal Power Commission, 1973, Steam Electric Plant Air and
and Water Quality Control Data for the Year Ending December 31,
1970, Based on FPC Form 67, Summary Report, Federal Power Commiss-
ion, Washington, D.C.
Ka76 Kaakinen J.W., Jordan R.M., Lawasani, H. and West R.E., 1975,
Trace Elements in Coal-fired Plants, Environmental Science &
Technology 9 (9): 862:869 (Sept 75).
Le77 Lee H., Peyton T.O., et al., 1977, Potential Radioactive
Pollutants Resulting from Expanded Energy Programs, EPA-600/7-
77-082, Office of Research and Development, Environmental
Protection Agency, Las Vegas, Nevada 89114.
-------
4.4-27
REFERENCEScontinued
NUS78 NUS Corporation, 1978, Commercial Coal Power Plants, Rock-
ville, Md.
Po75to78 Power Magazine, November-1975 through 1978, McGraw-Hill,
New York, N.Y.
Ri78 Rittenhouse R.C., 1978, New Generating Capacity: Where, When
and by Whom, Power Engineering 81 (4): 50-58 (1978).
Si77 Sittig, M., Particulates and Fine Dust Removal, Noyes Data
Corporation, Park Ridge, New Jersey, 1977.
-------
4.5-1
4.5 Metal and Nonmetal Mining and Milling
Uranium and thorium and their daughter products including
gaseous radon radionucTides are naturally occurring constituents of
the earth's crust. Therefore, any activity which involves the
disturbance of the earth's surface can result in some release of
these radionuclides to the atmosphere. Because mining and milling
activities involve the handling and processing of large quantities
of ore removed from below the earth's surface, these activities have
been assessed to determine their potential for release of
radioactive materials to the atmosphere and the health impact
resulting from these releases.
The task of identifying industrial segments to be assessed,
characterizing their emissions, and assessing their health impacts
is extremely difficult because there are approximately 15,000 mines
of many diverse sizes and types. Therefore, to reduce this
assessment to a manageable size, only certain types of industries
were selected.
The industries selected, iron, copper, zinc, clay, limestone,
fluorspar, and bauxite, are relatively large industries and are
considered to have the greatest potential for emitting radioactive
materials into the ambient air. Some of the criteria for selecting
these industries were: number of mines, production rate, working
level concentrations of radioactive materials in the mines,
ventilation rate, and size of waste tailings (table 4.5-1).
4.5.1 General Description
Iron
Crude iron ore production in the United States amounted to
215 million tons in 1975. Open pit mines produced 96 percent of
total output and nearly all the ore was shipped to beneficiation
plants. The average iron content of all crude ore mined in 1975 was
33 percent and the average iron content of all usable ore produced
was 61 percent. The average annual production of crude ore was 3.2
million tons per mine.
Twenty States have iron ore deposits but the Lake Superior
States of Michigan, Minnesota, and Wisconsin produced 85 percent of
the ore mined in 1975.
The iron industry has a projected annual growth rate of 1.0
to 2.0 percent over the period 1975 to 2000.
-------
4.5-2
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4.5-3
Copper
Approximately 263 million short tons of crude copper ore were
produced in the United States in 1975. This total yielded only 1.41
million short tons of copper because almost all of the crude ore is
waste, containing only 0.75 percent copper. Usable ore obtained
from beneficiation contains 25 percent copper.
Six States, Arizona, Utah, New Mexico, Montana, Nevada, and
Michigan accounted for 98 percent of the total copper production in
the United States in 1975. Arizona produced 58 percent of the total
recoverable copper or 813 thousand short tons.
The averge annual output of the significant mines in the
United States was about 10 million tons of crude ore per mine with
open pit mines accounting for 80 percent of mine production.
The growth rate of the copper industry is variable--!.0
percent to 15 percentdepending on the state of the Nation's
economy.
Zinc
About 500 thousand short tons of zinc were produced in the
United States in 1975. The major ore producing States are
Tennessee, New York, Missouri, Colorado, and Idaho, with the States
east of the Mississippi River accounting for one-half of the total
United States production.
Essentially all zinc mines are underground. The crude
domestic ore contains approximately 4 percent zinc, 7 percent usable
ore and 93 percent waste. The usable ore contains approximately 55
percent zinc. The average annual production of crude ore per mine
is about 500 thousand short tons.
The zinc mining industry does not appear to have any
significant growth rate in the immediate future.
Clay is produced in 47 States with Georgia, Texas, and Ohio
the leading producers. In 1975, 1317 clay mines in the United
States produced 49.4 million tons of clay. About 90 percent of clay
production comes from surface mines.
The average annual production per mine was 40 thousand short
tons. There is no waste in clay mining.
-------
4.5-4
Clay is categorized as kaolin, ball clay, fire clay,
bentonite, fuller's earth, or common clay and shalethe latter
category is produced by about 60 percent of the mines.
An annual growth rate of 3.5 percent has been projected for
this industry.
Limestone
The crushed limestone industry converts naturally occurring
limestone deposits to a form that is predominantly used by the
construction industry. Processing plants produced about 6.7 x 107
short tons of crushed limestone in 1975 from more than 2900 quarries
or underground mines in 46 States. Illinois, Texas, Pennsylvania,
Missouri, and Ohio accounted for 38 percent of the 1975 limestone
production.
Crushed limestone plant production ranges from 300 thousand
to 2.4 million short tons per year, with average plant production
about 600 thousand short tons per year. There is no waste
associated with the product.
Since 1945 the total output of crushed stone has multiplied
seven times. The annual growth rate of this industry is
approximately 5 percent.
Fluorspar
Fluorspar producing mines in the United States are located in
Illinois, Utah, Texas, Montana, Nevada, Kentucky and New Mexico.Mine
production in 1975 amounted to about 375 thousand short tons, with
about 75 percent of the production coming from Hardin County,
Illinois. The ore was beneficiated by 8 plants resulting in 132
thousand short tons of recovered material. Fluorspar ore is 70
percent waste. Most mines are underground and the average
production per mine was 15 thousand short tons.
The estimated annual growth rate for the industry is 3
percent.
Bauxite
The United States imported 86 percent of the bauxite consumed
in 1975, mining only 14 percent or 2 million short tons in 12 mines
located in Arkansas, Georgia, and Alabama. Arkansas produced 87
percent of the bauxite mined in 1975.
-------
4.5-5
Bauxite mines in the United States are open pit mines except
for 1 underground mine in Saline County, Arkansas. The ore is 45
percent waste. The industry's growth rate is minimal.
4.5.2 Process Description
The three basic processes associated with these industries
are mining, milling, and smelting. However, an assessment of the
smelting process has not been included because sufficient
measurement data on emissions are not available at this time to
assess the effect of this process. Future assessments will include
emissions from smelting or similar type operations.
Mining
Mining is the initial step in the acquisition of ore and is
performed both underground and above ground. Extraction techniques
vary with minerals and locations of deposits. Mining includes
operations such as drilling, blasting, loading, hauling and removal
of rock. Under some conditions, bulldozing and dredging may be
substituted for drilling and blasting. These operations are
carefully planned and executed because of high costs and the results
affect the operating costs of subsequent steps. A single blast can
fragment thousands of tons of ore. Large shovels with capacities of
7 to 13 cubic meters (9-17 cubic yeards) are used for loading
operations. Large capacity trucks (50-350 tons) are frequently used
for hauling; however, railroad trains containing as many as 160 cars
are often used to haul ore from remote mines to mills (e.g., 50
miles). Each of the foregoing mining operations and transfer points
has the potential for producing radioactive emissions.
Milling
Milling procedures, which are often performed at or near the
mine, enhance mineral recovery. Some of the procedures are
crushing, grinding, screening, blending, concentrating, classifying,
briquetting, sintering, and agglomerating. The physical and
chemical properties of the crude ore determine the milling method.
After recovery of the desirable'concentrate, a large volume of
tailings remains as residue or waste from the milling procedures.
The source of particulate emissions in the processing of ore
are those locations, including tailing piles, where the ore is in
motion or the atmosphere surrounding the ore or its products is in
motion. These locations are mining, crushing, screening, grinding,
bagging, drying, drilling, loading, hauling, blasting and all points
of transfer such as truck and railroad car loading points.
-------
4.5-6
Control Technology
Dust is produced at many of the stages in the mining and
milling processes. Methods to reduce emissions include wet dust
suppression, dry collection and a combination of the two. In wet
dust suppression, moisture is introduced in the material flow
causing fine particulate matter to remain with the material flow
rather than becoming airborne. Dry collection involves hooding and
enclosing dust-producing points and exhausting emissions to a
collection device. Both methods may be combined at different stages
throughout the processing plant. Structures enclosing process
equipment are also effective control methods.
4.5.3 Emissions of Radionuclides
Because little data are presently available on radioactive
emissions from the nonuranium, nonphosphate mining and milling
industry, most of the emissions data used in this section are rough
estimates (first approximations) calculated primarily from the
uranium content of the ore. In a few cases, however, the initial
results of a field study program (described below) were used and
these data are identified in the appropriate tables.
All calculations and assumptions used in developing emission
estimates are presented in the appendices. The general approach
used in these calculations was to estimate the uranium content of
the ore and to extrapolate information on emissions from the uranium
mining and milling industry to those of nonuranium mining and
milling activities based on the uranium content of the ore (i.e.,
normalizing to uranium content). In these calculations, the
assumption was made that uranium-238 was in equilibrium with its
decay products and further, that the same specific activity of
uranium and decay products present in the ore was also present in
the dust emissions.
For all of the industries evaluated, with the exception of
copper, a single best estimate of the uranium-238 content of the ore
was made. For the copper industry, the uranium-238 content is
expressed as a range because information in the literature indicated
that some copper deposits contain elevated levels of the uranium-238
decay series (Fi76).
Since there is a lack of data on the thorium content of ores,
the emissions data used in this section are limited to members of
the uranium decay series. Information is being developed on the
thorium content of ores and future assessments will include impacts
from emissions of thorium and its decay products.
-------
4.5-7
Through measurement studies EPA is presently developing data
on radioactive emissions from mining and milling operations.
Measurements are being made of the emissions of radon-222 and
radioactive participates from mine vents, mill and smelter stacks
and vents and waste piles at several facilities for each of the
industries listed in table 4.5-1. In a limited number of cases, the
preliminary data from these studies are used in this report.
However, for most facilities the data will not be available until
late in 1979 or early 1980. These studies, which are quite limited
in relation to the large size of the mining and milling industry,
will be useful only in characterizing the emissions from these
industries in a general way. However, these studies should identify
any significant problems, existing on a broad scale within the
industries studied, with emissions of radioactive materials.
4.5.4 Model Facility
To estimate the radioactive emissions and health impacts from
the selected mines, mills, and tailings piles, a model mine, mill,
and tailings piles for each industry were developed by assigning
values for the parameters which are important in assessing impacts
(table 4.5-2).
Emissions
The atmospheric emissions of radioactive material from the
model mines, mills, and tailing piles are listed in table 4.5-3.
4.5.5 Health Impact Assessment of Model Mine and Mill
Tables 4.5-4 and 4.5-5 estimate annual working level
exposures and radiation doses resulting from radioactive emissions
from the model mines, mills and tailings piles. The estimates for
the copper mining and milling industry are for a remote, sparsely
populated site in the Southwest; the site has minimal impact on the
general population (Site E, Appendix A). The estimates for all
other industries were based on a rural site in south central United
States (Site D, Appendix A). The individual lifetime risks and
number of fatal cancers estimated to result from the radioactive
doses for the selected industries are shown in tables 4.5-6 and
4.5-7.
-------
4.5-8
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-------
4.5-9
Table 4.5-3. Atmospheric emissions of radionuclides from
the model mines, mills, and tailing piles
of selected industries
Emissions
Source Iron Copper Zinc Clay
Mi ne
Radort-222
Mill
(Ci/y) 67
Ur,-tmurn-238 (yCi/y)a 91
Ta i ! inqs pi
Radon-222
le
(Ci/y) 160
27-1518 115
7-396 3.6
(b)-1500 (b)
18
6
(b)
Emissions
Emissions
Source Limestone
Mini-
Radt,-'-??? (Ci/y) 7
Mil 6
U> r !-m- -<8 (uCi/y)*
Tail : is pi !e (b)
Ra-!' -2?2 (Ci/y)
Fluorspar Bauxite
0.2 10
1.7 9
(b) 7
3 ht same release rate also applies to each member of the
uran-um series decay chain released in the dust emissions.
bOoe.; not differ from background.
-------
Table <*. 5»<'f, W'rr.nq ie\/c"i expo-Mrj, (TO
adon-?22 emissions rrcin moJe"; mine1; ati.l i.ai
piles for Sfifcted indusLri^s5-
Source iirh'vn'd
(WL)
Mining operations:
Iron
Copper (Lowest)
(Highest)
Zinc
Clay
Limestone
Fluorspar
Bauxite
ailings disposa1
Iron
Copper (Lowest)
(Highest)
Zinc
Clay
Limestone
Fluorspar
Bauxite
aEssentiall.y no radon -s r^le-seii In t^e mi;!viq o
NA Not available
-------
4.5-11
Table 4.5-5. Annual radiation doses
emissions from the model mills
Organ
Lung
Bone
Kidney
Liver
Thyroid
G.I. Tract
Other soft
Iron
Maximum
individual
(mrem/y)
0.85
1.26
0.43
0.23
0.40
0.16
tissue 0.41
Population
(person-rem/y)
1.8E-1
1.9E-1
1.2E-1
5.7E-2
8.1E-2
3.8E-2
8.3E-2
from radioactive participate
for selected industries
Copper
Maximum
individual
(mrem/y)
0.12-6.9
0.18-10
0.06-3.5
0.03-1.9
0.06-3.2
0.02-1.3
0.06-3.3
Population
(person-rem/y)
4.6E-4-2.6E-2
1.2E-3-6.5E-2
4.7E-4-2.6E-2
1.7E-4-9.7E-3
3.2E-4-1.8E-2
9.3E-5-5.2E-2
3.3E-4-1.8E-2
Zinc
Maximum
individual
(mrem/y)
3.4E-2
5.0E-2
1.7E-2
9.1E-3
1.6E-2
6.4E-3
1.6E-2
Population
(person-rem/y)
7.2E-3
7.6E-3
4.8E-3
2.2E-3
3.2E-3
1.5E-3
3.3E-3
Clay
Organ
Lung
Bone
Kidney
Liver
Thyro i d
G.I. Tract
Other soft
tissue
Maximum
individual
(mrem/y)
5.6E-2
8.3E-2
2.8E-2
1.5E-2
2.6E-2
1.1E-2
2.7E-2
Population
(person-rem/y)
1.2E-2
1.3E-2
8.0E-3
3.7E-3
5.3E-3
2.5E-3
5.5E-3
Limestone
Maximum
individual
(mrem/y)
5.6E-2
8.3E-2
2.8E-2
1.5E-2
2.6E-2
1.1E-2
2.7E-2
Population
(person-rem/y)
1.2E-2
1.3E-2
8.0E-3
3.7E-3
5.3E-3
2.5E-3
5.5E-3
Fluorspar
Maximum
individual
(mrem/y)
9.4E-3
1.4E-2
4.7E-3
2.5E-3
4.4E-3
1.8E-3
4.5E-3
Population
(person-rem/y)
3.4E-3
3.6E-3
2.3E-3
1.1E-3
1.5E-3
7.2E-4
1.6E-3
Bauxite
Organ
Lung
Bone
Kidney
Liver
Thyroid
G.I. Tract
Other soft
tissue
Maximum
individual
(mrem/y)
8.4E-2
1.2E-1
4.3E-2
2.3E-2
3.9E-2
1.6E-2
4.0E-2
Population
(person-rem/y)
1.8E-2
1.9E-2
1.2E-2
5.6E-3
8.0E-3
3.8E-3
8.3E-3
-------
4.5-12
Table 4.5-6. Individual lifetime risks due to radioactive
emissions from model mines, mills, and tailing piles
for selected industries
Industry
Mines
Maximum Average
individual individual
Mills
Maximum Average
individual individual
Iron
Copper (Lowest)
(Highest)
Zinc
Clay
Limestone
Fluorspar
Bauxite
2.2E-5
1.7E-5
9.4E-4
3.9E-4
4.5E-5
8.7E-6
6.9E-7
1.2E-5
5.9E-7
2.2E-7
1.2E-5
l.OE-6
1.6E-7
6.2E-8
1.8E-9
8.9E-8
7.1E-6
l.OE-6
5.7E-5
2.9E-7
4.7E-7
4.7E-7
7.9E-8
7.1E-7
2.4E-09
1.5E-10
8.5E-09
9.6E-11
1.6E-10
1.6E-10
4.5E-11
2.4E-10
Industry
Tailings
Maximum Average
Individual Individual
Iron
Copper
Zinc
Clay
Limestone
Fluorspar
Bauxite
(Lowest)
(Highest)
5.2E-5
-
9.3E-4
-
-
-
-
9.2E-6
1.4E-6
-
1.2E-5
-
-
-
-
6.6E-8
-------
4.5-13
Table 4.5-7. Estimated number of fatal cancers from
model mines, mills, and tailings piles
Estimated fatal cancers per year of operation
Industry
Mines
(Fatal cancers)
Mills
(Fatal cancers)
Tailings
(Fatal cancers)
Iron 5.0E-3
Copper (Lowest) 1.1E-4
(Highest) 6.2E-3
Zinc 8.6E-3
Clay 1.3E-3
Limestone 5.2E-4
Fluorspar 1.5E-5
Bauxite 7.4E-4
2.0E-5
8.0E-8
4.5E-6
7.9E-7
1.3E-6
1.3E-6
3.7E-7
2.0E-6
1.2E-2
-
6.1E-3
_
-
-
-
5.5E-4
-------
4.5-14
REFERENCES
MESA78 Mining Enforcement and Safety Administration, 1978, Letter
from Aurel Goodwin, Department of the Interior, 4025 Wilson
Blvd., Arlington, Va. 22203, to Charles Robbins, Environmental
Protection Agency, February 2, 1978.
