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EPA 520/1-83-001
Draft
Background Information Document
Proposed Standards for Radionuclides
March 1983
U.S. Environmental Protection Agency
Office of Radiation Programs
Washington, D.C. 20460
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CONTENTS
Page
1. INTRODUCTION 1-1
2. DEPARTMENT OF ENERGY FACILITIES
Summary 2.0-1
2.1 Argonne National Laboratory 2.1-1
2.2 Brookhaven National Laboratory 2.2-1
2.3 Fermi National Accelerator Laboratory 2.3-1
2.4 Hanford Reservation 2.4-1
2.5 Idaho National Engineering Laboratory 2.5-1
2.6 Lawrence Livermore National Laboratory 2.6-1
2.7 Los Alamos National Laboratory 2.7-1
2.8 Oak Ridge Reservation 2.8-1
2.9 Paducah Gaseous Diffusion Plant 2.9-1
2.10 Portsmouth Gaseous Diffusion Plant 2.10-1
2.11 Rocky Flats Plant 2.11-1
2.12 Savannah River Plant 2.12-1
2.13 Feed Materials Production Center 2.13-1
2.14 Ames Laboratory 2.14-1
2.15 Battelle-Columbus Laboratory 2.15-1
2.16 Bettis Atomi Power Laboratory 2.16-1
2.17 Knolls Atomic Power Laboratory 2.17-1
2.18 Lawrence Berkeley Laboratory 2.18-1
2.19 Mound Facility 2.19-1
2.20 Nevada Test Site 2.20-1
2.21 Pantex Plant 2.21-1
2.22 Pinellas Plant 2.22-1
2.23 Rockwell International Corporation 2.23-1
2.24 Sandia National Laboratories 2.24-1
2.25 Shippingport Atomic Power Station 2.25-1
2.26 Stanford Linear Accelerator Center 2.26-1
2.27 Worldwide Impact of Selected Radionuclides 2.27-1
2.28 Future operations at DOE Facilities 2.28-1
2.28(A) Resumption of Operations at the PUREX Plant 2.28-1
2.28(B) Resumption of L-Reactor Operations at the
Savannah River Plant 2.28-3
iii
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CONTENTS (Continued)
Page
3. NRC-LICENSED FACILITIES AND NON-DOE FEDERAL FACILITIES
3.1 Research and Test Reactors 3.1-1
3.2 Accelerators 3.2-1
3.3 Radiopharmaceutical Industry 3.3-1
3.4 DOD Facilities 3.4-1
3.4A Armed Forces Radiobiology
Research Institute 3.4A-1
3.4B U.S. Army Facilities 3.4B-1
3.4C U.S. Navy Facilities 3.4C-1
3.5 Radiation Source Manufacturers 3.5-1
4. COAL-FIRED UTILITY AND INDUSTRIAL BOILERS 4.0-1
4.1 Utility Boilers 4.1-1
4.2 Industrial Boilers 4.2-1
5. URANIUM MINES 5-1
6. PHOSPHATE INDUSTRY FACILITIES
6.1 Phosphate Rock Processing Plants 6.1-1
6.2 Wet Process Fertilizer Plants 6.2-1
6.3 Elemental Phosphorus Plants 6.3-1
7. MINERAL EXTRACTION INDUSTRY FACILITIES
Metal Mines, Mills, and Smelters
7.1 Aluminum Industry 7.1-1
7.2 Copper Industry 7.2-1
7.3 Zinc Industry 7.3-1
7.4 Lead Industry 7.4-1
APPENDICES
A. ASSESSMENT METHODOLOGY A-l
B. THE BASIS FOR RISK ESTIMATES ' B-l
C. CALCULATIONS OF RADON-222 CONCENTRATIONS C-l
iv
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Chapter 1: INTRODUCTION
The purpose of this report is to serve as a background information
document in support of the Environmental Protection Agency's (EPA)
proposed standards for sources of emissions of radionuclides pursuant
to Section 112 of the Clean Air Act. It presents an analysis of the
public health impact caused by radionuclides emitted into the air from
facilities that are the subject of this rulemaking.
These facilities are examined as six major source categories:
(1) Department of Energy (DOE) facilities
(2) Nuclear Regulatory Commission (NRC) licensedC1) and non-DOE
Federal facilities
(3) Coal-fired utility and industrial boilers
(4) Uranium mines
(5) Phosphate industry facilities
(6) Mineral extraction industry facilities
For each source category, we present the following information:
(1) A general description of the source category
(2) A brief description of the processes that lead to the
emission of radionuclides into air
(l)Sources are licensed by the Nuclear Regulatory Commission
(NRC) or States that 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.
1-1
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(3) A summary of emissions data
(4) Estimates of the radiation doses and health risks to both
individuals and populations
This is a draft document. After a public comment period, it will
be revised as necessary and issued in final form when the standards are
promulgated.
Emission Data
Insofar as possible, measured radionuclide emission data were used
to estimate health impacts. In the absence of measured data, estimates
were used that were based on calculated or extrapolated values. The
data for DOE facilities were obtained from DOE's Effluent Information
System for the calendar year 1981 (DOE81); the data for NRC-licensed
facilities were obtained from NRG annual effluent reports; and the data
for the other categories, such as coal-fired utility and industrial
boilers, uranium and nonuranium mines, and the various extraction
industries, were usually obtained from special reports prepared on
contract for the EPA.
Health Impact Assessment
The public health assessment includes estimates of the following
radiation exposures and health risks:
(1) Dose-equivalent rates to the individuals at highest risk
(maximum individual)
(2) Collective dose-equivalent rates to population groups
(3) Lifetime risks to the maximum 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 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
DOE facilities were analyzed individually on a site-by-site
basis. The NRG- 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 by selecting a reference facility
that represents the source category. Dose and risk were calculated by
1-2
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using AIRDOS-EPA and RADRISK computer models developed by Oak Ridge
National Laboratory under contract to the EPA. These computations are
based upon the latest information on transport, uptake, and metabolic
behavior of the various radionuclides and are described in detail in a
recent EPA report (EPA79).
Information on emission control technology for facilities in this
report is published in documents that are available in Docket A-79-11,
Central Docket Section, Gallery One, West Tower Lobby, EPA, 401 M Street,
S.W., Washington, D.C.