BOM75 Bureau of Mines, 1975. Minerals Yearbook, Volume 1, Metals,
Minerals, and Fuels, Department of the Interior, Washington, D. C.
MSHA78 Mine Safety and Health Administration, 1978, Letter from
D. K. Walker, Department of Labor, P.O. Box 25367, DFC, Denver,
CO 80225, to Charles Robbins, Environmental Protection Agency.
EPA78 Environmental Protection Agency, 1978, Source Assessment:
Crushed Limestone, State of the Art, EPA-600/2-78-004E, Industrial
Environmental Research Laboratory, Cincinnati, OH 45268.
Fi76 Fitzerald, J.E. Jr., 1976, Radioactivity in the Copper Ore
Mining and Dressing Industry: A Preliminary Assessment,
Proceedings of the 10th Midyear Topical Symposium of the Health
Physics Society, Saratoga Springs, New York.
-------
4.6 Rad_pn from Water
4.6A Geot.henna! Ppwei ' ' "!" ':tfr,
4.6A.] General 'es:.i ;.''.'
Geethermi' erf-, ;/, .»-, ;--^,,-ray ..on,,-in;-: ;n the bea4: o" lh.j
earth, is a rc^uiro' ~err -o 'rcred1 inu ,c "ien4" IT ir .jnd public
attention. Gt of h^rr-a i - , ;.' < fhp f0,r! ,,f P|0»- c,pr-i;nq ,,d;, neen
used for y<-:ars ; ! pr '.'' > 'i/.-a ;.hfi/ '.-.! .-er^esinng wa^m paths" at
spas and i-sci-;--,, K^-''- / '-,, * K i - t-nvMV hei.ree ;; .'vine;
invest iqat pd in ^ frvi. ;'-.., ih^re are ".eerenilv I?
power piants -t "he 'n v^ i ' ;t" an i- eet io '; 'opacity oj lir^t 60U
MW(e;. The gpotherm'l (-' !:.- 01 \ at .-.-; iev- ireas ar^ now "i"inq
evaluatpd by i !ie jep^rtv.uj"' - '" "re'";y . tn:J HP ir ,r. ..
4.6.A,.? Process [)>?sc_r i_pv , '
T'ieve a'>; rht'ee p- ! >
1) hvdrothermn I corivoctior .
Hydr ot.;.t::'; I'll 'Oii\-' '"i s/slerni ,!>(- created wher. 3 -^oie'ee of
heat, usually mo'ten i'C': ; ti^gra, cone:-, in euritact with d
permeable >"ock rormatior. conta ininq water which expands and rises
upward as it is heater! by ~i.e niolten rock below.. Above me
permeable roc* layer an Tifjenneable layer ef rork traps the
superheated w^ie-"- Jf !"MH nperineable r<.r-''* 1ayev- contcr^v- cvaeks
through w'nicn v/otpr , -\>, > , i he fijir' v-"'! i appear on t h^ earth's
surface either as stearn ;" ^ vai-or-dorr ;nate,1 sy-teni or <:s no' wate*-
in a liquid-dominated sy',i'.;;
At, o^e-c'-nr, on;y '".t^ ,'.,:.., i-~ d(;')i i n;3t eil syciem in tht; UreM T!
States pro'lii'LS r^owe' i -,[':!'' ''v---*"h- system at """he Ge.ys :"' 'n
Californn. At 'his ' '".;" . -ret ; <. ;in it1" ,y ! (.. :-t earn-p> ne,j., i,-,
wells have been rVM!er e . ;t steai' -';e i-" m i layer ;.r "f ^ ..'"'.'
tied together t.u produr.- - < .ui( ^.te,!^ ro drive ,, tuvti'Me,
Noncondensable ..ases. ''?:: , t idon an:! h., -(r'"opn su^f :ijf . -f"
di-"cha^g*1 d ti'» t"'^ al; ^ ,."<-'' ,.< '. <.<~''\('.< ".\ il es "if'- p'lhf"-' :i- "k...;^(-d
-------
4.6A-2
to the surface water system, or more commonly, are pumped down wells
to recharge the geothermal reservoir. When the turbine is shut
down, the wells supplying that turbine are usually vented directly
to the atmosphere to prevent condensate from building up in the
well. A well lasts about 15 years before it is depleted.
Only two other vapor-dominated systems have been identified
in the United States. The first is the Mud Volcano System in
Yellowstone National Park, Wyoming, and the other is an unconfirmed
system in Mt. Lassen National Park, California. Neither of these
systems is used for commercial purposes.
About 63 high- and medium-temperature (90° Q) not water or
liquid-dominated systems have been identified in the United States.
Although none of these systems is currently being used for the
commercial production of power, several experimental plants are
under development.
Hot Igneous Systems
Hot igneous systems consist of magma (molten rock occurring
near the surface of the earth) and hot dry rock (the solidified
margins around the deposits of magma and the overlying roof rock).
These systems differ from hydrothermal convection systems in that
the rock formations are generally not permeable enough to trap
water. The recovery of geothermal energy directly from magma or
from hot dry rock is not yet feasible, although some development
work is under way.
Conduction-dominated Systems
Most of the earth's heat is transferred to the surface
through the process of conduction through solid rock. The first of
these processes, called "normal gradient," is not presently being
used as a source of geothermal energy.
The second process is called geopressured geothermal
reservoirs. Like the normal gradient, the temperature of a
geopressured resource increases with depth at a constant, normal
rate. The geopressured reservoir differs in being a formation of
methane-saturated water trapped in layers of sand and shale beneath
impermeable rock. The weight of the sediment creates extremely high
water temperatures and pressures. A geopressured zone is known to
exist beneath an area extending from the Rio Grande in Texas to the
mouth of the Pearl River in Louisiana and from several kilometers
inland to the edge of the continental shelf. This resource is
presently in the experimental stage.
-------
4.6A-3
4.6A.3 Emissions of Radlonuclldes
The primary source of radioactivity from geothermal
installations is due to the natural uranium decay chain, most
commonly radium-226 and radon-222. Radium-226 would tend to remain
a water pollution or land disposal problem because it is
nonvolatile; therefore, the primary air pollutant is radon-222, a
gas.
Radon-222 is released when the water or steam from a
geothermal resource contacts the air. In an electric
power-generating system, this could occur at the wellhead before the
well is connected to a turbine-generator or where the effluent
(steam and/or water) from a power-generating system is discharged to
a surface stream or a cooling tower. In a closed-cycle system,
Ahe»"e all the effluent from power generation is returned to the
geothermal resource through a deep well, radon-222 is not a problem
unless noncondensable gases are vented to the atmosphere.
At the present time, radon-222 measurement data are available
for only a few geothermal sites such as The Geysers, the Niland
Geolhermal Test Facility, the Bureau of Reclamation's East, Mesa
Facility, the Chevron Oil Company Project in Heber, California, and
one geopressure well at Vermillion Bay, Louisiana. These data which
are primarily in the form of concentrations in air ejector gases,
noncondensable gases, and steam condensate, are summarized in table
The variation in these data highlights the expected change in
radon-222 over time, its dependency on flow conditions, and in some
part, on the sample collection method. Relating radon-222 to
ncncondensable gases, steam condensate, or unflashed brine is
dependent upon the objectives of the technical study which produced
these data.
At the present time, radon-222 emission estimates are
dva:?able only for The Geysers site since other sites are still in
the development stage.
4.6A.4 Typical Facility
In order to estimate health impacts from geothermal power
sites, The Geysers site was adopted as a typical site and the
parameters from this site were used in assessing impacts (table
4.6A-2). Atmospheric emissions from the typical facility are shown
in table 4.6A-3.
-------
Table 4.6/ui. ^jmrnary o4" radon-?-1; emission:, dat-i from
'jeo+herrn-i nowe' -,'tes
Site Radcn-222 .-'ata References
/.'. -il. ) :K";/L nt Me
jrruje i-j^t;-
20' '}~6l ii! P1' '".. LtvtiTl
. '"Hdc IS'1 L.
33.'-1151 f,Ci/L
Venn i1 h on
3ny , Loins "ana
The Geysers - 1.4 Ci/day released from
11 pc,«r plar,~;s (li^z]
- 2.1 pd'/L nnncondensable (LFE75)
'j&ses at we ! ihead
- 2400-6800 pf.-i/l. nonccndensable (LFR/5)
(jases -3t oi'- e.jertcr
- IGhOO-13000 nCi/L <1C-'K (St75)
r o r o e ,1 . a :r
Niland
Easl Mesa - ]2'*n pi, ^;? - Me, weM f>-1 ;jkD76)
Bureau of - i2u/ pC . kg ^--'^, We M 6--1 (ORP76)
Rec lamdtion - 69 p::i " hr"-,, (^75)
-------
4.6A-5
Table 4.6A-2. Typical geothermal power site
Parameter
Value
Electrical capacity
Availability factor
Number of units
Average electrical
capacity per unit
Stack height
600 Mw(e)
0.9
12
50 MW(e)
(Range 11-106)
10 meters
Table 4.6A-3 Atmospheric emissions of radionucTides
from typical geothermal power sites (AN75)
Radionuclide
Emissions3
(Ci/y)
Radon-222
540
a45 Ci/y from each unit.
Since there is presently only one geothermal power site in
operation (Geysers site), it is difficult to know how representative
the impacts from this site will be of the impacts from future
sites. It is important therefore that any interpretation of the
impact assessment for this category take this limitation into
consideration.
4.6A.5 Health Impact Assessment of a Typical Geothermal Power Site
The estimated working level exposures that would result from
radon-222 emissions from a typical geothermal power site are listed
in table 6.4A-4. These estimates are for a site similar to that of
The Geysers (80 km population, 310,000). The typical geothermal
power site consists of 12 electrical generating units, each
-------
4.6A-6
releasing an equal amount of radon-222. These units are widely
dispersed over a large site; two units are usually located in close
proximity to each other. The working level exposures to the highest
exposed group of individuals were calculated at a location 500
meters from the two units in the predominant wind direction assuming
a combined release of 90 Ci/y for the two units.
Table 6.4A-5 estimates the individual lifetime risks and
number of fatal cancers resulting from these exposures. The
lifetime risk of lung cancer to the highest exposed group of
individuals is about 6.0E-4 and the lifetime risk of lung cancer to
the average individual in the region is estimated to be about
7.0E-6. The annual number of fatal lung cancers is estimated to be
about 0.03.
Table 6.4A-4. Working level exposures from radon-222
emissions from a typical geothermal power site
Maximum
individual
(WL)
Regional
population
(Person-WL)
Geothermal power
site
4.3E-4
1.5
Table 6.4A-5. Individual lifetime risks and number of fatal
cancers due to radon-222 emissions from a typical
geothermal power site
Individual
Maximum
individual
lifetime risks
Average
individual
Expected fatal cancers
per year of operation
(Fatal cancers)
Geothermal power
site
6.5E-4
7.3E-6
3.1E-2
-------
4.6A-7
REFERENCES
An78 Anspaugh L.R. and P. L. Phelps, 1978, Final Report on the
Investigation of the Impact of the Release of Rn-222, Its
Daughters, and Possible Precursors at The Geysers Geothermal Field
and Surrounding Area. Lawrence Livermore Laboratory, under
contract W-7405-ENG-48 to ERDA.
LFE75 LFE Corporation, 1975, Investigation of the Release of
Radon-222, Its Daughters, and Possible Precursors at the Geysers
Geothermal Field and Surrounding Areas. Final Report LFE
Reference No. 16650, 2030 Wright Avenue, Richmond, California,
94804 (March 1975).
ORP76 Office of Radiation Programs, 1976, Radioactivity Associated
with Geothermal waters in the Western United States, Technical
Note ORP/LV-75-8A. Las Vegas Facility, Environmental Protection
Agency, Las Vegas, Nevada, 1976.
Ro78 Robertson D., 1978, Battelle Northwest Laboratory personal
communication with Office of Radiation Programs, Las Vegas, Nev.
St75 Stoker A.K. and P. Kruger, 1975, Radon measurements in
geothermal systems. Stanford University, SFP-TR-4. LFE
Corporation, 2030 Wright Avenue, Richmond, California 94804.
-------
"i ret s .=i: i c ontain
When this
'(S 'l^nte-
3," ia\o. As a
i hi-- aquifer. The
y r., -:; cee^s the
d tn withdraw
:;:; ' ur- Hr /n,, the
--.. ? Jgr.^d to
,-,-_,-,--or - jr 'Oats to
,". i.P' ; r jn i-.i
-h , ,- ,-)-. i r
fn-i. --' d>- ft
-------
4.6B-2
=3
O
0)
CD
> c
S_ LU
X
fO O)
S-
4J .*
C 00
CL)
-------
trays
These
met a 1
4.6B-3
The natural draft type of aerator uses successive slats or
to break up falling water into small droplets or a fine mist.
fine droplets are exposed to
; such as iron or manganese.
air which oxidizes the dissolved
4.6B.3 Emissions of Radionuclides
All ground water contains some quantity of radon 222 as the
result of decay of radium-226, both in the water ^nd th.. rc.-.-l and
soil matrix surrounding the water. Concentrations or radon-£22
measured in public ground water supplies in the United States
(except for New England) ranged up to 50,000 pCi/L (Du76). About 75
percent of the supplies fall within a range from nondetectable to
2,000 pf i .'L and about 5 percent of the supplies
10.000 C
(figure 4.nB-l).
100
80 -
a. 40
100
a- 40
2000 10000
100,000
Maine
(244 Samples )
2,000 10,000 100,000 °°
M
-------
4.6B-4
In order to obtain additional information on radon--222 in
drinking water supplies, the EPA has initiated a sampling and
measurement program for radon-222 in drinking water supplies in the
United States. The initial results of these measurement1: have been
recently reported (EPA/9).
Although no direct measurements o1 radon-22? emissions from
ground water treatment plants have been made, the amount of
radon-222 released from a plant during the aeration proreso can be
estimated from measurements of the raw ar.d treated water-
concentrations. Measurements of this type have indicated that a
large fraction of the radon-222 is released from the wate-r dur'ng
aeration (Pa79). Table 4.6B-1 presents data on the amour,ts of
radon-222
plants.
removed from water durinc aeration at 8 water treatment
Table 4.6B-1.
Radon-222 removal
at water treatment
'rrcm wate'" during feration
plants (fM79'
Location
Plant
Plant
Plant
Plant
Plant
Plant
Plant
Plant
Radon-222 conrentrat
Raw .vater Treat
746
823
260
225
455
235
460
415
ion (pi" i/L )
ed wa1f:r
213
417
70
150
<16
40
100
160
Percent
removed
71
49
73
33
>96
83
78
61
4.6B.4 Model Facility
In order to estimate the potential radioactive emissions and
health impacts from ground vater treatment plants, a model ground
water treatment plant Itabl- 4.6B-?) was -'eveloped by assigning the
various parameters neerpd i: assessing irp),Kf?. Thr model plant
emissions are s^own p"n lab!1 i.f>B-?,
-------
4.6B-5
Table 4.6B-2. Model ground water treatment plant
Parameter
Value
Plant capacitya
Type of treatments
Number of wellsa
Raw water
Radon-222 concentration
Treated water
Radon-222 concentration
Radon-222 removal efficiencya
1.1E+7 liters/day
Aeration: lime-soda ash
softening
2 deep wells (760 meters)
5 shallow wells (15 meters)
1,000 pCi/L
150 pCi/L
0.85
aThese parameters were taken from data on the underground
water supply serving West Des Moines, Iowa (Pa79).
Table 4.6B-3. Atmospheric emissions of radon-222 from
the model underground treatment plant
Radionuclide
Emissions
(Ci/y)
Radon-222
3.4
-------
4.6B-6
4.6B.5 Health Impact Assessment of the Model Underground
Water Treatment Plant
The estimated working level exposures that result from
radon-222 emissions from the model underground water treatment plant
are listed in table 4.6B-4. These estimates are for a moderate
population density suburban type site (Site B, Appendix A).
Table 4.6B-5 estimates the individual lifetime risks and
number of fatal cancers resulting from these working level
exposures. The lifetime risk of lung cancer to the highest exposed
group of individuals is about I x 10-5. The lifetime risk of lung
cancer to the average individual in the region is estimated to be
3 x 10~8. The number of fatal cancers per year of plant operation
is estimated to be 1 x 10~3.