The Maximum Individual
Dose-equivalent rates and radon decay exposures are presented for
the maximum individual. Each location on the assessment grid is
assessed to determine the lifetime risk to an individual from all
pathways. The position on the grid that provides the greatest
calculated lifetime risk is considered to be the location of the
maximum individual.
The dose rates presented for the maximum individual are 70-year
committed dose equivalents. This is also the dose-equivalent rate in
the 70th year following the start of exposure.
Radon decay 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 pCi/L radon-222 = 0.7 WL).
Regional Population
The term regional 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 radon decay product exposures
are expressed in units of person-rem/year and person-working levels,
and are the sum of the dose-equivalent rates or radon decay product
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 Risk and Number of Fatal Cancers
The individual lifetime risk is the probability of fatal cancer to
an individual due to a lifetime of exposure (70 years on the average)
to the concentrations of radionuclides estimated for that individual.
1-3
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The number of fatal cancers per year is that potential number of
cancers in the population committed by one year's release of
radionuclides from the facility. They are expected to occur many years
after the year in which the releases take place.
Numeric Notation and Units
Throughout this report, numeric values are frequently expressed in
a modified scientific format. For example, 0.00123, which is equal to
1.23 x 10~3, may be expressed as 1.23E-3; 3210, which is equal to
3.21 x 103, may be expressed as 3.21E+3.
Metric units have been used for reporting data, except in a few
instances where referenced data are presented in the original units.
1-4
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REFERENCES
BEIR72 Advisory Committee on the Biological Effects of Ionizing
Radiation, The Effects of Population Exposures to Low Levels
of Ionizing Radiation, National Academy of Sciences,
Washington, D.C., 1972.
DOE81 Department of Energy, Effluent Information System (EIS),
Calendar year 1981, DOE, Washington, D.C.
EPA77 Environmental Protection Agency, Radiological Quality of the
Environment in the United States, EPA 520/1-77-009, Office of
Radiation Programs, Washington, D.C., 1977.
EPA79 Environmental Protection Agency, AIRDOS-EPA, A Computerized
Methodology for Estimating Environmental Concentrations and
Dose to Man from Airborne Releases of Radionuclides,
EPA 520/1-79-009, December 1979.
Mo77 Moore R.E., The AIRDOS-II Computer Code for Estimating
Radiation Dose to Man from Airborne Radionuclides in Areas
Surrounding Nuclear Facilities, ORNL-5245, Environmental
Sciences Division Publication No. 974, Oak Ridge National
Laboratory, Oak Ridge, Tennessee 37830, 1977.
1-5
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Chapter 2: DEPARTMENT OF ENERGY FACILITIES
Summary
The radiation doses and risks of fatal cancers to individuals and
populations around Department of Energy facilities were estimated using
the methods in Appendices A and B. These estimates are summarized in
Tables 2-A through 2-D. More detailed information, including a general
description of the facility, a summary of the processes causing the
emissions, estimates of the amount of emissions, and more detailed
estimates of dose and risk are found in the respective sections of this
chapter.
Table 2-A. Summary of dose rates and risks to nearby individuals
for facilities with the largest emissions
Facility
Argonne
National
Laboratory
Brookhaven
National
Laboratory
Feed Materials
Production
Center
Fermi National
Accelerator
Laboratory
Principal
Radio-
nuclide
Ar-41
Kr-85
H-3
0-15
Ar-41
Xe-127
U-234
U-238
H-3
C-ll
emissions
Quantity
(Ci/y)
0.4
6.7
660
36,000
170
2.3
0.11
0.11
0.42
1500
Maximum individual
Principal
organ
Pulmonary
Bone(b)
Muscle
Red marrow
Pulmonary
Bone(b)
Muscle
Red marrow
Pulmonary
Bone(b)
Red Marrow
Kidneys
Bone(b)
Muscle
Red marrow
Dose rate
(mrem/y)
<0.1
<0.1
<0.1
<0.1
0.4
0.6
0.5
0.5
88(c)
26
1.8
12
0.7
0.6
0.7
Risk(a)
(x 10-6)
0.0006
9
200
10
See footnotes at end of table.
2,0-1
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Table 2-A. Summary of dose rates and risks to nearby individuals
for facilities with the largest emissions (Continued)
Principal
Facility
Hanford
Reservation
100 Area
200 Area
300-400 Areas
Idaho National
Engineering
Laboratory
Lawrence Livermore
National
Laboratory
Los Alamos
National
Laboratory
(12 Technical
Areas)
Technical
Area 33
Oak Ridge
Reservation
Radio-
nuclide
H-3
Ar-41
Kr-88
Cs-138
Cs-137
Pu-239
Kr-88
H-3
Ar-41
Kr-85
1-131
H-3
N-13
0-15
H-3
C-ll
N-13
0-15
Ar-41
H-3
H-3
Kr-85
1-131
Xe-133
U-234
emissions
Quantity
(Ci/y)
18
65,000
540
11,000
.05
.0004
450
400
2,500
59,000
0.055
2,600
170
170
1,100
130,000
25,000
200,000
1,400
6,100
11,000
6,600
0.6
32,000
0.12
Maximum individual
Principal
organ
Pulmonary
Bone(b)
Muscle
Red marrow
Red marrow
Pulmonary
Pulmonary
Bone(b)
Pulmonary
Bone(b)
Thyroid
Muscle
Red marrow
Red marrow
Stomach
Kidneys
Bone(b)
Muscle
Red marrow
Spleen
Bone(b)
Muscle
Red marrow
Spleen
Pulmonary
Bone(b)
Thyroid
Kidneys
Lower large
intestine
Dose rate Risk
(mrem/y) (x 10"6)
2 40
2
2
2
<0.1 0.3
<0.1
1 30
2
<0.1 0.5
<0.1
0.12
<0. 1
<0. 1
1 30
2
1
11 200
9
11
10
0.5 10
0.7
0.7
0.7
50(d) 40
8
9
5
5
See footnotes at end of table.
2.0-2
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Table 2-A. Summary of dose rates and risks to nearby individuals
for facilities with the largest emissions (Continued)
Principal emissions
Facility
Faducah Gaseous
Diffusion
Plant
Portsmouth Gaseous
Diffusion
Plant
Rocky Flats
Plant
Savannah River
Plant
Radio- Quantity
nuclide (Ci/y)
Tc-99
U-234
U-238
Tc-99
Th-234
U-234
H-3
U-234 /5/8
Pu-239/40
H-3
Ar-41
Kr-85
Kr-88
Xe-133
1-131
1-129
0.006
0.01
0.04
0.1
0.06
0.09
0.43
0.00003
0.000008
350,000
62,000
840,000
1,500
3,900
0.05
0.16
Maximum individual
Principal
organ
Pulmonary
Bone(b)
Thyroid
Kidneys
Pulmonary
Bone(b)
Thyroid
Kidneys
Pulmonary
Bone(b)
Red marrow
Bone(b)
Thyroid
Stomach
Dose rate
(mrem/y)
5(e)
7
0.2
4
7(e)
11
2
5
<0.1
<0.1
<0.1
2
5
2
Risk(a)
(x 10-6)
20
30
0.04
40
(a)Off-site location at point of highest dose equivalent.
(fe)Endosteal cells.
(c)Lung clearance class for uranium: one-third D, one-third W,
one-third Y.
(d)Lung clearance class for uranium: all uranium-238 and one-half
uranium-234, Y; one-half uranium-234, W.
(e)Lung clearance class for uranium: all W.
2.0-3
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Table 2-B. Summary of dose rates and risks to the regional population
for facilities with largest emissions
Principal emissions
Facility Radio-
nuclide
Argonne
National
Laboratory
Brookhaven
National
Laboratory
Feed Materials
Production
Center
Fermi National
Accelerator
Laboratory
Han ford
Reservation
100 Area,
200 Area, and
300-400 Areas
Idaho National
Engineering
Laboratory
Lawrence Livermore
National
Laboratory
Los Alamos
National
Laboratory
(12 Technical
Areas) and
Technical
Area 33
Ar-41
Kr-85
H-3
0-15
Ar-41
Xe-127
U-234
U-238
H-3
C-ll
H-3
Ar-41
Kr-88
Cs-138
H-3
Ar-41
Kr-85
1-131
H-3
N-13
0-15
H-3
C-ll
N-13
0-15
Ar-41
Quantity
(Ci/y)
0.4
6.7
660
36,000
170
2
0.11
0.11
0.4
1,500
18
65,000
990
11,000
400
2,500
59,000
0.055
2,600
170
170
7,200
130,000
25,000
200,000
1,400
Regional population
Principal Dose rate
organ (pers-rem/y)
Pulmonary
Bone(a)
Muscle
Red marrow
Pulmonary
Bone(a)
Muscle
Red marrow
Pulmonary
Bone(a)
Red Marrow
Kidneys
Bone(a)
Muscle
Red marrow
Pulmonary
Bone(a)
Muscle
Red marrow
Bone(a)
Thyroid
Red marrow
Stomach
Kidneys
BoneU)
Muscle
Red marrow
Spleen
<0. 1
<0. 1
<0. 1
<0. 1
3
3
3
3
440(b)
114
8
56
1
1
1
11
13
10
11
0.3
5.5
0.2
7
6
63
53
61
57
Fatal can-
cers/year of
operation
<0.001
<0.001
0.02
<0.001
0.003
<0.001
0.002
0.01
See footnotes at end of table.
2.0-4
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Table 2-B. Summary of dose rates and risks to the regional population
for facilities with largest emissions (Continued)
Principal emissions
Facility
Oak Ridge
Reservation
Paducah Gaseous
Diffusion
Plant
Portsmouth
Gaseous Dif-
fusion Plant
Rocky Flats
Savannah River
Plant
Radio- Quantity
nuclide (Ci/y)
H-3
Kr-85
1-131
Xe-133
U-234
Tc-99
U-234
U-238
Tc-99
Th-234
U-234
H-3
U-234 /5/8
Pu-239/240
H-3
Ar-41
Kr-85
11,000
6,600
0.6
32,000
0.12
0.006
0.01
0.04
0.1
0.06
0.09
0.4
0.00003
0.000008
350,000
62,000
840,000
Regional population
Principal Dose rate
organ (pers-rem/y)
Pulmonary
Bone(a)
Thyroid
Kidneys
Lower large
intestine
Pulmonary
Bone(a)
Thyroid
Pulmonary
Bone(a)
Thyroid
Kidneys
Pulmonary
Bone(a)
Red marrow
Pulmonary
Thyroid
Stomach
212(c)
22
15
15
13
3.4(d)
13
0.4
ll(d)
35
8
17
0.1
0.2
0.01
98
120
110
Fatal can-
cers/year of
operation
0.01
<0.0001
<0.001
<0.001
0.03
Kr-88
Xe-133
1-131
1-129
1,500
3,900
0.05
0.16
(a)Endosteal cells.
(b)Lung clearance class for uranium: one-third D, one-third W,
one-third Y.
(c)Lung clearance class for uranium: all uranium-238 and one-half
uranium-234, Y; one-half uranium-234, W.
clearance class for uranium: all W.
2.0-5
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Table 2-C. Summary of individual dose rates and risks to
nearby individuals for facilities with small health impact
Facility
Ames Laboratory
Battelle-Columbus
Laboratory
Bettis Atomic Power
Laboratory
Knolls Atomic Power
Lab. (Kesselring Site)
Knolls Atomic Power
Lab. (Knolls Site)
Knolls Atomic Power
Lab. (Windsor Site)
Lawrence Berkeley
Laboratory
Mound Facility
Nevada Test Site
Pantex Plant
Pinellas Plant
Rockwell International
Corp.
Sandia Laboratories
Shippingport Atomic
Power Station
Stanford Linear
Accelerator Center
Principal
organ
All organs
Bone(b)
Pulmonary
Bone(b)
(c)
Bone(b)
Thyroid
Bone(b)
Bone(b)
Pulmonary
Lower large
intestine
Bone(b)
Spleen
All organs
Bone(b)
Dose rate
(mrem/y)
0.001
0.008
0.004
0.08
(c)
0.003
1.6
0.2
0.002
0.005
0.4
0.00004
0.0009
0.0003
0.006
Risk(a)
(x 10~6)
0.02
0.04
0.02
1
(c)
0.04
10
4
0.03
0.01
5
0.00007
0.02
0.005
0.1
(a)off-site location at point of highest dose equivalent.
(b)Endosteal cells.
(c)Kesselring and Knolls sites were assessed as a single combined site.