Table 4.6B-4. Working level exposures from radon-222
emissions from model underground watertreatment plant
Maximum Regional
Source individual population
(WL) (person-WL)
Underground water 7.8E-6 6.0E-2
treatment plant
Table 4.6B-5. Individual lifetime risks and number of fatal
cancers due to radioactive emissions from the model
underground water treatment plant
Individual lifetime risk Expected fatal cancers
Source Maximum Average individual per year of operation
individual Region (Fatal cancers)
Radon-222 1.2E+5 3.4E-8 1.2E-3
-------
4.6B-7
REFERENCES
Du76 Duncan D. L., T. F. Gessell, and R. H. Johnson, Jr., 1976,
"Radon-222 in Potable Water." Submitted for publication in
Proceedings of the Health Physics Society 10th Midyear Topical
Symposium: Natural Radioactivity in Man's Environment.
EPA79 Environmental Protection Agency, 1979, Environmental Radiation
Data, Report 16, Eastern Environmental Radiation Facility, Environ-
mental Protection Agency, Montgomery, Ala. , April 1979.
Pa79 Personal communication with Jennings Partridge, Eastern
Environmental Radiation Facility, Environmental Protection
Agency, Montgomery, Ala., 1979.
-------
CHAPTER 5
OTHER SOURCE CATEGORIES
This section includes source categories not within the scope
of Chapters 2, 3, and 4--United States Army reactors, United States
Navy nuclear shipyards, and particle accelerators. In general,
these are minor sources of radioactivity.
5 . 1A U.S. J\rmy Facilities
5.1A.1 General Description
The U.S. Army Test and Evaluation Command operates two
reactors which are used principally for nuclear weapons radiation
effects testing of Army and DOD related materiel. One reactor is
located at the Army Pulse Radiation Facility (APRF) on the site of
the Aberdeen Proving Grounds in Maryland. The second, the Fast
Burst Reactor (FBR) is located on the Army's White Sands Missile
Range in New Mexico.
In addition to the pulse type reactors, the Army also
maintains surveillance at three nuclear power plant reactors that
have shut down and been decommissioned. Two of these reactors, the
MH-1A and SM-1 are located at Fort Belvoir, Virginia. The other
reactor, the SM--1A, is located at Fort Greely, Alaska.
Environmental monitoring programs are maintained around these sites
to detect and evaluate any significant changes in the radiation
levels that may be due to emissions from the reactors.
The Army had a TRIGA MARK F research reactor, operated by the
Diamond Ordinance Radiation Facility (DORF), which was shut down for
decommissioning in October 1977. This research reactor is located
in the Forest Glen section of Walter Reed Army Medical Center in
Washington, D.C.
5 . 1 A . 2 Process D esc r i p t i on
The pulse reactors operated by the Army's Test and Evaluation
Command are bare, unreflected, unmoderated, and fueled with enriched
uranium, The reactors are capable of self-limiting, super-prompt-
critical pulse operations and steady-state operations at low power
levels (less than 10 kW). Table 5.1A-1 summarizes the modes of
operations of the two reactors.
Forced air cooling systems are used to reduce the temperature
of the cores in order to minimize the time between operations. The
air exhausted from the reactor buildings is passed through HEPA
filters before being released.
-------
5.1A-2
Table 5.1A-1. Number and modes of operation of the
Army reactor facilities, 1976, (Aa77, De76)
Number of operations
Type of Operation APRF FBR
Pulse 237 277
Steady state 236 30
Unscheduled terminations - 11
Total 473 318
5.1A.3 Emissions of Radionuclides
Each reactor operation produces airborne radioactivity in the
reactor building due to neutron activation of material in the air
and release of fission products from the core. The airborne
radioactivity due to fission products from the core is normally
small compared to that due to activation. Emissions of radioactive
particulates are reduced by the HEPA filters in the exhaust system.
The gaseous radioactivity emitted to the atmosphere is principally
argon-41 from air activation. Table 5.1A-2 summarizes the emissions
from the two sites.
5.1A.4 Health Impact Assessment
No dose estimates were reported for the Fast Burst Reactor at
White Sands Missile Range.
At the Aberdeen reactor, gamma doses, from one pulse of
2.0E+17 fissions, were reported for distances of 411 and 1372
meters. The reported doses were 4.7 mrem and 0.008 mrem,
respectively.
The maximum annual gamma dose for 1976 was reported to be 500
mrem at a film badge area monitoring station located 1.6 kilometers
south of the APRF reactor. This point is within the Aberdeen
Proving Ground Military Reservation. A person who continually
occupied this location would have a lifetime risk of fatal cancer of
7.0E-3. No population dose assessments were reported for the area
surrounding the APRF.
-------
5.1A-3
Table 5.1A-2. Emissions of radionuclides
from Army pulse reactors, 1976 (Aa77, De76)
Month
Apr i 1
May
June
July
August
September
October
November
December
Total
APRF
Gross beta
concentrations9
(yCi/cm3)
1.4E-15
8.6E-14
3.2E-14
3.6E-14
6.1E-14
3.5E-14
6.0E-14
9.5E-14
1.6E-13
FBR
Gross gaseous
activity
(Ci)
January
February
March
6.3E-14
1.1E-13
1.4E-15
4.9E-1
1.2
6.3E-1
1.1
1.1
1.2
3.5
9.0E-2
1.3E-1
7.7E-1
8.5E-1
6.1E-1
1.2E+1
aOnly average monthly gross beta concentrations in the reactor
building stack were reported. No volumetric air exhaust rate or
total emissions were reported.
-------
5.1A-4
REFERENCES
Aa76 Aaserude R. A. 1976, APRF Army Pulse Radiation Facility
Environmental Radiological Monitoring Plan, Aberdeen Proving
Ground, Md.
Aa77 Aaserude R. A., R. W. Dickinson, H. G. Dubyoski, and A. H.
Kazi, 1977, APRF, Army Pulse Radiation Facility, 1976 Annual
Operating Report, Aberdeen Proving Ground, Md.
De76, De La Paz A., and R. W. Dressel, 1976, White Sands Missile
Range Fast Burst Reactor Facility Annual Operating Report, Janu-
ary-December 1976, White Sends Missile Range, N.M.
Gi77 Gieseler W. [_., E. D. McGarry, and B. R. Adcock, 1977, Opera-
tions Report of the Diamonc Ordnance Radiation Facility Nuclear
Reactor, Report No. 11, 1 January 1976 to 31 December 1976,
Adelphi, Md,.
Sc77a Schweitzer J. G., 1977, 1976 Annual Post-Decommiss'ioning En-
vironmental Monitoring Report for the Decommissioned SM--1 Nuclear
Power Plant at Fort Belvoir, Va., 22060, FESA-OD-7703, Fort Belvoir
Va.
Sc77b Schweitzer J. G., 1977, 1976 Annual Site Surveillance Report
for the Decommissioned SM-1A Nuclear Power Plant at Fort Greely,
Alaska, FESA-oD-7704, Fort Belvoir, Va.
-------
5.1B-1
5.IB U.S. Navy Facilities
5.1B.1. General Description
The United States Navy had 113 nuclear powered submarines and
11 nuclear powered surface ships in operation at the end of 1978.
The ships are powered by pressurized water nuclear reactors.
Associated with the operation of these ships are support facilities
involved in construction, maintenance, overhaul, and refueling of
the nuclear propulsion plants. These facilities include nine
shipyards, thirteen tenders, and two submarine bases.
5.IB.2 Process Description
Shipboard nuclear reactors primarily release small amounts of
radioactivity in liquid discharges. However, less than 100 curies
per year of carbon-14 is released into the atmosphere from U. S.
naval nuclear powered ships and their supporting facilities. Most
of the carbon-14 is released at sea, over twelve miles from shore.
The naval nuclear reactors and their support facilities involved
with handling radioactive materials have air exhaust systems
equipped with HEPA filters to reduce emissions. The exhausted air
is monitored during discharge. The concentrations of radioactivity
in air exhausts were below those levels normally present in the
atmosphere.
5.IB.3 Emissions of Radionuclides
Table 5.1B-1 summarizes estimates of airborne emissions
(Mi79) from a typical nuclear naval shipyard. These estimates, used
in environmental pathway analysis, are higher than any measurements
made in the past five years from any shipyard.
5.IB.4 Health Impact Assessment of a Typical Naval Shipyard
Table 5.1B-2 summarizes the annual total body doses to the
maximum individual and to the population within 80 kilometers of a
typical nuclear naval shipyard. The individual dose estimates were
based on the estimated releases shown in table 5.1B-1. The
population dose was based on both liquid and airborne emissions.
The individual lifetime risks and health effects from the operation
of a nuclear naval shipyard are summarized in table 5.1B-3.
Assuming all nine shipyards have similar health risks, the total
number of fatal cancers from radiological work at all naval
shipyards is estimated to be less than 1.8E-3 for each year of
operation.
-------
5.1B-2
Table 5.1B-1. Estimated atmospheric emissions of radionuclides
from a typical nuclear naval shipyard (M179)
Radionucl ide Emissions
(Ci/y)
Argon-41 4.1E-1
Cobalt-60 l.OE-3
Tritium l.OE-3
Carbon-14 l.OE-1
Krypton-83m 2.0E-2
Krypton-85m 2.4E-21
Krypton-85 l.OE-3
Krypton-87 5.0E-2'
Krypton-88 2.0E-21
Xenon-131 5.0E-3
Xenon-133m l.OE-2
Xenon-133 2.1E-1
Xenon-135 2.5E-1
Table 5.1B-2. Annual radiation doses from a typical
nuclear naval shipyard (Mi79)
Maximum
individual Population9
Organ (mrem/y) (person-rem/y)
Total body <2.0 <1.0
aTo the population within 80 km.
-------
5.1B-3
Table 5.1B-3. Individual lifetime risks and number of fatal cancers
from radioactive emissions at a typical nuclear naval shipyard
Individual lifetime risks Expected fatal cancersa
Source Maximum Average per year of operation
individual individual (Fatal cancers)
Nuclear
shipyard <2.8E-5 <1.4E-8 <2.0E-4
aTo the population within 80 km.
-------
5.1B-4
REFERENCES
Mi79 Miles M. E., G. L. Sjoblom, J. D. Eagles, 1979, Environmental
Monitoring and Disposal of Radioactive Wastes from U. S. Naval
Nuclear-Powered Ships and Their Support Facilities, 1978, Report
NT-79-1, Naval Sea Systems Command, Department of the Navy,
Washington, D. C.
-------
5. ?. 1 Genera 1 Desc r ijrti on
Accelerators are devices for imparting high kinetic energy to
electrons or positively charged particles (such as alpha particles,
protons, and deuterons; b> electrical or magnetic fields. In a
typical operation, the accelerated particles travel in an evacuted
tube or enclosure. ":he narticles impinge on a metallic or gaseous
target, producing secondary radiation.
There arK three basic accelerator designs categorized
according to the means used to achieve the particle velocity:
(1) constant direct current (DC) field machines, (2) incremental
acceleration machines, and (3) magnetic field accelerators.
The constant DC field machines, sometimes called "Potential-
drop" accelerators, operate at very high voltages. Charged
particles achieve greater kinetic energy as they are accelerated
through the direct current electric field toward the target. These
accelerators are named according to the type of power supply used to
generate the high DC voltage they require. The principal design
types in this class are the Van de Graaff, Cockcroft-Walton,
Dynamitron, Resonant Transformer and Insulating Core Transformer.
Incremental accelerators accelerate particles by temporally
varying electric fields which impart kinetic energy in discrete
increments. Therefore, the velocity of the particle increases
stepwise rather than in a continuous manner. The principal design
types of this class are the linear accelerator (linac) and the
cyclotron.
The betatron is the only type of magnetic field accelerator.
Electrons are continuously accelerated to desired energy levels in
the betatron b/ a temporal variation in the intensity of a magnetic
field.
Accelerators have a variety of applications. Potential-drop
machines are commonly used for radiography, activation analysis,
food sterilization and preservation, industrial processing (ion
implantation), radiation therapy, and research. Electron linear
accelerators are widely uc,tjd in research by universities,
-------
5.2-2
government, Industry, and medical laboratories. Accelerators have
become increasingly popular in medical radiation therapy as
replacements for teletherapy sources (e.g., cobalt-60). Cyclotrons,
which accelerate the heavier positively charged particles, are
commonly used in physics laboratories and in the radiopharmaceutical
industry. The betatron has many of the same applications; as the
electron linacs and electron accelerating potential-drop
accelerators.
Estimates of the number of particle accelerators in the
United States were published by the Bureau of Radiological Health
(BRH78). In 1977, over 1100 accelerators were reported in use in
this country, not including Federally-owned accelerators., Most of
the very high energy physics research accelerators are owned by the
Department of Energy and are briefly discussed in Chapter 3.
Linear accelerators are the most widely used machines in the
United States; about 70 percent of the linacs are used in medical
applications. The percent, by type of machine, of the total number
of accelerators reported is as follows: linacs, 50 percent; Van de
Graaff, 15 percent; Neutron Generators, 17 percent; Resonant and
Insulating Core Transformers, 6 percent; Cyclotrons, 3 percent;
Betatrons, 6 percent; Cockcroft-Walton (ion implanters), 3 percent.
5,2.2 Process Description
At research facilities there is diversity in operational
modes. The size or energy of the accelerators, the type of particle
accelerated, and the target material used are important variables
which make it difficult to predict airborne effluvent composition
over any extended time period. Also, there are a large number of
radionuclides that may be present in airborne emissions as a result
of nuclide production at medical or radiopharmeceutical facilities
(Section 2.3 of this report).
Possible sources of airborne contamination at accelerator
facilities include: (1) loss of target integrity, (2) handling of
targets in laboratory hoods and glove boxes, and (3) activation of
dust and room air.
Loss of target integrity may involve rupture of powder
targets, flaking-off of the activated surface layer of solid
-------
5.2-3
targets, or leakage of flow-type targets. The vacuum pump exhaust
may be an especially important pathway for gaseous radionuclides,
particularly tritium. Potential airborne radioactive contaminants
from a loss of target integrity depend on such factors as target
composition, particle acceleration, energy, and time of irradiation.
Although not directly related to the actual operation of the
accelerator, radiochemical operations in preparing targets for use
in the accelerator and in handling exposed targets in laboratory
hoods and glove boxes are potential sources of airborne radio-
nuclides at an accelerator facility.
Interaction of primary and secondary particles with dust in
the air is another potential source of emissions of radioactive
materials from accelerator facilities. This source may be minimized
by good housekeeping procedures in the accelerator room and by
filtration of the ventilation air at either intake or exhaust or
both. Another source of radioactive dust particles is nuclear
interaction with the structural material of the accelerator.
Welding, soldering, and working with accelerator parts may initiate
radioactive dust and vapors in the air presenting a potential
internal contamination problem for occupational exposures and
probably contribute little radioactivity to the environment.
If the primary particle beam emerges into the air before
striking the target and has sufficient energy, nuclear reactions
with gases in the air will occur. This interaction of the primary
and secondary particles with air in the accelerator hall will vary
depending upon the hall size, ventilation rate, type of particle
accelerated, energy, and the nature of the target. Protons and
heavy ions accelerated to high energies produce nuclear reactions
directly. All energetic accelerated particles give rise to protons
and neutrons as secondary radiations from interaction in various
targets. Table 5.2-1 lists some radionuclides and their half-lives
that are associated with dust and air activation at accelerator
facilities.
5.2.3 Emissions of Radionuclides
Emissions of radioactive materials at accelerator facilities
are principally dependent on production rates of air activation
-------
5.2-4
products which are functions of the following factors:
1) The incident flux or secondary flux produced
by the accelerator;
2) The density of the target gas atoms in the beam
path;
3) The appropriate cross-section for the reaction
at the interaction energy;
4) The buildup factor, which is based on exposure
time. Its maximum is at the equilibrium
condition;
5) The decay that occurs between termination of the
irradiation and the time of exposure to the radio-
nuclides.
The constant field machine delivers a beam in the energy
range of 1 to 19 MeV, which is, for the most part, below the
threshold energies for air activation shown in table 5.2-1. Gaseous
tritium, however, can be released from these machines when they are
used to produce neutrons. Neutrons are generated by bombarding a
Table 5.2-1. Half-lives of activation radionuclides produced
in accelerator facilities
Isotope Half-life
Oxygen-15
Nitrogen-13
Nitrogen-16
Oxygen- 14
Carbon-11
Argon-41
Beryllium- 7
Tritium
Sulfur-38
Fluorine-18
2
10
7
1
20
1.9
53
12
37
2
min
min
s
min
min
h
d
y
min
h
-------
5.2-5
tritium target with deuterium. During this process tritium may be
knocked off the target and released to the room or atmosphere
through the vacuum pump exhaustabout 100 to 300 mCi per target.
Typical targets contain about 5 Ci of tritium. The small neutron
flux produced (about 50 n/cm2/s) is not considered large enough to
produce significant activation products.
Of the incrementally accelerated machines, the most popular
is the 18 MeV medical electron linac. At this energy the photons
produced are usually below the threshold energies for air activation
shown in table 5.2-2. However, it is possible to use the measured
thermal neutron flux around an 18 MeV electron linac to calculate
the amount of argon-41 and carbon-14 produced in the air. The flux
was reported to be about 440 n/cm2/s. Assuming 2000 hours of
operation per year, the activity produced in a room 27 m3 is
l.OE-4 and l.OE-9 curies, respectively.
Measurements at a 40 MeV linac detected trace amounts of
chlorine-39 and argon-41. The common air activation products,
carbon-11, nitrogen-13, and oxygen-15, were also found to be
produced at rates that ranged from <0.1 uCi per pulse of 20 MeV
electrons without a bremsstrahlung target to 2 uCi per pulse of 45
MeV electrons with a bremsstrahlung converter. The total facility
release quantities were reported to be in the curie range.
Production rates of oxygen-15, nitrogen-13, and carbon-11 in
the presence of a high energy neutron flux at a cyclotron facility
have been estimated. The conditions assumed were a 1 yA proton
beam striking a thick beryllium target at 100 MeV. This arrangement
produces neutrons at a rate of of 2.0E+12 n/sec with energies
distributed around 100 MeV. Assuming a 1 cm? beam traversing 1
meter of air, and using the abundance of carbon, nitrogen, and
oxygen isotopes in air, production rates were calculated (table
5.2-3). Decay during production was not considered since activation
products are free to be exhausted immediately after production. An
estimate of the thermal neutron flux is not available for this
situation; thus, production rates of argon-41 and other (n,x)
products cannot be calculated.
Table 5.2-4 lists the estimated annual emissions from typical
particle accelerator facilities.
-------
5.2-6
Table 5.2-2. Nuclear reactions responsible for some airborne
radioactivity
Reaction
(y,n)
(Y,n)
(Y.n)
(n,2n)
(n,2n)
(n,2n)
(n,pn)
(n,Y)
Parent
nuclide
Nitrogen-14
Oxygen-16
Carbon-12
Nitrogen-14
Oxygen-16
Carbon-12
Oxygen-16
Nitrogen-14
Argon-40
Isotope
produced
Nitrogen-13
Oxygen-15
Carbon-11
Nitrogen-13
Oxygen-15
Carbon-11
Oxygen-15
Nitrogen-14
Argon-41
Threshold
energy
(MeV)
10.5
15.7
18.7
11.3
18.0
20.0
10.0
10.0
NA
NA Not applicable.
Table 5.2-3. Production rates of oxygen-15, nitrogen-13, and
carbon-11 at a 100 MeV cyclotron producing 2.0E+12 neutrons
per second
Reaction
12C(n,2n)nC
14N(n,2n)13N
160(n,2n)150
Production rate
(atoms/s)
6.0E+4
5.0E+7
9.0E+7
Amount produced
in 4 hours
(C1)
7.0E-5
5.3E-2
9.4E-2
-------
5.2-7
The most popular accelerator is of the incremental or cyclic
type. Half of the accelerators in the United States are linacs, 90
percent of which are electron linacs in the energy range of 1 to 19
MeV and 1 to 10 kilowatts in power. These characteristics are
typical of cancer therapy machines widely used in hospitals. For
this reason the 18 MeV medical linac is considered to be a typical
cyclic accelerator.
The betatron is a unique accelerator class. However, its
radiation characteristics and applications are similar to electron
linacs. The production of airborne radioactivity by the typical
linac can also be applied to the betatron. Therefore, a typical
magnetic field machine will not be discussed.
5.2.4 Typical Facility
The information for a typical facility is generalized to the
various classes of accelerators. Figure 5.2-1 shows the range of
operating characteristics for various types of accelerators.
Although accelerators vary within a given class, some generaliza-
tions can be made.
Of the constant potential type machines, the Van de Graaff
accelerators are the most prevalent. About 15 percent of all
registered accelerators in 1977 belonged in this category.
Furthermore, 78 percent of the Van de Graaff accelerators were in
the energy range of 1 to 19 MeV with intensity ranging from 1 yA
to 1 mA. To estimate health impacts, a 6 MeV Van de Graaff using a
tritium target is considered as typical of a constant potential type
of machine.
The other type of cyclic accelerator, the cyclotron,
generally operates at higher energies and power levels than other
machines in use and therefore has the potential of producing greater
quantities of airborne raidoactivity. A typical research cyclotron
accelerates ions at about 100 MeV and at very low currents into
scattering targets to produce secondary radiation. The secondary
radiation is in the form of hard gamma rays (>10 MeV). These
gammas are presumably responsible for producing nitrogen-13,
oxygen-15, and carbon-11. At lower energies, cyclotrons would
probably not produce secondary radiation with sufficient energy to
produce even nitrogen-13 via the (y>n) reaction. Although increased
research is being performed in the area of neutron production for
medical applications, it is difficult to estimate the annual
utilization of a cyclotron for neutron beam generation. For typical
cyclotron facilities, then, it will be assumed that 2 percent of the
yearly operating time (1000 hours) is allocated to neutron
production.
-------
5.2-8
ICT3
I- LOW VOLTAGE TRANSFORMERS
2-ELECTROMAGNETIC SYSTEMS
3-VAN DEGRAAFFS
4-TANDEM ACCELERATORS
5-BETATRONS
6-ELECTRON UN ACS
7-ION LINACS
8-CYCLOTRONS-
9-ELECTRON SYNCHROTRONS
O-PROTON SYNCHROTRONS
O.I
1.0
10
KT IOJ 10"
Particle Energy (MeV)
Figure 5.2-1. Operating characteristics of accelerators.
(Particle beam intensity versus particle energy, for several
types of accelerators. The dark-toned areas relate to electron
and ion accelerators; middle tones to ion accelerators only;
light tones to electron accelerators only.)
-------
5.2-9
Table 5.2-4 lists the estimated annual emissions from typical
accelerators. Table 5.2-5 summarizes the characteristics of a
typical accelerator facility.
Table 5.2-4. Estimated annual emissions from typical
particle accelerators (Te79)
Radio-
nuclide
Carbon-11
Nitrogen-13
Oxygen-15
Tritium
Carbon-14
Argon-41
18 MeV
100 MeV Electron 6 MeV
Cyclotron Linac Van de Graafa
(Ci) (Ci) (Ci)
2.0E-3
4.0E-2
1.0
1
l.OE-9
l.OE-4
aTritium target used for neutron generation; release estimates
include emissions from laboratory hoods due to tritium target
handling operations.
-------
5.2-10
Table 5.2-5. Typical accelerator facility
Parameter Value
Type of accelerator: 6 MeV Van de Graaff with
tritium targetoperating
3000 h/y
18 MeV electron linac
operating 2000 h/y
100 MeV research cyclotron
operating 1000 h/y
Emissions control: None
Stack characteristics:
Height 16.8 meters (roof type)
5.2.5 Health Impact Assessment of Typical Accelerators
The diverse operational modes practiced by facilities using
accelerators make it difficult to predict airborne effluent
compositions over any extended time period. Even at medical
accelerators, routine changes in radionuclide production and target
characteristics make it difficult to predict realistic source term
values. However, table 5.2-6 represents estimates of the annual
radiation doses resulting from radioactive emissions from typical
accelerators. These estimates are for a site in the suburbs of a
large Midwestern city (Site B, Appendix A). The nearest resident
was assumed to live 1 kilometer from the site.
Individual fatal cancer risks and committed fatal cancers to
the population within 80 km are estimated in table 5.2-7.
-------
5.2-11
Table 5.2-6. Annual radiation doses due to radioactive
emissions from typical accelerators
Type of
accelerator
6 MeV
Van de Graaf
18 MeV
Electron linac
100 MeV
Research cyclotron
Maximum
individual
(mrem/y)
1.1E-4
4.2E-8
9.6E-5
Average
individual
(mrem/y)
2.4E-07
1.3E-10
2.1E-09
Population
(person-rem/y)
5.9E-4
3.1E-7
5.1E-6
Table 5.2-7. Individual lifetime risks and number of fatal cancers
due to radioactive emissions from typical accelerators
Individual lifetime risks Expected fatal cancers
Type of Maximum Average per year of operations
accelerator individual individual (Fatal cancers)
6 MeV
Van de Graaf
18 MeV
Electron linac
1.6E-09
5.9E-13
3.4E-12
1.8E-15
1.2E-07
6.4E-11
100 MeV
Research
Cyclotron 1.3E-09 2.9E-14 l.OE-09
-------
5.2-12
REFERENCES
Te79 Teknekron, 1979, Final Report. A Study of Radioactive
Airborne Effluents from Particle Accelerators, EPA Contract No. 68-
01-4997, McLean, Va.
Ei72 Eicholz G. G. (Ed.), 1972, Radioisotope Engineering, Marcel
Dekker, Inc., New York, N. Y.
BRH75 Bureau of Radiological Health, 1975, The Use of Electron
Linear Accelerators in Medical Radiation Therapy, Overview Report
Number 2, Market and Use Characteristics: Current Status and
Future Terms, PB-246-226, U. S. Department of Commerce, Spring-
field, Va.
BRH78 Bureau of Radiological Health, 1978, Report of State and
Local Radiological Health Programs, Fiscal Year 1977. HEW Pub.
No. 78-8034, FDA, Department of Health, Education and Welfare,
Rockville, Md. 20852.
PHS68 Public Health Service, 1968, Particle Accelerator Safety
Manual, Report by William M. Brobeck and Associates under
Contract PH 86-67-193, MORP 68-12, Department of Health, Educa-
tion, and Welfare, Rockville, Md. 20852.
S179 Slaback L. A., 1979, Health Physics Aspects of 20- to 40-
MeV Linac Operation, Armed Forces Radiobiological Research
Institute, Bethesda, Md. (To be published)
Ka67 Kase K.R., 1967, Radioactive Gas Production at a 100 MeV
Electron Linac Facility, Health Physics, Vol. 13, 1967.
-------
APPENDICES
-------
APPENDIX A
ASSESSMENT METHODOLOGY
A.I Introduction
The general methodology used in the generic assessments
presented in this report consisted of the following parts:
1) a description of a model or typical facility for the
source category,
2) a choice of one or more generic sites appropriate to the
source category,
3) an assignment of a source term (Ci/y) and source related
quantities (e.g., release height, plume rise),
4) a calculation of the individual, average, and collective
doses due to air immersion, ground surface exposure, inhalation, and
ingestion of radionucTides,
5) a health risk assessment based on the doses to the various
organs and the working-level exposure from radon-222.
Assumptions made at each step were based on being realistic
but not underestimating the impact of a release. The following
sections describe these steps in more detail. (See Appendix B for
health risk assessment details.)
A.2 Model/Typical Facility
For each source category, a representative facility was
designated. In some instances (e.g., nuclear power plants),
extensive information was available on release rates and source
considerations influencing dispersion (e.g., release height and exit
velocity). In such cases, a model facility was designed to
represent an average facility for the source category. For other
source categories (e.g., radiopharmaceutical industry), industry
wide information was sparse. In these cases, data for a particular
facility considered representative of the source category was used
for the assessment and the facility was identified as a typical
facility.
A.3 Generic Sites
Six generic sites were characterized for the purpose of
assessing different source categories. These sites were chosen by
-------
A-2
first identifying locations of facilities within each source
category and then identifying a few of them which typified the types
of locations where such facilities might be located. Factors which
entered into this judgment included geographic location, population
density, and food crop production.
On the basis of similarities between representative sites for
the different source categories, six generic sites (designated A, B,
C, D, E, and F) were chosen which were believed to adequately
represent potential sites for all of the source categories
considered. For some source categories, one site was sufficient
(e.g., uranium mining) while others required as many as four sites
to represent the source category (e.g. fossil fuel power plants).
While the data used to characterize the generic sites were obtained
for specific locations, there would not necessarily be a facility at
that location for any specific source category.
Sites A and B represent urban locations. Site A
characterizes a very large metropolitan city: the maximum case with
respect to population density and overall population within 80 km
(New York City, New York). Site B represents the near suburbs of a
large Midwest city (St. Louis, Missouri). Site C was selected to
depict the phosphate industry since this location has a heavy
concentration of phosphate mining and milling (Polk County, Florida,
near Bartow). Site D represents a rural setting in the central
portion of the United States (near Little Rock, Arkansas). Site E
exhibits the characteristics associated with the uranium industry
and other mining endeavors (Grants, New Mexico). Site F is a
remote, sparsely populated location in the Northwest which
represents a minimal impact on the general population (near
Billings, Montana). Sites C through F are considered rural
locations. Table A-l gives the important characteristics of these
generic sites.
A.4 Source Characterization
Sources were characterized by the release rate (Ci/year) of
each emitted radionuclide. Radionuclides released as particulates
or reactive vapors were assigned a deposition velocity V,j of 1
cm/s unless otherwise indicated and a precipitation scavenging
factor of approximately 1E-5 C s'1 where C is the average
precipitation rate in m/year. See table A-3 for the actual values
used for the generic sites. An effective release height was
assigned to each source based on the release height and any expected
plume rise. In general, no credit was given for plume rise unless
it was clearly indicated. Because the depletion calculation method
in AIRDOS-II required very long run times for low level releases,
ground level releases were given an effective release height of 10 m.
-------
A-3
Table A-l. Characteristics of the generic sites
Site ANew York
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Mixing Height:
Population (0-8 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Crop Area:
(0-80 km)
New York/LaGuardia (WBAN=14732)
A-F
65/01-70/12
102 cm
1000 m
9.23E+5 persons
1.71E+7 persons
1.72E+5 head
1.17E+5 head
3.77E+4 ha
Site BMissouri
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Mixing Height:
Population: (0-8 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Food Crop Area:
(0-80 km)
St. Louis/Lambert (WBAN=13994)
A-6
60/01-64/12
102 cm
600 m
1.34E+4 persons
2.49E+6 persons
3.80F+4 head
6.90E+5 head
1.64E+4 ha
-------
A-4
Table A-l. Characteristics of the generic sitescontinued
Site CFlorida
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Mixing Height:
Population: (0-8 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Crop Area:
(0-80 km)
Orlando/Jet Port (WBAN=12815)
A-E
74/01-74/12
142 cm
1000 m
6.67E+3 persons
1.51E+6 persons
2.76E+4 head
2.57E+5 head
1.39E+4 ha
Site D--Arkansas
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Mixing Height:
Population: (0-8 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Crop Area:
(0-80 km)
Little Rock/Adams (WBAN=13963)
A-F
72/02-73/02
127 cm
600 m
1.18E+4 persons
5.92E+5 persons
1.19E+4 head
2.57E+5 head
2.94E+3 ha
-------
A-5
Table A-l. Characteristics of the generic sites--continued
Site E--New Mexico
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Mixing Height:
Population: (0-8 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Crop Area:
(0-80 km)
Grants/Gnt-Milan (WBAN=93057)
A-F
54/01-54/12
20 cm
800 m
0 persons
3.60E+4 persons
2.30E+3 head
8.31E+4 head
2.78E+3 ha
Site F--Montana
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Mixing Height:
Population: (0-8 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Crop Area:
(0-80 km)
Billings/Logan
A-F
67/01-71/12
20 cm
700 m
0 persons
1.19E+4 persons
1.86E+3 head
1.47E+5 head
1.77E+4 ha
(WBAN=24033)
-------
A-6
A.5 Environmental Pathway Modeling Computer Program
AIRDOS-II (Mo77) was used to calculate the individual and
collective doses and working level exposures for these assessments.
The program was modified to assess area sources using the
methodology described by Culkowski and Patterson (Cu76). Working
levels associated with radon-222 were calculated on the assumption
of a 70 percent equilibrium, a value considered representative of
indoor exposure conditions (Ge78).
Air concentrations are ground level sector averages.
Dispersion is calculated from annual average meteorological data.
Depletion due to dry deposition and precipitation scavenging was
calculated for particulates and reactive vapors. AIRDOS-II provides
no means for calculating resuspended air concentrations or
subsequent redeposition to the ground surface. The air
concentrations are the basis for inhalation and submersion dose
calculations.
Ground surface concentrations are calculated on the basis of
the deposition rate due to dry deposition and precipitation
scavenging. A 100-year accumulation period was generally used
unless otherwise indicated. Radiological decay was the only process
considered to deplete the ground surface concentration. Effective
decay constants were used to calculate the concentrations at the end
of the accumulation period for nuclides which are members of decay
chains.
Ingestion rates were calculated using the TERMOD portion of
AIRDOS-II. Initial assessments indicated that transfers from the
soil pool were often unrealistically high, especially for the
long-lived radionuclides. This result was due in part to the
equilibrium (rather than 100-year) values calculated by TERMOD and
in part to the TERMOD element dependent parameters (Ki76). Since no
satisfactory resolution of these difficulties was practicable, the
decision was made to consider only those transfers associated with
foliar deposition on food and forage. For selected nuclides, the
calculated ingestion intake rates are a low estimate of the values
after 100 years' accumulation. However, this approach is not
expected to significantly underestimate the overall risk from the
source categories considered and avoids some extreme overestimates
that would otherwise have been made.
-------
A-7
A.6 Individual Dose Assessment
The maximum individual was assessed on the following basis:
1) The maximum individual for each source category is
intended to represent an average of the individuals living near each
facility within the source category. The individual was assumed to
be located approximately 500 meters from the point of release in the
predominant wind direction. For area sources this location was
nominally 500 meters from the edge of the source.
2) The organ dose-equivalent rates in the tables are based on
the calculated environmental concentrations. For inhaled or
ingested radionuclides, the dose-equivalent rates are actually the
50-year committed dose-equivalent rates, i.e., the internal
dose-equivalents which would be delivered up to 50 years after an
intake. The individual dose equivalent rates in the tables are in
units of mrem/y.