2.0-6
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Table 2-D. Summary of dose rates and risks to the regional population
for facilities with small health impact
Facility
Ames Laboratory
Batte lie-Columbus
Laboratory
Bettis Atomic Power
Laboratory
Knolls Atomic Power
Lab. (Kesselring Site)
Knolls Atomic Power
Lab. (Knolls Site)
Knolls Atomic Power
Lab. (Windsor Site)
Lawrence Berkeley
Laboratory
Mound Facility
Nevada Test Site
Pantex Plant
Pinellas Plant
Rockwell International
Principal Dose rate
organ (pers-rem/y)
Average all organs
Bone
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2.1 Argonne National Laboratory; Argonne. Illinois
2.1.1 General Description
Argonne National Laboratory (ANL) occupies the central 6.88 km2
of a 15.14 km2 tract in DuPage County, 43 km southwest of downtown
Chicago, and 39 km due west of Lake Michigan. It lies in the Des
Plaines River Valley, south of Interstate Highway 55 and west of
Illinois Highway 83.
Argonne is an energy research and development laboratory with
several principal objectives. It conducts a broad program of research
in the basic energy and related sciences (physical, chemical, material,
nuclear, biomedical, and environmental) and serves as an important
engineering center for the study of nuclear and nonnuclear energy
sources.
A significant portion of these laboratory studies requires the use
of radioactive and chemically-toxic substances.
2.1.2 Process Description
The principal nuclear facilities at the Laboratory are a 200 kW
light-water cooled and moderated biological research reactor (Janus)
fueled with fully-enriched uranium; one critical assembly or zero power
reactor (ZPR-9), that is fueled at various times with plutonium,
uranium, or a combination of the two; the Argonne Thermal Source
Reactor (ATSR), a 10 kW research reactor fueled with enriched uranium;
a prototype superconducting heavy ion linear accelerator; a 60-inch
cyclotron; several other charged particle accelerators (principally of
the Van de Graaff and Dynamitron type); a large fast neutron source
(IPNS, Intense Pulsed Neutron Source) in which high energy protons
strike a heavy metal target to produce the neutrons; cobalt-60
irradiation sources; chemical and metallurgical plutonium laboratories;
and several hot cells and laboratories designed for work with
multicurie quantities of the actinide elements. Two major facilities,
a 12.5 GeV proton accelerator (ZGS, the Zero Gradient Synchrotron) and
a 5 MW heavy water-enriched uranium reactor (CP-5) were not in
operation during 1981 and are awaiting decontamination and
decommissioning.
2.1.3. Radionuclide Emissions and Existing Control Technology
Airborne emissions from Argonne National Laboratory for 1981 are
identified in Table 2.1-1. The emissions for years 1979 through 1981
are summarized in Table 2.1-2. The primary source of tritiated water
vapor and argon-41 prior to September 1979 was the CP-5 reactor that
was taken out of service at that time. This explains a significant
reduction in air emissions as indicated in Table 2.1-1. The only
significant releases originated from the JANUS Reactor and the hot cell
2.1-1
-------
facility in Building 212. No controls are reported for the JANUS
Reactor; however, the exhaust of the hot cell facility employs both wet
scrubbers and electrostatic precipitators. Calculations of health
impact were based on a single release point (stack height of 61 meters)
located approximately 500 meters south of the plant boundary.
2.1.4 Health Impact Assessment of Argonne National Laboratory
The estimated annual radiation dose rates from radionuclide
emissions from Argonne National Laboratory are shown in Table 2.1-3.
The maximum individual is located 900 meters north of the assumed
release point, approximately 400 meters beyond the site boundary due to
the elevation of the release point (61 meters). The primary exposure
pathway is external exposure resulting from argon-41.
Risks of having fatal cancer from exposure to the radioactive
emissions from this facility are identified in Table 2.1-4. The
highest individual lifetime risk is 6E-10 and the risk for the regional
population is 7E-7 fatal cancers each year of operation.
Table 2.1-1. Radionuclide emissions from Argonne National Laboratory,
1981
o n j. T.J Emissions
Source Radionuclide
JANUS Reactor Argon-41 3.8E-1
Hot cell exhaust Krypton-85 6.7
Antimony-125 1.7E-5
Chemical Engineering
Laboratory Tritium 6.9E-7
2.1-2
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Table 2.1-2. Radionuclide emissions from Argonne National Laboratory,
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Antimony-125
Argon-41
Tritium
Krypton-79
Kryp ton-85
Krypton-85m
Xenon-133
Xenon-135
8.5E-5
7 . 1E+3
6.6E+2
1.5E-4
9.0
1.4E-5
3.6E-5
4.7E-4
1.5E-4
8.1E-1
9.0
7.1E-4
5.1
7.6E-5
1.4E-5
6.2E-5
1 . 7E-5
3.8E-1
6.9E-7
0
6.7
0
0
0
Table 2.1-3. Radiation dose rates from radionuclide emissions
from Argonne National Laboratory
Q Maximum individual Regional population
(mrem/y) (person-rem/y)
Endosteal
Spleen
Red Marrow
Muscle
Pulmonary
3.3E-5
3.3E-5
3.1E-5
3.1E-5
3.1E-5
2.6E-3
2.6E-3
2.5E-3
2.4E-3
2.5E-3
Weighted Sum 2.9E-5 2.3E-3
Table 2.1-4. Fatal cancer risks due to radionuclide emissions from
Argonne National Laboratory
Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
ANL 6E-10 7E-7
2.1-3
-------
REFERENCES
DOE82 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
Go82 Golchert N.W., Duffy T.L. and Sedlet J., Environmental
Monitoring at Argonne National Laboratory—Annual Report for
1981 (ANL-82-12), March 1982.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean, Virginia,
1979.
2.1-4
-------
2.2 Brookhaven National Laboratory; Long Island, New York
2.2.1 General Description
Brookhaven National Laboratory (BNL) is a large scientific
research facility located near the center of Long Island approximately
113 kilometers east of New York City. BNL was originally established
as a nuclear science research center but has been expanded to include
facilities for non-nuclear energy studies and environmental research.
Current activities at Brookhaven deal with the transmission, use, and
environmental effects of nuclear and nonnuclear energy sources;
physical, chemical, and biological radiation studies; and applied
nuclear studies, such as those dealing with radioisotopes.
2.2.2 Process Description
A wide variety of scientific programs are conducted at
Brookhaven. The major facilities at the laboratory to carry out the
research and development programs include several accelerators,
reactors, and groups of laboratories. The major facilities at the
laboratory that release radioactivity to the atmosphere are described
briefly below.