3) Since the risk assessment is based on an entire lifetime
spent at the calculated environmental concentrations, an adult model
was appropriate for dosimetry.
4) The individual is assumed to home-grow a portion of his or
her diet consistent with the type of site. Individuals living in
urban areas were assumed to consume much less home produced food
than an individual living in a rural area. The fractions of home
produced food consumed by individuals for the generic sites are
shown in table A-2. For Site B the portion of the the individual's
diet that was not locally produced was assumed to come from the
average of the assessment area. Subsequent trial runs showed
little difference between assuming that the balance of the maximum
individual's diet comes from the assessment area or that it is
imported from outside the assessment area. Some assessments for
Site B were performed using the rural food source fractions for
Sites C-F in table A-2. These are identified in the text as Site B
with rural food source fractions.
A.7 Average Individual
Dose rates and working level exposures for an average
individual within 80 km of a source were obtained by dividing the
collective dose rates and working level exposures for the region
(see A.8) by the population of the region.
-------
A-8
Table A-2. Sources of food for the maximum individual
Food
Site
A
Site
B
Sites
C-F
Fl F2 F3 Fl F2 F3 Fl F2 F3
Vegetables
Meat
Milk
7.6
.8
0.
0.
0.
0.
92.
99.
100.
4
2
7.6
.8
0.
92.4
99.2
100.
0.
0.
0.
70
44
39
.0
.2
.9
30.0
55.8
60.1
0.
0.
0.
Fl and F2 are the percentages produced at the individual's
location and within the 80 km assessment area, respectively. The
balance of the diet, F3, is considered to be imported from outside
the assessment area with negligible radionuclide concentrations due
to the assessed source.
A.8 Collective Dose Assessment
The collective dose assessment to the population within an 80
km radius of the facility under consideration was performed as
follows:
1) The population distribution around the generic site was
based on the 1970 census. The population was assumed to remain
constant in time.
2) Average agricultural production data for the State in
which the generic site is located were assumed for all distances
greater than 400 meters from the source. For distances less than
400 meters no agricultural production takes place.
3) The population in the assessment area consumes food from
the assessment area to the extent that the calculated production
allows. Any additional food required is assumed to be imported
without contamination by the assessment source. Any surplus is not
considered in the assessment.
4) The collective organ dose-equivalent rates are based on
the calculated environmental concentrations. Fifty-year dose
commitment factors (as for the individual case) are used for
-------
A-9
ingestion and inhalation. The collective dose equivalent rates in
the tables can be considered to be either the dose commitment rates
after 100 years of plant operation or equivalently the doses which
will become committed for up to 100 years from the time of release
from one year of plant operation.
A.9 AIRDOS-II Parameters and Input Data
A sample computer input for the AIRDOS-II code as used in
these assessments (table A-4) is annotated to enable the reader to
understand the input format. Moore (Mo77) supplies a detailed
explanation of the code.
Mixing Height and Scavenging
Table A-3 summarizes the mixing heights, rainfall rates, and
scavenging coefficients used for the generic sites. A dry
deposition velocity of .01 m/s was used for particulates and
reactive vapors (e.g., elemental iodine) unless otherwise indicated.
Table A-3. Some site descriptors used with AIRDOS-II
Average mixing
Generic
site
Site A
Site B
Site C
Site D
Site E
Site F
height
(m)
1000
600
1000
600
800
700
Rainfall
rate
(cm/y)
102
102
142
127
20
20
Scavenging
coefficient
(s-i)
l.OE-5
l.OE-5
l.OE-5
l.OE-5
2.0E-6
2.0E-6
The average mixing height is the distance between the ground
surface and a stable layer of air where no further mixing occurs.
This average was computed by determining the harmonic mean of the
morning mixing height and the afternoon mixing height for the
location (Ho72). The rainfall rate (USGS70) determines the value
used for the scavenging coefficient. No attempt was made to be more
accurate
-------
A-10
than one significant figure for both average mixing height and
scavenging coefficient. Sites E and F are relatively dry locations
as reflected by the scavenging coefficients.
Meteorological Data
To demonstrate the methodology of converting from a standard
format for meteorological data (STAR format) to the AIRDOS-II
format, tables A-5 and A-6 are included. Site B meteorological data
are presented as an example in table A-5. A utility Fortran IV
program (Mo78) converts the STAR data to meet the input requirements
of the AIRDOS-II code (see lines 1200-4200 of table A-4).
STAR (an acronym for Stability ARray) meteorological data
summaries were obtained from the National Climatic Center,
Asheville, North Carolina. Data for the station considered most
representative for each generic site were used. Generally, these
data are from a nearby airport. The station used is identified by
the corresponding WBAN number in table A-l.
Dairy and Beef Cattle
Dairy and beef cattle distributions are part of the AIRDOS-II
input. A constant cattle density is assumed except for the area
closest to the source or stack in the case of a point source, i.e.,
no cattle wthin 400 m of the source. The cattle densities are
provided by State in table A-7. These densities were derived from
data developed by NRC (NRC75). Milk production density in units of
liters/day-square mile was converted to number of dairy cattle /
square kilometer by assuming a milk production rate of 11.0
liters/day per dairy cow. Meat production density in units of
kilograms/day-square mile was changed to an equivalent number of
beef cattle/square kilometer by assuming a slaughter rate of .00381
day-1 and 200 kilograms of beef/animal slaughtered. A 180-day
grazing period was assumed for dairy and beef cattle.
Population
The population data for each generic site were generated by a
computer program (At74) which utilizes an edited and compressed
version of the 1970 United States Census Bureau's "Master
Enumeration District List with Coordinates" containing housing and
population counts for each census enumeration district (CED) and the
geographic coordinates of the population centroid for the district.
In the Standard Metropolitan Statistical Areas the CED is usually a
"block group" which consists of a physical city block. In other
-------
A-ll
areas the district used is called the "enumeration district," and it
may cover several square miles in a rural area.
There are approximately 250,000 CEDs in the United States and
the average population is about 800. The position of the population
centroid for each CED was marked on the district maps by the
individual census official responsible for each district and is
based only on his judgment from inspection of the population
distribution on the map. The CED entries are sorted in ascending
order by longitude on the final data tape.
The resolution of a calculated population distribution cannot
be better than the distribution of the CEDs. In a metropolitan area
the resolution is often as small as one block, but in rural areas it
may be on the order of a mile or more.
Vegetable Crop Area
A certain fraction of the land within 80 km of the source is
used for vegetable crop production which is assumed to be uniformly
distributed throughout the entire assessment area with the exception
of the first 400 meters from the source. Information on the
vegetable production density in terms of kilograms (fresh weight)/
day-square mile were obtained from NRC data (NRC75). The vegetable
crop fractions (table A-7) by State were obtained from the
production densities by assuming a production rate of 2 kilograms
(fresh weight)/year-square meter (NRC77).
Food Intake
Referring to the sample input table (table A-4), lines
21800-21900 determine what percentage of milk, beef and vegetables
is produced and consumed at each environmental location; is produced
throughout the assessment area (average value for the whole
assessment area) and consumed locally; and is produced outside the
assessment area (imported) and consumed locally. The imported
fraction is assumed to have no radioactivity content.
Table A-2 summarizes the ingestion values used for each
generic site for the maximum individual. These values are based on
a USDA report (USDA72). Sites A and B utilize data on urban
locations while rural sites are based on rural-farm situations. The
Fl ratios are obtained by dividing the home-produced quantity by the
quantity from all sources. The beef ratios are actually meat
values, i.e., beef and pork. The vegetable ratios only include
fresh vegetables.
-------
A-12
For population exposure estimates, the AIRDOS-II code
determines the imported fraction needed to supply the nutritional
requirements of the entire population within 80 km. The quantity of
food that is not imported is assumed to be grown or produced
throughout the entire assessment area and consumed by the population
within the assessment area as an average value for the entire
assessment area. For a site that produces more food than is needed
for the population within the assessment area, this food is assumed
to be exported outside the assessment area. No collective doses
were calculated for such exported foods.
The ingest ion pathway is calculated by the TERMOD portion
(Ki76) of AIRDOS-II. The input values shown in table A-8 were used
and are independent of location. The soil transfer parameters,
TAURG and TAUSP, were set to zero since only contamination from
foliar deposition was considered.
Internal Doses
Internal doses to each organ were calculated for inhalation
and ingestion using dose conversion factors from table A-9. Sources
for the dose conversion factors in table A-9 are listed in table
A-10. The choice of organs to be considered was dictated by
AIRDOS-II. The dose to the bone is an average dose to hard bone,
not bone marrow. The G.I. tract dose is the dose delivered to the
lower large intestine. The total body dose is generally a mass
weighted mean of the dose to a number of specific organs. Where
dose conversion factors for particular organs are not shown in
table A-9, the total-body factor is used as a default value.
Doses to other soft tissue used for risk assessments (see
Appendix B) were estimated by the muscle dose values. In those
cases where the muscle dose conversion factor is actually a total
body dose conversion factor which included bone as one of the
specific organs, the muscle dose may be overestimated,.
Inhalation dose conversion factors for uranium-235 and
members of the uranium and thorium series were calculated on the
basis of the Task Group Lung Model (TGLD66, ICRP72) for clearance
class Y and an effective particle size (AMAD) of 1.0 micrometer.
Inhalation dose conversion factors for strontium-90 and
technetium-99 were similarly calculated but for clearance classes D
and W respectively. Details of the calculational models and
assumptions for other nuclides may be found in the references in
table A-10.
-------
A-13
Working level exposures associated with radon-222 were
calculated assuming an indoor exposure at 70 percent equilibrium
(6e78) (i.e., 100 pCi/L radon-222 = 0.7 working level). Accord-
ingly, no inhalation doses were calculated for radon-222 or its
short-lived progeny.
TERMOD Input Data
Element dependent input parameters for the TERMOD
calculations were taken from Killough (Ki76, Table 2-7). The value
of fm for radium was corrected from 1.5E+2 to 1.5E-2.
External Doses
External doses to each organ were calculated for air
submersion and ground surface exposure using the appropriate dose
conversion factors from table A-9. Air submersion doses are for a
semi-infinte exposure to the ground level air concentration. Ground
surface exposures are for an infinite plane with the ground surface
concentration at the particular location. Noble gases (e.g., xenon
and krypton) were assumed to be nondepositing. Therefore, even
through table A-9 indicates surface DCFs for some noble gases there
were no surface doses from these radionuclides.
-------
A-14
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00 0s! O *~* CN] CO -3" *O ^-D I1"*- OO (^ O *-^ cvj <"") ^" in \O r~^ CO O"»
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CsICNjCSCMCSCSlCNCslCSC^CsICs|CslCvl
-------
A-21
Table A-5. Site B meteorological data (STAR format)
WSl
WS2
WS3
WS4
WS5
WS6
STATION PERIOD SEQUENCE
NUMBER (2) OF RECORD NUMBER
N
NNE
NE
ENE
E
ESE
SE
SSE
S
ssw
sw
wsw
w
WNW
NW
NNW
N
NNE
NE
ENE
E
ESE
SE
SSE
S
SSW
sw
wsw
w
WNW
NW
NNW
N
NNE
NE
ENE
E
ESE
SE
SSE
S
SSW
SW
WSW
W
WNW
NW
NNW
A
A
A
A
A
A
A
A
A
A
A
A
A
A
A
A
B
B
B
B
B
B
B
B
B
B
B
B
B
B
B
B
C
C
C
C
C
C
C
C
C
C
C
C
C
C
C
C
o.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
o.
0.
0.
o.
o.
o.
0.
o.
o.
0.
o.
0.
0.
0.
o.
o.
o.
o.
0.
0.
0.
0.
0.
o.
0.
o.
o.
0.
o.
o.
o.
000220.
000140.
000080.
000120.
000160.
000090.
000180.
000050.
000130.
000070.
000150.
000100.
000200.
000170.
000080.
000110.
000540.
000670.
000590.
000740.
000610.
001140.
000700.
000330.
000800.
000530.
000560.
000670.
001150.
000610.
000740.
000610.
000250.
000190.
000380.
000390.
000240.
000570.
000530.
000350.
000430.
000170.
000350.
000420.
000720.
000520.
000280.
000260.
000320.
000160.
000160.
000180.
000250.
000180.
000250.
000250.
000390.
000230.
000340.
000250.
000340.
000340.
000250.
000300.
001260.
001370.
001760.
001740.
001260.
001580.
001550.
001350.
001870.
001510.
001960.
002190.
001640.
001960.
001370.
001160.
001300.
001390.
001370.
001190.
001460.
002210.
002010.
001620.
001990.
000820.
002150.
002170.
002240.
001920.
001370.
001120.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
000000.
oooooo.
000000.
000730.
000890.
000660.
000780.
000870.
001100.
000840.
001070.
001070.
000890.
001710.
001280.
001070.
000960.
001100.
000820.
002420.
002210.
001870.
002650.
003220.
003790.
003860.
005090-
007100.
004860.
005000.
006160.
004820.
004910.
003740.
003010.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
oooooo.
000550.
000320.
000550.
000320.
000300.
000530.
000620.
001030.
001670.
000870.
001140.
001350.
000910.
000980.
000960.
000570.
OOOOOO
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
oooooo
000020
oooooo
oooooo
oooooo
000020
000090
000210
000070
000070
oooooo
000070
000140
000020
000020
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00002
.00002
.00002
13994 1A1760
13994 2A1760
13994 3A1760
13994 4A1760
13994 5A1760
13994 6A1760
13994 7A1760
13994 8A1760
13994 9A1760
1399410A1760
1399411A1760
1399412A1760
1399413A1760
1399414A1760
1399415A1760
1399416A1760
13994 1B1760
13994 2B1760
13994 3B1760
13994 4B1760
13994 5B1760
13994 6B1760
13994 7B1760
13994 8B1760
13994 9B1760
1399410B1760
1399411B1760
1399412B1760
1399413B1760
1399414B1760
1399415B1760
1399416B1760
13994 1C1760
13994 2C1760
13994 3C1760
13994 4C1760
13994 5C1760
13994 6C1760
13994 7C1760
13994 8C1760
13994 9C1760
1399410C1760
1399411C1760
1399412C1760
1399413C1760
1399414C1760
1399415C1760
1399416C1760
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
-------
A-22
Table A-5. Site B meteorological data (STAR format)--continued
(D(2) WSl
WS2
WS3
WS4 WS5
WS6
STATION PERIOD SEQUENCE
NUMBER (2) OF RECORD NUMBER
N D 0.000640.003080.009160.010530.001070.00002
NNE D 0.000710.003470.007510.006710.000460.00007
NE D 0.001060.003880.006830.003900.000230.00005
ENE D 0.000900.003970.006670.003700.000210.00000
E D 0.000950.003720.009680.005270.000460.00005
ESE D 0.001020.004290.012560.009410.000870.00007
SE D 0.000850.003630.014750.014230.001740.00011
SSE D 0.000580.003240.015210.023130.003720.00034
S D 0.000830.003700.020890.032150.004340.00039
SSW D 0.000360.002330.010460.011620.001580.00009
SW D 0.000480.002900.009640.008810.000870.00014
WSW D 0.000600.003220.008010.011670.001440.00041
W D 0.000830.004040.009360.012950.003290.00119
WNW D 0.000880.003770.012060.030030.012860.00244
NW D 0.000780.004110.012810.026490.006210.00112
NNW D 0.000460.002630.010640.012970.001940.00023
N E 0.000000.001440.002810.000000.000000.00000
NNE E 0.000000.002190.001940.000000.000000.00000
NE E 0.000000.003240.001370.000000.000000.00000
ENE E 0.000000.002510.001620.000000.000000.00000
E E 0.000000.002580.002030.000000.000000.00000
ESE E 0.000000.004110.004660.000000.000000.00000
SE E 0.000000.005180.005800.000000.000000.00000
SSE E 0.000000.003400.008930.000000.000000.00000
S E 0.000000.003880.015140.000000-000000.00000
SSW E 0.000000.002470.006900.000000.000000.00000
SW E 0.000000.003010.006350.000000.000000.00000
WSW E 0.000000.002950.006740.000000.000000.00000
W E 0-000000.004060.008220.000000.000000.00000
WNW E 0.000000.003240.009060.000000.000000.00000
NW E 0.000000.001990.004610.000000.000000.00000
NNW E 0.000000.001960.003970.000000.000000.00000
N F 0-000530.002720.000000.000000.000000.00000
NNE F 0.000710.003770.000000.000000.000000.00000
NE F 0.001490.004610.000000.000000.000000.00000
ENE F 0.001050.003010.000000.000000.000000.00000
E F 0.001580.003240.000000.000000.000000.00000
ESE F 0.001600.006280.000000.000000.000000.00000
SE F 0.002570.010250.000000.000000.000000.00000
SSE F 0.001560.007670.000000.000000.000000.00000
S F 0.001550.008220.000000.000000.000000.00000
SSW F 0.000690.004750.000000.000000.000000.00000
SW F 0.000810.005640.000000.000000.000000.00000
WSW F 0.001140.006120.000000.000000.000000.00000
W F 0.002180.011190.000000.000000.000000.00000
WNW F 0.001240.007100.000000.000000.000000.00000
NW F 0.000910.003490.000000.000000.000000.00000
NNW F 0.000650.002650.000000.000000.000000.00000
13994 1D1760 16412
13994 2D1760 16412
13994 3D1760 16412
13994 4D1760 16412
13994 5D1760 16412
13994 6D1760 16412
13994 7D1760 16412
13994 8D1760 16412
13994 9D1760 16412
1399410D1760 16412
1399411D1760 16412
1399412D1760 16412
1399413D1760 16412
1399414D1760 16412
1399415D1760 16412
1399416D1760 16412
13994 1E1760 16412
13994 2E1760 16412
13994 3E1760 16412
13994 4E1760 16412
13994 5E1760 16412
13994 6E1760 16412
13994 7E1760 16412
13994 8E1760 16412
13994 9E1760 16412
1399410E1760 16412
1399411E1760 16412
1399412E1760 16412
1399413E1760 16412
1399414E1760 16412
1399415E1760 16412
1399416E1760 16412
13994 1F1760 16412
13994 2F1760 16412
13994 3F1760 16412
13994 4F1760 16412
13994 5F1760 16412
13994 6F1760 16412
13994 7F1760 16412
13994 8F1760 16412
13994 9F1760 16412
1399410F1760 16412
1399411F1760 16412
1399412F1760 16412
1399413F1760 16412
1399414F1760 16412
1399415F1760 16412
1399416F1760 16412
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
-------
A-23
Table A-5. Site B meteorological data (STAR format)--continued
(1)(2) WSl WS2 WS3 WS4
WS5
WS6
STATION PERIOD SEQUENCE
NUMBER (2) OF RECORD NUMBER
N
NNE
NE
ENE
E
ESE
SE
SSE
S
SSW
SW
WSW
W
WNW
NW
NNW
G
G
G
G
G
G
G
G
G
G
G
G
G
G
G
G
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
.001770.
.003170.
.004710.
.003580.
.004030.
.005210.
.008970.
.005350.
.003490.
.001500.
.002040.
.004080.
.008110.
.005890.
.002490.
.001400.
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
.000000.
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.000000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
.00000
13994 1G1760
13994 2G1760
13994 3G1760
13994 4G1760
13994 5G1760
13994 6G1760
13994 7G1760
13994 8G1760
13994 9G1760
1399410G1760
1399411G1760
1399412G1760
1399413G1760
1399414G1760
1399415G1760
1399416G1760
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
16412
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
0603
(1) - Wind direction (from indicated direction)
(2) - Stability class (A through G)
WSl - Wind speed class 0-3 knots
WS2 - Wind speed class 4-6 knots
WS3 - Wind* speed class 7-10 knots
WS4 - Wind speed class 11-16 knots
WS5 - Wind speed class 17 - 21 knots
WS6 - Wind speed class 21+ knots
Period of record: January 1960 - December 1964
-------
A-24
Table A-6. Listing of STAR code
100
200
300
400
500
600
700
800
900
1000
1100
1200
1300
1400
1500
1600
1700
1800
1900
2000
2100
2200
2300
2400
2500
2600
2700
2800
2900
3000
3100
3200
3300
3400
3500
3600
3700
3800
3900
4000
4100
4200
4300
4400
4500
4600
4700
4800
4900
5000
5100
: UTILITY PROGRAM TO CONVERT STAR FORMAT METEOROLOGICAL DATA TO
} AIRDOS-II INPUT FORMAT. MODIFIED BY C. B. NELSON 12/15/78.
DIMENSION WDCS(16,8,6),WS(6),RWS(6),DICAT(16,8),TWDC(16,8),
1RWDC(16,8),PERD(16)
DATA WDCS/768*0./
DO 1 IC=1,8
DO 1 IS=1,16
ID=MOD(25-IS,16)+1
1 READ(50,200,END=2)(WDCS(ID,IC,IW),IW=1,6)
200 FORMAT(T8,6F7.5)
2 CONTINUE
WS(1)=1.5*.5148
WS(2)=5.*.5148
WS(3)=8.5*.5148
WS(4)=13.5*.5148
WS(5)=19.*.5148
WS(6)=23.*.5148
DO 4 1=1,6
4 RWS(I)=1./WS(I)
DO 5 ID=1,16
DO 6 IC=1,8
DICAT(ID,IC)=0
DO 7 IW=1,6
7 DICAT(ID,IC)=WDCS(ID,IC,IW)+DICAT(ID,IC)
6 CONTINUE
5 CONTINUE
DO 8 ID=1,16
DO 9 IC=1,8
SUM1=0
SUM2=0
SUM3=0
DO 10 IW=1,6
SUM1=SUM1+WDCS(ID,IC,IW)*WS(IW)
SUM2=SUM2+WDCS(ID,IC,IW)*RWS(IW)
10 SUM3=SUM3+WDCS(ID,IC,IW)
IF(SUM3.EQ.O)TWDC(ID,IC)=0
IF(SUM3.EQ.O)RWDC(ID,IC)=0
IF(SUM3.EQ.O)GO TO 80
TWDC(ID,IC)=SUM1/SUM3
RWDC(ID,IC)=SUM2/SUM3
80 CONTINUE
9 CONTINUE
8 CONTINUE
DO 40 ID=1,16
DO 41 IC=1,7
IF(RWDC(ID,IC).EQ.O)GO TO 81
RWDC(ID,IC)=1./RWDC(ID,IC)
81 CONTINUE
41 CONTINUE
40 CONTINUE
-------
A-25
Table A-6. Listing of STAR codecontinued
5200 DO 50 ID=1,16
5300 PERD(ID)=0
5400 DO 51 IC=1,7
5500 51 PERD(ID)=PERD(ID)+DICAT(ID,IC)
5600 DO 52 IC=1,7
5700 52 DICAT(ID,IC)=DICAT(ID,IC)/PERD(ID)
5800 50 CONTINUE
5900 SUM=0
6000 DO 70 ID=1,16
6100 70 SUM=SUM+PERD(ID)
6200 DO 71 ID=1,16
6300 71 PERD(ID)=PERD(ID)/SUM
6400 WRITE(51,300)(PERD(I),I=1,16)
6500 300 FORMAT(' ',T20,16F5.3)
6600 WRITE(51,302)((RWDC(ID,IC),ID=1,16),IC=1,7)
6700 302 FORMAT(' ',T20,16F5.2)
6800 WRITE(51,302)((TWDC(ID,IC),ID=1,16),IC=1,7)
6900 WRITE(51,301)((DICAT(ID,IC),IC=1,7),ID=1,16)
7000 301 FORMAT(' ',T20,7F10.4)
7100 PUNCH 400,(PERD(I),1=1,16)
7200 400 FORMAT(16F5.3)
7300 PUNCH 402,((RWDC(ID,IC),ID=1,16),IC=1,7)
7400 402 FORMAT(16F5.2)
7500 PUNCH 402,((TWDC(ID,IC),ID=1,16),IC=1,7)
7600 PUNCH 401,((DICAT(ID,IC),IC=1,7),ID=1,16)
7700 401 FORMAT(7F10.4)
7800 STOP
7900 END
-------
A-26
Table A-7. Cattle densities and vegetable crop
distributions for use with AIRDOS-II
State
Alabama
Arizona
Arkansas
California
Colorado
Connecticut
Delaware
Florida
Georgia
Idaho
1 1 1 i no i s
Indiana
Iowa
Kansas
Kentucky
Louisiana
Maine
Maryland
Massachusetts
Michigan
Minnesota
Mississippi
Missouri
Montana
Nebraska
Nevada
New Hampshire
New Jersey
New Mexico
New York
Dairy cattle
density
#/km2
7.02E-1
2.80E-1
5.90E-1
2.85
3.50E-1
2.50E-1
2.72
1.37
8.63E-1
8.56E-1
2.16
2.80
3.14
8.00E-1
2.57
9.62E-1
8.07E-1
6.11
3.13
3.51
4.88
8.70E-1
1.89
9.27E-2
8.78E-1
5.65E-2
1.58
3.29
1.14E-1
8.56
Beef cattle
density
#/km2
1.52E+1
3.73
1.27E+1
8.81
1.13E+1
3.60
6.48
1.28E+1
1.43E+1
7.19
3.33E+1
3.34E+1
7.40E+1
2.90E+1
2.65E+1
1.08E+1
7.65E-1
1.09E+1
2.90
7.90
1.85E+2
1.75E+1
3.43E+1
7.29
3.50E+1
1.84
1.40
4.25
4.13
5.83
Vegetable
crop fraction
km2/km2
4.16E-3
2.90E-3
1.46E-3
1.18E-2
1.39E-2
7.93E-3
5.85E-2
6.92E-3
2.17E-3
7.15E-2
2.80E-2
2.72E-2
2.43E-2
5.97E-2
3.98E-3
4.35E-2
5.97E-2
1.11E-2
4.96E-3
1.70E-2
3.05E-2
1.07E-3
8.14E-3
8.78E-3
2.39E-2
8.92E-3
6.69E-2
1.82E-2
1.38E-3
1.88E-2
-------
A-27
Table A-7. Cattle densities and vegetable crop
distributions for use with AIRDOS-1 Icontinued
State
North Carolina
North Dakota
Ohio
Oklahoma
Oregon
Pennsylvania
Rhode Island
South Carolina
South Dakota
Tennessee
Texas
Utah
Vermont
Virginia
Washington
West Virgina
Wisconsin
Wyoming
Dairy cattle
density
#/km2
1.26
6.25E-1
4.56
7.13E-1
4.53E-1
6.46
2.30
7.02E-1
8.85E-1
2.00E-1
5.30E-1
4.46E-1
8.88
1.84
1.50
6.00E-1
1.43E+1
5.79E-2
Beef cattle
density
#/km2
1.02E+1
1.18E+1
2.03E+1
2.68E+1
4.56
9.63
2.50
8.87
2.32E+1
2.11E+1
1.90E+1
2.84
4.71
1.31E+1
5.62
6.23
1.81E+1
5.12
Vegetable
crop fraction
km2/km2
6.32E-3
6.29E-2
1.70E-2
2.80E-2
1.59E-2
1.32E-2
4.54E-2
1.84E-3
1.20E-2
2.72E-3
5.77E-3
1.83E-3
1.08E-3
8.70E-3
5.20E-2
1.16E-3
1.78E-2
1.59E-3
-------
A-28
Table A-8. 'Site independent parameters used for AIRDOS-II
generic site assessments
Symbolic
variable
GRAZ
PTPMV
PTPMB
PTPMM
BRTHRT
T
A
ASUBG
DSUBF
DSUBG
SMALLD
KSUBB
MSUBB
RHO
SI
Description
Length of grazing season
Period of time from vegetable
production to human consumption
Period of time from beef production
to human consumption
Period of time from milk
production to human consumption
Human breathing rate
Buildup time for surface
deposition
Soil surface area to furnish
food crops for one person
Pasture area per cow
Dry weight area density of
above surface food
Dry weight area grass density
Depth of plow layer
Rate of increase of steer
muscle mass
Muscle mass of steer at slaughter
Soil density
Fraction of deposited
Value
180. days
0.0 days
0.0 days
0.0 days
9.58E+5 cm3/n
3.65E+4 days
l.OE+3 m2
l.OE+4 m2
l.OE-1 kg/m2
1.5E-1 kg/m?
2.0E+1 cm
4.0E-1 kg/day
2.0E+2 kg
1.4 gm/cm3
l.OE-1
radionuclides intercepted by
above-surface food crop
-------
A-29
Table A-8. Site independent parameters used for AIRDOS-II
generic site assessmentscontinued
Symbolic
variable
Description
Value
S2
S3
TAUBEF
TAUMLK
TAUBM
TAUCM
TAUES
TAUGR
TAUPD
TAURD
TAURG
TAUSP
U
V
Fraction of deposited
radionuclides on soil surface
below above-surface food crop
Fraction of deposited
radionuclides intercepted
by pasture grass
Fraction of beef herd
slaughtered per day
Number of mi Ik ings per cow
Human beef consumption rate
Human milk consumption rate
Above-surface food crop to
soil surface transfer rate
Pasture grass to pasture
soil transfer rate
Soil pool to soil sink
transfer rate
Pasture soil to soil sink
transfer rate
Pasture soil to pasture
grass transfer rate
Soil surface to soil pool
transfer rate
Milk capacity of udder
Human consumption rate
of vegetable food
9.0E-1
2.5E-1
3.81E-3 day-1
2.0 day-1
3.0E-1 kg/day
1.0 L/day
4.95E-2 day-1
4.95E-2 day-1
1.1E-4 day-1
1.1E-4 day-1
0.0 day-1
0.0 day-1
5.5 L
2.5E-1 kg/day
-------
A-30
Table A-8. Site independent parameters used for AIRDOS-II
generic site assessmentscontinued
Symbolic
variable
VSUBC
VSUBM
DD1
DD2
DD3
DD4
Description
Grass consumption rate
(dry) of cow
Milk production rate
of cow
Dietary correction factor
for above-surface food
Dietary correction factor
for uptake from soil
Dietary correction factor
for beef
Dietary correction factor
Value
10.
11.
1.0
1.0
1.0
1.0
kg/day
L/day
for milk
-------
A-31
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A-38
Table A-9, Dose conversion factors for use in preliminary
Clean Air Act assessments continued
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-------
A-39
Table A-10. Sources for dose conversion factors used in table A-9
Radionuclides
Inhalation
Ingest ion
Air submersion
and surface
exposure
Americium-241 aEPA77
Plutonium-242,241,240,238 Mo78
Plutonium-239 bEPA77
Uranium-238,235,234 Ho74
Uranium-236 ORNL78
Uranium-233 Mo78
Protactinium-234m,234 Mo78
Thorium-234,232,231,228 Ho74
Thorium-230 Ho74
Radium-228,224 Ho74
Radium-226 dHo74
Bismuth-214
Bismuth-212 Ho74
Lead-214
Lead-212 Ho74
Lead-210 Wa78a
Polonium-210 Wa78a
Thallium-208
Xenon-138,135m,135,133m,133 Mo78
Xenon-131m hWo78
Cesium-137,134 Mo78
Iodine-133,129 Mo78
Iodine-131,125 Ke76
Technetium-99m,99 Ki78
Strontium-90 Ki78
EPA77
Mo 78
CEPA77
Mo 78
ORNL78
Mo 78
Mo 78
Mo 78
Ki76
Mo78
eWa78b
Mo78
Mo78
fWa78a
9Wa78a
Mo 78
Mo 78
Mo 78
Ke76
Ki78
Ki78
Mo 78
Mo78
Mo 78
Mo78
Ki76
Mo78
Mo 78
Mo78
Ki76
Ki76
Ki76
Ki76
Mo 78
Ki76
Mo 78
Mo78
Mo78
Ki76
Mo 78
Ki76
Mo78
Mo 78
Ki76
Ki76
Ki76
-------
A-40
Table A-10. Sources for dose conversion factors used in
table A-9continued
Air submersion
and surface
RadionucTides Inhalation Ingestion exposure
Krypton-88,87,85m,85,83m Mo78 Mo78 Mo78
Cobalt-60,58 i|
-------
A-41
REFERENCES
At74 Athey T. W., R. A. Tell, and D. E. Janes, 1974, The Use of an
Automated Data Base in Population Exposure Calculations, from
Population Exposures, Health Physics Society, CONF-74018,
October 1974.
Cu76 Culkowski W. M. and M. R. Patterson, 1976, A Comprehensive
Atmospheric Transport and Diffusion Model, ORNL/NSF/EATC-17, Oak
Ridge National Laboratory.
E178 Ellett W. H. 1978, Private Communication with P. Magno.
EPA77 Environmental Protection Agency, 1977, Proposed Guidance on
Dose Limits for Persons Exposed to Transuranium Elements in the
General Environment, EPA 520/4-77-016, Washington D.C.,
September 1977.
Ge78 George A. C. and A. J. Breslin, 1978, The Distribution of
Ambient Radon and Radon Daughters in Residential Buildings in
the New Jersey-New York Area. Presented at Symposium on the
National Radiation Environment III, Houston, Texas.
Ho72 Holzworth G. C., 1972, Mixing Heights, Wind Speeds, and
Potential for Urban Air Pollution Throughout the Contiguous
United States, Report AP-101, U. S. Office of Air Programs
1972.
ICRP72 International Commission on Radiological Protection,
1972, The Metabolism of Compounds of Plutonium and other
Actinides, ICRP Publication No. 19, Pergamon Press, N.Y.
Ho74 Houston J. R., et al., 1974, DACRIN - A Computer Program for
Calculating Organ Dose from Acute or Chronic Radionuclide
inhalation, BNWL-B-389, December 1974.
Ke76 Kereiakes J. G., P. A. Feller, F. A. Ascoli, S. R. Thomas,
M. J. Gelfand and E. L. Saenger, 1976, Pediatric Radio-
pharmaceutical Dosimetry, Radiopharmaceutical Dosimetry Symposium,
Proceedings of Conference held at Oak Ridge, Tennessee,
April 26-29, 1976 (Superintendent of Documents, U.S. Government
Printing Office, Washington, DC).
Ki76 KiHough G. G. and L. R. McKay, 1976, A Methodology for
Calculating Radiation Doses from Radioactivity Released to the
Environment, ORNL-4992, March 1976.