The High-Flux Beam Reactor (HFBR) is a 40-MW, fully enriched,
heavy-water-moderated, -cooled, and -reflected reactor. It provides
intense neutron beams for research. The core is contained in an
aluminum vessel and operated at a pressure of 14.1 kg/cm^. The
reactor, its auxiliary equipment, and its experimental facilities are
housed in a welded steel hemisphere 54 meters in diameter. The reactor
cover gas is helium, contaminated with air, fission products, D£0
decomposition products, D£0 vapor, and tritiated heavy water vapor
(DTO). Modifications to the primary water cooling system have been
made to allow the power level to be raised to 60 MW(th).
The Alternating Gradient Synchrotron (ACS) is a 33-GeV proton
accelerator used for ultra-high energy particle physics research.
Protons originate in a 0.75 MeV Cockcroft-Walton generator, are
accelerated by a 200-MeV linear accelerator (linac) and injected into
the ACS. The proton beam may be deflected to strike a target or into
one of the several experimental areas.
The 200-MeV linac also serves the Brookhaven Linac Isotope
Production Facility (BLIP or BLIP) and the Chemistry Linac Irradiation
Facility (CLIF). The BLIP was built to utilize the excess capacity of
the linac to produce significant quantities of radionuclides that can
be made in no other way. The principal component of the BLIP is a
10-meter deep, 2.4-meter diameter water-filled tank, into the bottom of
which the 200-MeV proton beam is directed horizontally. The targets
2.2-1
-------
are individually jacketed and lowered to the 20-centimeter diameter
irradiation chamber through J-shaped tubes. The CLIP, which is
operated in a similar way, provides convenient irradiation of targets
with surplus protons from the linac or secondary neutrons generated by
a converter beam stop. CLIF targets are shuttled into the beam via a
pneumatic target transfer system.
The tandem Van de Graaff accelerator consists of two electrostatic
accelerators capable of independent or tandem use. Maximum achievable
particle energy is 30 MeV. Particles ranging from hydrogen (light) to
chlorine (heavy) have been accelerated. During accelerator operation,
the particle beams are magnetically directed to various targets for
study of nuclear and atomic reactions.
The Brookhaven Medical Research Reactor (BMRR) is a tank-type,.
fully enriched, water-cooled and -moderated reactor. It operates
intermittently at power levels up to 3 MW (thermal) at a pressure of
246 g/cm2. it is an integral part of the Medical Research Center and
is used for various research programs requiring irradiation facilities.
BNL has several laboratories, one of which is the Hot Laboratory
Complex. The Hot Lab originally provided shielded areas for research
and development work with large amounts of radioactive material. It
includes three remotely operable hot cells, a large radioactive metals
hot cell, and several totally sealed systems for use with alpha-
emitting materials. Post-irradiation processing of BLIP targets is
done in one corner of the building. Liquid wastes generated within the
Hot Lab are pumped to storage tanks and evaporated to a slurry. The
distillate flows to the sanitary sewer, and the slurry is packaged at
the Waste Management Facility and shipped as solid waste for offsite
disposal.
Additional programs involving irradiations and/or the use of
radionuclides for scientific investigations are carried on at other
Laboratory facilities including the Biology Department, the Chemistry
Department, and the Department of Energy and Environment.
2.2.3 Radionuclide Emissions
Most of the airborne radioactive effluents at Brookhaven originate
from the HFBR, BLIP, and the research Van de Graaff, with lesser
contributions from the Chemistry and Medical Research Centers.
Radioactive releases occurred during 1981 from the seven stacks that
are identified in Table 2.2-1. The quantities discharged to the
atmosphere are listed in Table 2.2-2. Tritium is the most frequently
discharged contaminant, although oxygen-15 (t^/2 = 122 sec) is
discharged in greatest quantity. About 63 percent of the tritium is
released from the Van de Graaff stack (S-2). The BLIP stack
(S-7) contributes all of the oxygen-15, while the HFBR stack (S-3)
contributes 37 percent of the tritium and all xenon-127 and
2.2-2
-------
unidentified beta, gamma-ray releases. Thus, only small quantities of
radionuclides are released from the other four sources.
Table 2.2-1. The stacks at BNL from which radionuclides
were released during 1981
Stack Number
S-l
S-2
S-3
S-4
S-5
S-6
S-7
Location
Chemistry Building-555
Van de Graaff Ace., Building-901
HFBR - Hot Laboratory
Hazardous Waste Management Area
MRC, Building-490
MRR, Building-491
BLIP, Building-931
Height
(m)
UK
-------
The Brookhaven site covers approximately 21.3 square kilometers.
However, all airborne radioactive releases from the site, excluding
those from the Hazardous Waste Management Area, are located in an area
that is only slightly greater than 1 square kilometer. Because only
very small quantities of radioactivity are discharged from the 10 m
incinerator stack (S-4) in the Hazardous Waste Management Area (See
Table 2.2-2), it was decided to assess the Brookhaven Facility as
having only one airborne radioactive release point: a stack positioned
approximately central to the other six effluent stacks (S-l to S-3 and
S-5 to S-7).
As discussed above, nearly all effluents are released from three
stacks, S-2, S-3, and S-7, that have heights of 18 m, 98 m, and 46 m,
respectively (See Table 2.2-1). To be conservative, 18 m was selected
as the height of the hypothetical stack representing the point source
of airborne discharge. Table 2.2-3 compares the radionuclide emissions
for 1979 to 1981.
Table 2.2-3. Radionuclide emissions (Ci/y) from
Brookhaven National Laboratory, 1979 to 1981
Radionuclide
Argon-41
Beryllium- 7
Garb on- 14
Iodine-125
Iron-59
Oxygen- 15
Phosphorus-32
Sulfur-35
Tin-113
Tritium
Unidentified
beta + gamma
Xenon- 12 7
1979
3.2E+2
NR
NR
NR
2.8E+4
NR
NR
NR
2.3E+2
1.7E-4
1.0
1980
2.6E+2
NR
NR
NR
NR
2.6E+4
NR
NR
NR
5.5E+2
7.8E-5
1.6
1981
1.7E+2
2.6E-3
8.2E-4
9.9E-4
2.5E-4
3.6E+4
1.5E-4
5.7E-3
2.6E-4
6.6E+2
1.8E-4
2.3
(a)
NR - none reported.