-------
A-42
REFERENCEScontinued
Ki78 Killough G. G., et al., 1978, Estimate of Internal Dose
Equivalent to 22 Target Organs for Radionuclides Occurring in
Routine Releases from Nuclear Fuel Cycle Facilities. Vol. 1,
NUREG/CR-0150, ORNL/NUREG/TM190, June 1978.
Mo77 Moore R. E., 1977, The AIRDOS-II Computer Code for Estimating
Radiation Doses to Man from Airborne Radionuclides in Areas
Surrounding Nuclear Facilities, ORNL-5245, April 1977.
Mo78 Moore R. E., 1978, ORNL, private communication with
C. B. Nelson, EPA, 1978. Note - These dose conversion factors
were generally derived from Ki76.
NRC75 Memo from K. Eckerman, N. Dayem, R. Emch, Radiological
Assessment Branch, Division of Technical Review, Nuclear
Regulatory Commission, Code Input Data for Man-Rem Estimates,
(Washington, DC, October 15, 1975).
NRC77 Nuclear Regulatory Commission, 1977 Regulatory Guide 1.109,
Calculation of Annual Doses to Man from Routine Releases of Reactor
Effluents for the Purpose of Evaluating Compliance with 10 CFR Part
50, Appendix I, Revision 1, October 1977, Office of Standards
Development, NRC, Washington, D.C.
TGLD66 Task Group on Lung Dynamics, 1966, Deposition and Retention
Models for Internal Dosimetry of the Human Respiratory Tract,
Health Physics, Vol. 12, No. 2, pp. 173-207, February 1966.
USDA72 United States Department of Agriculture, 1972, Food
Consumption of Households in the United States (Seasons and Year
1965-1966), Household Food Consumption Survey 1965-1966, Report
No. 12, Agricultural Research Service, USDA, Washington, DC
(March 1972).
USGS70 U.S. Geological Survey, 1970, The National Atlas, U. S.
Department of the Interior, Washington, D.C.
Wa78a Watson A.P., 1978, Private communication from A.P. Watson,
ORNL, July 16, 1978DCFs were calculated using INREM-II.
Wa78b Watson A.P., 1978, Private communication from A.P. Watson,
ORNL, September 1978DCFs were calculateed using INREM-II.
-------
APPENDIX B
HEALTH RISK ASSESSMENT METHODOLOGY
The fatal cancer risks presented in this report were
estimated from the risk/rem and risk/WL-year factors shown in table
B-l. The risk/rem factors were developed from information in the
BEIR report (BEIR72). The risk/WL-year conversion factors were
developed primarily from information on uranium miner exposures
(EPA79 Section 4.0).
Risk estimates are limited to fatal cancers only. Our
current practice is to assume that for whole body exposure, the
number of genetic health effects and the number of nonfatal cancers
are each about the same as the number of fatal cancers (EPA77).
In applying these risk factors to the organ doses calculated
by the AIRDOS-II code, the following modifications were necessary:
(1) Since AIRDOS-II calculates only the dose to
bone (and not to bone marrow), a risk factor
of 3 x 10-5 fatal cancers per person-rem was
applied to the bone doses in order to take
into consideration the risk to both the bone
and red bone marrow. This factor is a
composite of the risk factors in table B-l of
4 x 10-5 f0r red bone marrow and 1 x 10"5
for bone (other organ) and was developed on
the basis that the average ratio of red bone
marrow dose to bone dose is about 0.5.
(2) A risk factor of 5 x 10-5 fatai cancers per
person-rem was applied to soft tissue doses
calculated by AIRDOS-II. This factor includes
the risk to breast of 4 x 10-5 anc) the risk
for one other soft tissue organ of 1 x 10-5.
(3) The kidney and liver were used as the
remaining two organs in the "all other"
category.
(4) Since stomach doses were not calculated by
AIRDOS-II, no corresponding risks were
estimated.
-------
B-2
(5) The total body risk factor was not used since
dose estimates for specific organs were
available.
The individual lifetime risks are the fatal cancers risks to
individuals which would result from an average lifetime exposure (70
years) to the dose rates and working levels estimated for those
individuals. The lifetime risk to the maximum individual was
obtained by multiplying the dose equivalent rates and working level
exposures by 70 to obtain the lifetime exposure and then multiplying
this value by the risk/rem or risk/WL-year factors shown in table
B-l.
The lifetime risk to the average individual was obtained by
dividing the population exposures in person-rem/year and
person-working levels by the total number of people in the exposed
population to obtain average annual exposure rates and then
proceeding as described above for the maximum individual.
The number of fatal cancers per year were obtained by
multiplying the annual collective dose equivalents and working level
exposures by the risk/rem or risk/WL-year factors in table B-l.
Table B-l. Risks of fatal cancer
Organ Risk of fatal cancer
(per person-rem) (per person-WL-y)
Total body 2E-4
Red bone marrow (leukemia) 4E-5
Lung 4E-5 a2.1E-2
Breast (average for both sexes) 4E-5
6. I. tract 2E-5
Stomach 2E-5
All others'3 (for each site) 1E-5
Thyroidc 1E_6
Exposures from radon-222 decay products.
"Up to four sites other than those listed above.
cNonfatal cancer risk of 1E-5.
-------
B-3
REFERENCES
BEIR72 Advisory Committee on the Biological Effects of Ionizing
Radiation, 1972, The Effects of Population Exposures to Low Levels
of Ionizing Radiation, National Academy of Sciences, Washington,
D.C.
EPA77 Environmental Protection Agency, 1977, Radiological Quality
of the Environment in the United States, EPA 520/1-77-009, Office
of Radiation Programs, Washington, D.C.
EPA79 Environmental Protection Agency, 1979, Indoor Radiation
Exposure Due to Radium-226 in Florida Phosphate Lands,
EPA-520/4-78-0013, Office of Radiation Programs, Washington, D.C.
-------
APPENDIX C
SOURCE TERM CALCULATIONS
C.I Metal and Nonmetal Mining
C.I.I Mines
Radon-222 emissions from mines were estimated either directly
from measurement data on concentrations of radon-222 in mine ventila-
tion air or in the absence of measurement data indirectly from the
uranium-238 content of the ore.
Radon-222 emissions for iron, zinc, and clay mines were
estimated from effluent measurement data as follows:
A = B c D (1)
where
A = radon-222 emissions, Ci/y
B = mine ventilation rate,
C = radon-222 concentration in the mine effluent, pCi/L
D = unit conversion factor = 1.49 x 10-5 L rcin Ci
ft3 y pCi
The values of the parameters A, B, and C are listed in table
C.l-1.
Radon-222 emissions from mining operations are comprised of
two components: radon-222 released during the actual mining of the
ore and radon-222 released from ore surfaces exposed during the
mining. Releases from copper, limestone, fluorspar, and bauxite
mines, estimated from the uranium-238 content of the ore, were
calculated as follows:
E = F G H I + Z (2)
where
E = radon-222 emissions, Ci/y
-------
C-2
F = amount of ore mined annually, short ton/y or ST/y
G = concentration of uranium-238 in ore, ppm
H = fraction of radon released from ore =0.2
-, Ci
I = unit conversion factor = 3.03 x
ppm
Z = annual release rate of radon resulting from the exposed
ore body
Radioactive equilibrium was assumed between uranium and its
daughters. The values of parameters E, F, G, Z are listed in table
C.l-2.
Table C.l-1. Values of parameters A, B, C
Type
of
mine
Iron
Zinc
Clay
A
Radon
emission
rate
(Ci/y)a
67
168
18
B
Ventilation
rate
(ft3/min)
500,000
220,000
43,000
C
Radon
effluent
concentration
(pCi/L)
9
35
28
aThese emission rates resulted from measurements performed by
EPA during 1978 and 1979.
-------
C-3
Table C.l-2. Values of parameters E, F, G, Z
Type
of
Mine
Copper
Limestone
Fluorspar
Bauxite
E
Radon
emission
rate
(Ci/y)
27 to 1518
7
0.2
10
F
Amount
of ore
mined
(short ton/y)
10E+6
600,000
15,000
200,000
G
Uranium-238
concentration
in ore
(ppm)
1 to 55a
2C
26
8f
Z
Radon from
exposed ore
surface
(Ci/y)
27 to 1485b
7d
0.2d
10d
<*Source: Fi76.
bFor an area of 2.6 x 10° nr (1 mile square) the rate of radon
released from exposed ore surface is estimated to be 27 Ci/y per ppm of
uranium-238.
cSource: EPA78. This value represents an average of 6 samples ranging
from 0.93 to 1.92 ppm uranium-238 from the Calera mines in Alabama.
^The radon release rate from exposed ore surface is assumed to be one
hundred times that resulting from mining.
eThis value represents an average of 4 samples from the Cave in Rock
mine in Illinois taken in 1978. The uranium-238 concentration ranged from
1.23 to 3.69 ppm.
fSource: Ad60.
C.I.2 Mills
The following formula was used to estimate the radioactive
emissions at the various milling operations:
6 = H I
(3)
where
= uranium-238 release rate, yCi/y
-------
C-4
H = uranium-238 concentration of ore, ppm
I = participate emission rate, g/s
s uCi
J = unit conversion factor = 7.2
y g ppm
The values of G, H, I for the various milling activities are listed
in table C.l-3.
Table C.l-3. Values of parameters G, H, I
Type of
mill
Iron
Copper
Zinc
Clay
Limestone
Fluorspar
Bauxite
G
Uranium-238
release rate
(Ci/y)
91
7.2 to 396
3.6
6.0
6.0
1.7
9.0
H
Uranium-238
concentration
in ore (ppm)
2
1 to 55
3
6
2
2
8
I
Parti cul ate
emission rate
(g/s)a
6.3
1
0.16
0.14
0.4
0.12
0.15
aThese emission rates are "rough estimates" based upon
judgmental observations made during the EPA environmental mine
sampling program. These values will be updated by actual
measurements currently under way.
-------
C-5
C.I.3 Tailings
The following formula was used to estimate the net radon-222
emissions from tailings piles:
K = (LM - NP) Q (4)
where
K = net radon-222 exhaled by the tailings pile, Ci/y
L = area of tailings pile in km2
M = specific radon-222 exhalation rate of the
tailings pile, pCi/m2min
N = exhalation area without tailings pile, km?
P = background exhalation rate = 30 pCi/m2min
n ,. min Ci m2
Q = unit conversion factor = 0.53 y^^2
The values of K, L, M, N are listed in table C.l-4.
-------
C-6
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-------
C-7
C.2 Radon-222 Release Rates from Large Area Sources
Radon-222 releases from large area sources (e.g., ore or coal
storage piles, waste piles, etc.) were estimated using a simple
one-dimensional diffusion model (ORNL75) as follows:
J = C E (XDe/v) (5)
where
J = radon-222 flux from source surface, pCi/cm2_s
C = radium-226 concentration per unit volume in source material,
E = radon emanation coefficient
X = radon-222 decay constant = 2.1E-6 s-1
De/v = diffusion coefficient/void fraction, cm^/s
The model assumes a source thick enough that no finite thickness
correction is necessary.
-------
C-8
REFERENCES
Ad60 Adams & Richardson, 1960, Economic Geology, 55, 1653.
EPA78 Environmental Protection Agency, 1978, Private communication
from the EPA Eastern Environmental Radiation Facility to the Office
of Radiation Programs, Environmental Protection Agency, Washington,
D. C. (December 10, 1978).
Fi76 Fitzgerald J.E.Jr., 1976, Radioactivity in Copper Ore Mining
and Dressing Industry. Proceedings of the 10th Midyear Topical
Symposium of the Helth Physics Society, Saratoga Springs, N. Y.
ORNL75 Sears M.B., et al., 1975, Correlation of Radioactive Waste
Treatment Costs and the Environmental Impact of Waste Effluents
in the Nuclear Fuel Cycle for Use in Establishing "as Low as
Practicable" GuidesMilling of Uranium Ores, ORNL-TM-4903,
Vol. 1, Oak Ridge, Tennessee.
-------
APPENDIX D
GLOSSARY OF TERMS AND ABBREVIATIONS
accelerator
A device for increasing the kinetic energy of charged
elementary particles (for example, electrons or protons)
through the application of electrical and/or magnetic forces.
activity
Radioactivity or radioactive materials.
activity
A measure of the rate at which a material is undergoing nuclear
transformations. The special unit of activity used in this
report is the curie (Ci).
AEC
Atomic Energy Commission (discontinued with formation of
ERDA and NRC on January 19, 1975).
agreement states
Those states which, pursuant to Section 274 of the Atomic Energy
Act of 1954, as amended, have entered into an agreement with the
NRC for assumption of regulatory control of byproduct, source,
and small quantities of special nuclear materials. Before
approving such an agreement, NRC must determine that the
State's radiation control program is compatible with NRC's
regulatory program and is adequate to protect public health and
safety.
alpha particle (a)
A positively charged particle emitted by certain radioactive
materials. It is made up of two neutrons and two protons;
hence, it is identical to the nucleus of a helium atom. The
alpha particle has a nuclear mass number of four and charge of
positive two.
-------
D-2
alpha radiation
An emission of alpha particles (helium nuclei) from a material
undergoing nuclear transformation.
aquifer
A water-bearing layer of permeable rock or soil.
A subsurface formation containing sufficient saturated
permeable material to yield significant quantities of water.
atomic number
The number of protons in the nucleus of a chemical element.
background radiation
Radiation in the natural human environment originating from
cosmic rays and from the naturally radioactive elements of the
earth, including those within the human body. The level of
radioactivity in an area which is produced by sources other
than the one of specific interest.
beta particle (B)
An elementary particle emitted from a nucleus during radio-
active decay. It has a single negative electric charge and
a mass equal to 1/1837 that of a proton. A beta particle is
identical to an electron.
biota
Plant and animal life.
body burden
The amount of a specified radioactive material or the summation
of the amounts of various radioactive materials present in an
animal or human body at a specified time.
boiling water reactor (BWR)
A type of nuclear power reactor that employs ordinary water
(^0) as coolant and moderator and allows bulk boiling
in the core so that steam is generated in the primary
reactor vessel.
-------
D-3
burial grounds
Areas designated for storage of containers of radioactive
wastes by near-surface burial in geologic media.
byproduct material
Radioactive material produced in a nuclear reactor, ancillary
to the reactors main purpose of producing power or fissile
materials. Fission products are usually considered to be
byproduct material.
calcination
Solidification method of disposal of liquid wastes involving
atomizing and coating of liquid on small granular solids,
followed by heating to drive off moisture.
calcine
Material heated to a temperature below its melting point to
bring about loss of moisture and oxidation resulting in a
chemically stable form.
canyon building
A heavily shielded building used in the chemical processing
of irradiated fuel and target elements. Operation and main-
tenance is by remote control.
Ci
curies
clay
As a soil separate, clay contains mineral soil particles that
are less than 0.002 millimeters in diameter. As a soil
textural class, the soil material is 40 percent or more clay,
less than 45 percent sand, and less than 40 percent silt.
cladding
The outer jacket of nuclear fuel elements preventing
corrosion of the fuel and the release of fission products into
the coolant. Stainless steel and zirconium alloys are common
cladding materials.
Synonym: hull
-------
D-4
coarse fragments
Gravel, cobblestones, or stones in soil that range in size
from 2 millimeters to 3 feet.
controlled area
Any specific region of the Hanford Reservation into which entry
by personnel is regulated by physical barrier or procedure.
counts per minute (cpm)
The number of events per unit time recorded by an instrument
designed to detect radioactive particles; especially used to
indicate the relative amount of radioactive contamination.
criticality safety
Those procedures and understandings necessary to the handling
of fissile materials in a manner that will prevent them from
reaching a critical condition.
curie
The basic unit used to describe the intensity of radioactivity
in a sample of material. One curie (Ci) equals 37 billion
disintegrations per second.
daughter products
The nuclides formed by the radioactive disintegration of a first
nuclide (parent).
deactivated
The condition of a facility or disposal site where steps have
been taken to preclude further operation or the further addition
of waste materials.
decay
The spontaneous radioactive transformation of one nuclide
into a different nuclide or into a different energy state of
the same nuclide. Every decay process has a definite half-
life.
-------
D-5
decay chain
DOE
The sequence of radioactive disintegrations in succession from
one nuclide to another until a stable daughter is reached.
Department of Energy. Established by Executive Order in
October 1977. Comprises the following former agencies:
Energy Research and Development Administration, Federal
Energy Administration, Federal Power Commission, and
parts of the Department of Interior.
dose
The energy imparted to matter by ionizing radiation per unit
mass of irradiated material at a specific location. The unit
of absorbed dose is the rad. A general term indicating the
amount of energy absorbed from incident radiation by a
specified mass.
dose commitment
The integrated dose which results from an intake of radioactive
material when the dose is evaluated from the beginning of intake
to a later time (usually 50 years;) also used for the long term
integrated dose to which people are considered committed because
radioactive material has been released to the environment.
environmental surveillance
A program to monitor the impact on the surrounding region
of the discharges from industrial operations.
enriched uranium
Uranium in which the percentage of the fissionable isotope
uranium-235 has been increased above the 0.7% contained in
natural uranium.
EPA
Environmental Protection Agency
-------
D-6
ERDA
Energy Research and Development Administration (the nuclear
program components of ERDA were formerly part of the AEC),
now part of Department of Energy.
exposure
The condition of being made subject to the action of radiation.
extraction'
A chemical process for selectively removing materials
from solutions.
fallout
Those radioactive materials deposited on the earth's surface
and in the atmosphere following the detonation of nuclear
weapons.
fertile material
A material (for example, uranium-238) not fissionable, but
which can be converted into a fissionable material
by irradiation in a reactor.
fission
The splitting of a heavy nucleus into two roughly equal parts
(which are nuclei of lighter elements), accompanied by the
release of a relatively large amount of energy and frequently
one or more neutrons.
fission products
Nuclei formed by the fission of heavy elements. Many are
radioactive. Examples: strontium-90, cesium-137.
fissionable material
Any material readily fissioned by neutrons, for example,
uranium-235 and plutonium-239.
-------
D-7
fuel cycle (nuclear, reactor)
The series of steps involved in supplying fuel for nuclear power
reactors. It includes mining, refining, the original
fabrication of fuel elements, their use in a reactor, chemical
processing to recover the fissionable material remaining in the
spent fuel, re-enrichment of the fuel material, and
refabrication into new fuel elements.
fuel (nuclear, reactor)
Fissionable material used as the source of power when placed
in a critical arrangement in a nuclear reactor.
fuel separation
(fuel reprocessing)
Processing of irradiated (spent) nuclear reactor fuel to recover
useful materials as separate products, usually involving
separation into Plutonium, uranium, and fission products.
fuel rod
A tube containing U02 or mixed oxide fuel; part of a fuel
assembly in a nuclear reactor.
fuel element
A tube, rod, or other form into which fissionable material is
fabricated for use in a reactor.
grams
gamma rays (y)
High-energy, short-wavelength electromagnetic radiation
emitted by a nucleus. Gamma radiation usually accompanies
alpha and beta emissions and always accompanies fission.
gastrointestinal dose (GI dose)
The dose to the stomach and lower tract of humans and animals
via external exposure or via internal transport of radioactive
material.
-------
D-8
g
grams
ground water
Water in the zone of saturation beneath the land surface.
half-life
The time in which half the atoms of a given quantity of a
particular radioactive substance disintegrate to another
nuclear form. Measured half-lives vary from millionths of
a second to billions of years.
half-life, biological
The time required for a living organism to eliminate, by
natural processes, half the amount of a substance that has
entered it.
half-life, effective
The time required for a radionuclide contained in a biological
system to reduce its activity by half due to the combined
result of radioactive decay and biological elimination.
heavy water
Deuterium oxide, D20. Water in which normal hydrogen atoms
have been replaced with deuterium atoms. Having a low neutron
absorption cross section, D20 readily dissipates the energy
of the high-energy neutrons which sustain a fission reaction.
Hence, D£0 is used as a moderator in some nuclear reactors.
In SRP reactors it is used as both the moderator and the primary
coolant.
HEPA
High efficiency particulate air filter. A type of filter
designed to remove 99.9 percent of particles down to 0.3 urn in
diameter from a flowing air stream.
hood
A canopy and exhaust duct used to confine hazardous materials in
order to reduce the exposure of industrial workers.
-------
D-9
hypothetical maximum individual (max man)
A postulated person who is assumed to receive the maximum
credible radiological dose through each of the exposure
pathways from the source being considered.
inactive
The condition of a facility or disposal site which is not
presently being operated or to which materials are not being
added.
ICRP
International Commission on Radiological Protection
ion exchange
A reversible chemical reaction between a solid and a fluid
mixture by means of which ions may be interchanged.
isotope
One of two or more forms of an element that differ in atomic
weight. Nuclides with the same atomic number, (i.e., the
same chemical element, characterized by the number of protons
contained in the atomic nucleus) but with different atomic
masses (i.e., different numbers of neutrons contained in the
nucleus). Although chemical properties are the same,
radioactive and nuclear (radioactive decay) properties may be
quite different for each isotope of an element.
km
kilometers (1 kilometer = 1000 meters or 0.621 mile)
leaching
Extracting material from a solid by passing water
or a solution through the solid material.
light water
Normal water (H20), as distinguished from heavy water (DgO).
-------
D-10
light water reactor
A reactor in which ordinary water (h^O) is used as the coolant
and moderator. In such reactors the water is either allowed to
boil (boiling water reactor or BWR) or pressurized to prevent
boiling (pressurized water reactor or PWR).
long-lived nuclides
Radioactive isotopes with half-lives greater than about 30
years. Most long-lived nuclides of interest to waste
management have half-lives on the order of thousands to
millions of years. For example:
239Pu _ 24,400 years; 99Tc - 2.1 x 105 years;
129I - 1.6 x 107 years.
low-level waste
Wastes containing types and concentrations of radioactivity
such that shielding to prevent personnel exposure is not
required.
m
1. meter
2. as prefix, milli. See "miHi."
micro (y)
Prefix indicating one millionth (1 microgram = 1/1,000,000
of a gram or 10-6 gram).
milli
Prefix indicating one thousandth
millirem
One thousandth of a rem
ml
milliliters
-------
D-ll
moderator
A material, such as heavy water, used in a reactor to slow
down the high-velocity neutrons which sustain a fission
reaction.
mR
millirads
mrem
millirem
nano
Prefix indicating one thousandth of a micro unit (1 nano-
curie = 1/1000 of a microcurie or 10-9 cur-je).
natural (normal) uranium
Uranium as found in nature. It is a mixture of the fertile
uranium-238 isotope (99.3%), the fissionable uranium-235
isotope (0.7%), and a minute percentage of uranium-234.
neutron
An uncharged elementary particle with a mass nearly equaled by
that of the proton. Neutrons are part of the fission chain
reaction in a nuclear reactor. They can also be generated by
spontaneous fission and by collision of high energy y rays
and a particles with some nuclei.
NRC
Nuclear Regulatory Commission (formerly part of AEC).
nuclide
Any atomic nucleus specified by its atomic weight, atomic
number, and energy state. A radionuclide is a radioactive
nuclide.
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D-12
nuclear radiation
Particles and electromagnetic energy given off by
transformations occurring in the nucleus of an atom.
off-gas
The gas given off in any stage of an industrial process.
pCi
picocuries
permeability, soil
That quality of the soil that enables it to transmit water or
air. Terms used to describe permeability in inches per hour
are:
Very slow Less than 0.06 inches
Slow 0.6 to 0.2 inches
Moderately slow 0.2 to 0.6 inches
Moderate 0,6 to 2.0 inches
Moderately rapid 2.0 to 6.0 inches
Rapid 6.0 to 2.0 inches
Very rapid More than 20 inches
person-rem
Used as a unit of population dose; the average dose per
individual expressed in rems times the population affected.
PH
A measure of the hydrogen ion concentration in aqueous solu-
tions. Acidic solutions have a pH from zero to 7. Basic
solutions have a pH from 7 to 14.
pico
Prefix indicating one millionth of a micro unit (1 picocurie
1/1,000,000 of a microcurie or 10-12 curie).
-------
D-13
population dose
(population exposure)
The summation of individual radiation doses received by all
those exposed to the source or event being considered.
Plutonium
A radioactive element with atomic number 94. Its most im-
portant isotope is fissionable plutonium-239, produced by
neutron irradiation of uranium-238.
power reactor
A nuclear reactor designed to produce heat for conversion to
electrical energy or mechanical propulsion.
ppm
parts per million
precipitation scavenging
The process by which rain or snow removes particulates or
reactive vapors from the atmosphere and deposits them on the
ground surface.
production reactor
A nuclear reactor designed primarily for large-scale production
of plutonium, tritium, and other radionuclides by neutron
irradiation. A nuclear reactor designed for transforming one
nuclide into another; usually, a conversion of natural uranium
into plutonium.
Purex
rad
A solvent extraction process in which uranium and plutonium
are selectively separated from each other and from fission
products by extraction from nitric acid solutions with
tributylphosphate in a hydrocarbon diluent.
Radiation absorbed dose. The basic unit of absorbed dose of
ionizing radiation. One rad is equal to the absorption of
100 ergs of radiation energy per gram of matter.
-------
D-14
radiation (ionizing)
Particles and electromagnetic energy emitted by nuclear
transformations which are capable of producing ions when
interacting with matter; gamma rays and alpha and beta
particles are primary examples in INEL waste.
radioiodines
Isotopes of iodine which are radioactive.
radioactive (decay)
Property of undergoing spontaneous nuclear transformation in
which nuclear particles or electromagnetic energy are emitted.
radioactivity
The spontaneous decay or disintegration of unstable atomic
nuclei, accompanied by the emission of radiation.
radionuclide
An unstable nuclide of an element that decays or disintegrates
spontaneously, emitting radiation.
radioisotope
A radioactive isotope. An unstable isotope of an element that
decays or disintegrates spontaneously, emitting radiation. More
than 1300 natural and artificial radioisotopes have been
identified.
radwaste
Waste materials which are radioactive.
reactor
A device by means of which a fission chain reaction can be
initiated, maintained, and controlled. A nuclear reactor.
recycle
The returning of uranium and plutonium (recovered in spent fuel
reprocessing) for reuse in new reactor fuel elements.
-------
D-15
release limit
(release guide)
A control number which regulates the concentration or amount
of radioactive material released to the environment in an
industrial situation; usually dose to persons in the environ
ment derived from environmental behavior of the released
material so that the dose is kept below a selected control
value.
rem
A dose unit which takes into account the relative biological
effectiveness (RBE) of the radiation. The rem ("roentgen
equivalent man") is defined as the dose of a particular type
of radiation required to produce the same biological effect as
one roentgen of (0.25 Mev) gamma radiation. Amillirem (mrem)
is one thousandth of a rem.
roentgen (R)
A measure of the ability of gamma or X rays to produce
ionization in air. One roentgen corresponds to the absorption
of about 86 ergs (100 ergs = 6.24 x 10? million electron
volts, Mev) of energy from X- or gamma radiation, per gram of
air. The corresponding absorption of energy in tissue may be
from one-half to two times as great, depending on the energy of
the radiation and the chemical composition of the tissue. The
roentgen is thus more useful as a measure of the amount of
gamma or X rays to which one is exposed than as a measure of
the dose of such radiation actually received.
reprocessing
Chemical processing of irradiated nuclear reactor fuels to
remove desired constituents.
retention basin
An excavated and lined area used to hold contaminated fluids
until radioactive decay reduces activities to levels permissible
for release.
-------
D-16
retired facility
A facility which has been shut down with no intentions of
restarting and which has had appropriate controls and safeguards
placed on it.
salt cake
The solid residue resulting from a concentration of high-level
liquid waste in underground waste storage tanks.
scavenging
See precipitation scavenging.
separations
Chemical processes used to separate nuclear products from
byproducts and from each other.
short-lived nuclides
Radioactive isotopes with half-lives not greater than about
30 years, e.g., 137Cs and 90Sr.
solidification
Conversion of radioactive waste to a dry, stable solid.
solid wastes (radioactive)
Either solid radioactive material or solid objects which contain
radioactive material or bear radioactive surface contamination.
solvent extraction
A process in which materials are selectively removed from an
aqueous solution by contact with an immiscible organic solvent.
source material
Uranium or thorium or any ores which contain at least 0.05% of
uranium or thorium
source term
Release rates (in curies per year) to the atmosphere from each
point or area source are known collectively as the source term.
-------
D-17
special nuclear material (SNM)
Plutonium, 233^ 235U} or uran-jum enriched to a higher
percentage than normal of the 233 or 235 isotopes.
stability (atmospheric)
A description of the atmospheric forces on a parcel of air
following vertical displacement in an atmosphere otherwise
in hydrostatic equilibrium; if the forces tend to return the
parcel to its original level, the atmosphere is stable; if the
forces tend to move the parcel further in the direction of
displacement, the atmosphere is unstable, and if the air parcel
tends to remain at its new level the atmosphere has neutral
stability.
standby
The condition where a facility or burial ground, etc., is placed
in a nonoperating condition but is maintained in readiness for
subsequent operation
storage basin
A water-filled facility for holding irradiated reactor fuels
with the water acting as a shield.
tank
A large metal container located underground for storage of
liquid wastes
tank farm
An installation of interconnected underground containers
(tanks) for storage of high-level waste.
thorium
A naturally radioactive element with atomic number 90 and, as
found in nature, and atomic weight of approximately 232. The
fertile thorium 232 isotope is abundant and can be transmuted
to fissile uranium 233 by neutron irradiation.
total body dose
The radiation dose to the entire body.
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D-18
tracer
A radionuclide(s) or chemical introduced in minute quantities
to a system or process for the purpose of using radiation or
chemical detection techniques to follow the behavior of the
process or system.
transmutation
The process whereby one nuclide changes (or is changed) into
another, usually by means of bombardment with nuclear particles.
transuranic elements
Elements with mass numbers greater than 92, including
neptunium, plutonium, americium, and curium.
trench
A long and narrow excavation in the ground for solid waste.
Unless qualifying descriptions are given, a trench is unlined,
and its walls are unsupported. After the solid wastes are
placed in position, the trench is filled to grade level with
some of the removed soil.
TRIGA reactor
A research and training pool type nuclear reactor built by
Gulf General Atomics which has a compact zirconium hydride
core and which can operate in either a steady state or a pulsing
mode.
tritium
A radioactive isotope of hydrogen with two neutrons and one
proton in the nucleus. It is heavier than deuterium (heavy
hydrogen). Tritium (T or 3^) is used in industrial thickness
gages, as a label in tracer experiments, in controlled nuclear
fusion experiments, and in thermonuclear weapons. It is
produced primarily by neutron irradiation of lithium-6.
water table
Upper boundary of an unconfined aquifer surface below which
saturated groundwater occurs; defined by the levels at which
water stands in wells that barely penetrate the aquifer.
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D-19
wind rose
A diagram designed to show the distribution of prevailing
wind directions at a given location; some variations include
wind speed groupings by direction.
u ran i urn
A naturally radioactive element with the atomic number 92 and
an atomic weight of approximately 238. The two principal
naturally occurring isotopes are the fissionable uranium-235
(0.7% of natural uranium) and the fertile uranium-238 (99.3%
of natural uranium).
USAEC
United States Atomic Energy Commission (see AEC)
waste, radioactive
Equipment and materials (from nuclear operations) that are
radioactive or have radioactive contamination and for which
there is not recognized use or for which recovery is
impractical.
water table
The upper surface of the ground water.
mu, a prefix. Same as "micro."
microcuries
yg
micrograms
urn
micrometers
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APPENDIX E
LIST OF ELEMENTS
ELEMENT
actinium
aluminum
americium
antimony
argon
arsenic
astatine
barium
berkelium
beryllium
bismuth
boron
bromine
cadmium
calcium
californium
carbon
cerium
cesium
chlorine
chromium
cobalt
columbium
copper
curium
dysprosium
einsteinium
erbium
europium
fermium
fluorine
francium
gadolinium
gallium
germanium
gold
hafnium
helium
holmium
hydrogen
indium
iodine
iridium
iron
krypton
lanthanum
lawrencium
lead
lithium
lutetium
magnesium
manganese
mendelevium
SYMBOL
Ac
Al
Am
Sb
Ar
As
At
Ba
Bk
Be
Bi
B
Br
Cd
Ca
Cf
C
Ce
Cs
Cl
Cr
Co
Nb
Cu
Cm
Dy
Es
Er
Eu
Fm
F
Fr
Cd
Ca
Ge
Au
Hf
He
Ho
H
In
1
Ir
Fe
Kr
La
Lw
Pb
Li
Lu
Mg
Mn
Md
ATOMIC NUMBER
89
T3
95
51
18
33
85
56
97
4
83
5
35
48
20
98
6
58
55
17
14
27
(see niobium)
29
%
16
99
68
63
100
9
87
64
31
32
79
72
2
67
1
49
53
77
26
36
57
103
82
3
71
12
25
101
ELEMENT
mercury
molybdenum
neodymium
neon
neptunium
nickel
niobium
nitrogen
nobelium
osmium
oxygen
palladium
phosphorus
platinum
plutonium
polonium
potassium
praseodymium
promethium
protactinium
radium
radon
rhenium
rhodium
rubidium
ruthenium
samarium
scandium
selenium
silicon
silver
sodium
strontium
sulfur
tantalum
technetium
tellurium
terbium
thallium
thorium
thulium
tin
titanium
tungsten
uranium
vanadium
wolfram
xenon
ytterbium
yttrium
zinc
zirconium
SYMBOL
Hg
Mo
Nd
Ne
Np
Ni
Nb
N
No
Os
O
Pd
P
Pt
Pu
Po
K
Pr
Pm
Pa
Ra
Rn
Re
Rh
Rb
Ru
Sm
Sc
Se
Si
Ag
Na
Sr
S
Ta
Tc
Te
Tb
Tl
Th
Tm
Sn
Ti
W
U
V
W
Xe
Yb
Y
Zn
Zr
ATOMIC NUMBER
80
42
60
10
93
28
41
7
102
76
8
46
15
78
94
84
19
59
61
91
88
86
75
45
37
44
62
21
34
14
47
11
38
16
73
43
52
65
81
90
69
50
22
74
92
23
(see tungsten)
54
70
39
30
40
U.S. GOVERNMENT PRINTING OFFICE. 1979 0-300-984/6452
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