2.2-4
-------
2.2.4 Health Impact Assessment of Brookhaven National Laboratory
The health impact assessment for this facility is summarized in Table
2.2-4. The assessment was based on all emissions being combined into one
point source. The individual receiving the maximum exposure resided 1300
meters north-northwest from the hypothetical 18 m stack. The population
within the 80 km radius assessment area is about 4.6 million.
The dose equivalents to the various organs fell within a narrow
range, 0.41 to 0.56 mrem/y, with the endosteal cells and red marrow
receiving the larger dose equivalents. About 94 percent of the dose was
due to oxygen-15 through the air immersion pathway, and tritium
contributed over 5 percent of the dose, mainly through the ingestion
pathway. The collective dose equivalents to the various organs were
relatively uniform, ranging from 2.9 to 3.3 person-rem/y. The exposure to
the regional population was primarily due to tritium (76 percent) and
oxygen-15 (22 percent).
The risk of having fatal cancer as a result of exposure to the
radioactive emissions from this facility are listed in Table 2.2-5. The
highest individual lifetime risk is 9E-6, while the risk within the
regional population for the combined sources is 9E-4 fatal cancers each
year of facility operation.
Table 2.2-4. Radiation dose rates from radionuclide emissions
from Brookhaven National Laboratory
Maximum individual Regional population
0rSan (mrem/y) (person-rem/y)
Pulmonary 4.4E-1 3.1
Red marrow 5.4E-1 3.3
Muscle 4.7E-1 3.2
Liver 4.1E-1 3.0
Endosteal 5.6E-1 2.9
Weighted sum 4.4E-1 3.2
2.2-5
-------
Table 2.2-5. Fatal cancer risks due to radionuclide emissions from
Brookhaven National Laboratory
0 Lifetime risk Regional population
Source , a , K .
to maximum individual (Fatal cancers/y of operation;
BNL 9E-6 9E-4
2.2-6
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
Na82 Naidu J.R. and Olmer L.L., Editors, 1981 Environmental
Monitoring Report, Brookhaven National Laboratory, Safety and
Environmental Protection Division, April 1982.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, In., McLean, Virginia,
1979.
2.2-7
-------
2*3 Fermi National Acc^Urat£rJ1aboratory; Batavia, Illinois
2.3.1 General Description
The Fermi National Accelerator Laboratory (FNAL) is located in the
Greater Chicago area just east of Batavia, Illinois, on a 27.5 km2
tract of land. The site is roughly 4.8 km square and is operated for
the Department of Energy (DOE) by Universities Research Association.
The facility is composed of three basic elements: the accelerator,
experimental areas, and support facilities.
The primary purpose of FNAL is fundamental research in high energy
physics. In addition, cancer patients are being treated using neutrons
released by the interaction of 66 MeV protons from the second stage of
the accelerator. A major program is in progress to construct, install,
and operate a ring of superconducting magnets. The goal is to produce
higher energy protons using less electrical power.
The surrounding area is rapidly changing from farming to
residential use. There are many manicipalities in the vicinity,
resulting in a distinct pattern of high population concentration.
Within a 3-km distance from the Laboratory boundaries, Batavia (pop.
12,169), Warrenville (pop. 7,185), and West Chicago (pop. 12,444), are
located. The total population within a 80 km radius of FNAL is more
than 7.5 million.
2.3.2 Process Description
The FNAL is a proton synchrotron with an original design energy of
200 GeV (billion electron volts). As a result of accelerator
improvements, protons were accelerated to an energy of 500 GeV in 1976
and operation at 400 GeV is now routine.
The proton beam extracted for high energy physics from the 2-km
diameter main accelerator is taken to three different experimental
areas on site, the Meson, Neutrino, and Proton Areas. All three areas
received proton beams for the first time in 1972. Radioactivity is
produced as a result of the interaction of the accelerated protons with
matter. The total number of protons accelerated in 1981 was 1.4 X
1019.
2.3.3 Radionuclide Emissions and Control Technology
Radioactivation of air in measurable concentrations will occur
wherever the proton beam or the spray of secondary particles resulting
from its interactions with matter passes through the air. Along most
proton beam lines (paths of the protons from the accelerator), the
protons travel inside evacuated pipes. Thus, radioactivation of air is
usually caused by secondary particles.
2.3-1
-------
Radioactive gas, primarily carbon-11, was produced by interaction
of secondary particles with air. Monitoring was carried out by
detecting the beta particles emitted in the radioactive carbon-11
decay. A release of 1.45 kCi occurred from the labyrinth stack in the
Neutrino Area during 1981.
There was also a controlled release of tritium in tritiated water
evaporated as a means of disposal for the first time at Fermilab in
1981. The total quantity released to the atmosphere was 420 mCi. The
release occurred from the Meson Area.
A debonding oven was placed in operation in 1979. Its purpose is
to debond magnets by decomposing the epoxy adhesives at high
temperatures. Most of these magnets are radioactive, having failed
during accelerator operations. Thirty magnets were debonded in 1981,
and the total tritium release was approximately 5 mCi. Table 2.3-1
list the activity, location, and stack heights of the FNAL airborne
releases for 1981. Table 2.3-2 summarizes the airborne releases from
1979 to 1981. The primary control of airborne radioactive emissions is
hold-up confinement. The accelerator is designed for high efficiency,
so that proton losses will be small during acceleration, extraction,
and transport to the experimental-area targets.
The accelerator, beam-transport, and target systems are all within
well-shielded housings, while the beam travels in evacuated pipes, thus
reducing the radioactivation of air.
2.3.4 Health Impact Assessment of Fermi Laboratories
The estimated annual radiation doses resulting from radionuclide
emissions from the Fermi Laboratories are listed in Table 2.3-3. The
maximum individual is located 1300 meters north of the release
location. The predominant exposure pathway is that of air immersion.
The dose is primarily (greater than 99%) from carbon-11.
Table 2.3-4 list the estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses. The lifetime risk to the maximum individual is estimated
to be 1E-5 and the total number of fatal cancers per year of operation
of the Fermi Laboratory to be 3E-4.
2.3-2
-------
Table 2.3-1. Radionuclide emissions from
Fermi National Accelerator Laboratory, 1981
Source(a)
Neutrino Area
Meson Area
Debonding oven
Radionuclide
Carbon-11
Tritium
Tritium
Emissions
(Ci)
1.5E+3
4.2E-1
5.0E-3
(a'Stack height = 10 meters.
Table 2.3-2. Radionuclide emissions from
Fermi National Accelerator Laboratory, 1981
(Ci/y)
Radionuclide
Carbon-11
Tritium
1979
4.0E+3
2.8E-1
1980
1.3E+3
2.4E-1
1981
1.5E+3
4.2E-1
Table 2.3-3. Radiation dose rates from radionuclide emissions from
Fermi National Accelerator Laboratory, 1981
Organ
Red marrow
Endosteal
Testes
Spleen
Muscle
Weighted Sum
Maximum individual
(mrem/y )
6.7E-1
6. 9E-1
6.5E-1
6.2E-1
5.8E-1
5.4E-1
Regional population
(person-rem/y)
1.4
1.5
1.4
1.3
1.2
1.1
2.3-3
-------
Table 2.3-4. Fatal cancer risks due to radionuclide
emissions from Fermi National Accelerator Laboratory
Lifetime risk to Regional population
Source Maximum individual (Fatal cancers/y of operation)
FNAL 1E-5 3E-4
2.3-4
-------
REFERENCES
Da80 Dave M. J. and Charboneau R., Baseline Air Quality Study at
Fermilab, ANL Report, ANL/EES-TM-110, 1980.
DOE81 Department of Energy, Environmental Monitoring Report for
Fermi National Accelerator Laboratory, Annual Report for CY
1981, FERMILAB 82/22, Universities Research Association Inc.,
Batavia, Illinois, 1981.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, In., McLean, Virginia,
1979.
2.3-5
-------
2-4 Hanford Reservation; Richland. Washington
2.4.1. General Description
The Hanford Reservation is a 1,500 square-kilometer site located
270 kilometers southeast of Seattle, 200 kilometers southwest of
Spokane, Washington, and 230 kilometers east of Mt. St. Helens. The
Columbia River flows through the northern edge of the Hanford site and
forms part of its eastern boundary.
Facilities on the Hanford Reservation include the historic reactor
facilities for plutonium production along the Columbia River in the
100 Area. The reactor fuel-processing and waste-management facilities
are on a plateau about 4.3 kilometers (2.7 miles) from the river in the
200 Area. The 300 Area, just north of the city of Richland, contains
the reactor fuel manufacturing facilities and research and development
laboratories. The Fast Flux Test Facility (FFTF) is located in the
400 Area approximately 3.8 kilometers (2.1 miles) northwest of the
300 Area, and the WPPSS power reactor site is about 4.3 kilometers (2.7
miles) north of the 300 Area.
Privately-owned facilities located within the Hanford Reservation
boundaries include the WPPSS generating station adjacent to N Reactor,
the WPPSS power reactor site and office buildings, and a radioactive
waste burial site. The Exxon fuel fabrication facility is located
immediately adjacent to the southern boundary of the Hanford
Reservation.
The facilities for these programs are located at five operational
areas on the reservation (designated 100, 200 East, 200 West, 300, and
400). These facilities are operated for the Department of Energy by
four prime contractors. The current contractors and their primary
roles are:
- Rockwell International's Rockwell Hanford Operations (RHO):
waste management and fuel processing
- Douglas United Nuclear Industries (UNI): reactor operation and
fuel fabrication
- Battelle's Pacific Northwest Laboratory (PNL): research in
biophysics and biomedicine and development of advanced waste
management technologies
- Westinghouse Hanford Company: operation of Hanford Engineering
Development Laboratory (HEDL) and advanced energy research.
2.4-1
-------
2.4.2 Process Description
The Hanford Reservation was originally established in 1943 to
produce plutonium for nuclear weapons. At one time, nine production
reactors were in operation, including eight with once-through cooling.
Between December 1964 and January 1971, all eight reactors with
once-through cooling were deactivated. N-Reactor, the remaining
production reactor in operation, has a closed primary cooling loop.
Steam from N-Reactor operation is used to drive turbine generators that
produce up to 860 million watts of electrical power in the Washington
Public Power Supply System's (WPPSS) Hanford Generating Plant. By the
end of 1976, N-Reactor had supplied enough steam to produce nearly 35
billion kilowatt-hours of electrical energy, which was fed to the
Bonneville Power Administration grid covering the Pacific Northwest.
Presently, plutonium production has decreased and other programs
have been introduced and developed. Current operations include
plutonium production and fabrication, management and storage of
radioactive wastes, reactor operations and fuel fabrication, energy
research and development, and biophysical and biomedical research.
100 Area
The 100 Area is the location of the original nine plutonium
production reactors in the northern area of the Hanford site
approximately 8 to 10 kilometers from the northern site boundary and
adjacent to the Columbia River. The 100 Area is approximately 45
kilometers north-northwest of Richland. Eight of the reactors have
been deactivated and placed on standby. Operating facilities in the 100
Area include the N-Reactor and the 1706 Laboratory.
The N-Reactor is operated by UNI and is the only plutonium
production reactor still in operation on the Hanford Reservation.
Pacific Northwest Laboratory operates the 1706 Laboratory located
in the 100-K Area. The laboratory conducts studies of water quality,
filtration, and corrosion in support of N-Reactor operations.
Small-scale decontamination studies are also done at the laboratory.
200 Area
The 200 Area is divided into the 200 East Area and the 200 West
Area. The 200 East Area is located in the center of the Hanford site,
approximately 15 kilometers from the east and west site boundaries and
35 kilometers north-northwest of Richland. Activities conducted in
this area include irradiated fuel processing, waste management and
storage, and laboratory research. The 200 West Area is adjacent to the
2.4-2
-------
200 East Area. Activities conducted in the area include waste
treatment and storage, equipment decontamination, plutonium and uranium
processing, and laboratory research.
The PUREX Plant, located in the 200 East Area, is the fuel
reprocessing facility at Hanford. Since 1972 the PUREX Plant has been
held in standby and is scheduled to resume operation no later than
April 1984 and continue through the year 2000. See Section 2.16 for a
discussion of the future operations of DOE facilities.
Another facility in the 200 East Area is the Critical Mass
Laboratory which is operated by PNL. This laboratory is used for
research on the criticality safety of plutonium in its various forms
and combinations with other elements. All of the remaining facilities
in the 200 East Area are used for waste treatment and storage.
Included among these facilities are B-Plant, C-Plant, the AR and CR
vaults, and the numerous tank farms.
Major facilities in the 200 West Area include the U03 plant, the
Z-Plant, and the Redox Plant. Uranyl nitrate hexahydrate solution
(UNH) is converted to U03 at the U03 Plant. The Z-Plant has been
used to finish the processing of plutonium separated during the PUREX
process. Currently, a capability to complete the processing of
plutonium oxide has been added to the PUREX plant; therefore, the
Z-Plant will no longer be used for this purpose. The Z-Plant presently
reclaims plutonium from scrap. The Redox facility currently houses
Laboratories 222-S and 219-S which conduct studies in support of
B-Plant operations and waste management processes.
Support facilities in the 200 West Area include the T-Plant, used
for equipment repair and decontamination projects; the Plutonium
Metallurgy Laboratory, operated by BNL; facility tank farms; the 242T
waste evaporator; and the laundry facility.
300 Area
The 300 Area, which is in the southeast corner of the reservation,
is the site of most of the laboratory and research facilities at
Hanford. This area is 8 kilometers north of Richland and adjacent to
the east site boundary. The major facilities are the Hanford
Engineering Development Laboratory (HEDL), the fuel fabrication
facility, and the Life Sciences Laboratory.
The Hanford Engineering Development Laboratory is the major
facility in the 300 Area. It consists of numerous laboratories,
testing facilities, and storage areas utilized in support of the Fast
Breeder Reactor (FBR) program at Hanford. These facilities are
operated by Westinghouse Hanford Company for the Department of Energy.
The fuel fabrication facility is operated by UNI. It is used in
the production of fuel pins for the N-Reactor. The Life Sciences
Laboratory is operated by PNL; current programs include biophysical and
2.4-3
-------
biomedical research. Studies on the inhalation of plutonium which were
formerly conducted in the 100 areas were transferred to this facility
in 1975, In addition, BNL operates two laboratories that conduct
research in advanced waste management techniques and metallurgical
techniques. These laboratories are the Metal Fabrication Laboratory
and the 3720 Laboratory.
Previous programs at Hanford generated radioactive wastes which
were buried in the 300 Area. These areas are not presently in use, and
radioactive wastes that are being generated by current programs are
shipped to the 200 Areas for processing and disposal. No airborne
effluents are released from the buried wastes.
400Area
The 400 Area is the newest of the operational areas to be
developed at Hanford. The area is approximately 9 kilometers northwest
of the 300 Area and 5 kilometers from the south and east site
boundary. At present, the Fast Flux Test Facility (FFTF) is in
operation in the 400 Area and the Fuel Materials Examination Facility
(FMEF) is under construction in the 400 Area. When these facilities
are both in operation, the 400 Area will be the center for the fast
breeder reactor development program at Hanford.
2.4.3 Radionuclide Emissions and Control Technology
The airborne releases at Hanford Reservation are presented in
Table 2.4-1. The site is large, covering an area of 1,500 square
kilometers. For the purposes of analysis, Hanford is regarded as
having three point sources for emissions, each at a height of 1 m above
the surface. These are located in the 100 Area, 200 Area, and the
combined 300-400 Area. The release point in the 100 Area is 8
kilometers from the northern site boundary at the location of
N-reactor. The 200 Area stack is 10 kilometers from the southern site
boundary and is located at a point midway between 200 east and 200 west
Areas. The 300-400 Area release point is 0.25 kilometers from the
southern boundary.
Existing Control Technology
All particulates released from Hanford operations are less than 1
micron in size. Airborne effluents from the N Reactor constitute more
than 95 percent of the releases in the 100 Area. Releases from the
N-Reactor are passed through HEPA filters and activated charcoal
filters, while emissions from the 1706 Laboratory are exhausted through
HEPA filters only.
In the 200 Area, residual operations presently occurring at the
PUREX Plant account for the majority of the plutonium released in the
area. Airborne effluents from all 200 Area release points are passed
2.4-4
-------
through acid scrubbers, deentrainers, fiberglass filters, and HEPA
filters prior to release. In addition, releases from the PUKEX plant
are passed through a silver nitrate reactor to remove elemental
iodine. Emissions from all waste management functions in the 200 East
Area account for the significant release of beta- and gamma-emitting
nuclides and one-third of the plutonium emissions.
In the 200 West Area, emissions from the Z-Plant include 70
percent of the area beta-gamma releases. These releases are filtered
through either multilayered sand filters or HEPA filters. In addition,
80 percent of the plutonium from the U-Plant (adjacent to the 1103
Plant) is released untreated. Discharges of plutonium-239 from Z-Plant
represent more than 80 percent of the total plutonium released in the
area. All of the release points at the Z-Plant are fitted with one,
two, or three HEPA filters to control particulate emissions.
In the 300 Area, the fuel fabrication facility is responsible for
most of of the natural uranium discharged in the area. All discharges
pass through HEPA filtration prior to release.
2.4.4 Health Impact Assessment of the Hanford Site
A separate health risk assessment was performed for each of the
three sources considered at this site. Summaries of these analyses are
given in Table 2.4-2 and in Table 2.4-3. The size of the regional
population differs for each source (266,000 for the 100 Area, 259,000
for the 200 Area, and 199,000 for the 300-400 Area). The maximum
individual in the 100 Area is 7500 m northwest of the source. For the
200 Area, the maximum individual is 16,000 m south of the release
point. The maximum individual in the 300-400 Area is also south of the
facility, although the distance is 2000 meters.
The lifetime fatal cancer risk to the maximum individual ranges
from 3E-7 to 4E-5 or 3 in 10 million to 4 in 100,000 (see Table
2.4-3). The lowest individual risk results from exposure to emissions
from the 200 Area, while the highest risk is associated with the 100
Area. The number of fatal cancers expected per year within the
regional population varies from 6E-5 at the 200 Area to 2E-3 at the 100
Area.
Organs receiving the five highest dose equivalent rates from
emissions from the 100 Area range from 2.2 mrem/y to the pulmonary
region to 2.4 mrem/y to the endosteal and spleen cells. Argon-41
contributed 92 percent of the weighted sum organ dose equivalent and 93
percent of the fatal cancer risk.
In the 200 Area, the five organs receiving the highest dose
equivalent rate ranged from 1.2E-2 mrem/y to testes to 8.4E-2 to
endosteal cells. Barium-137m (from decay of cesium-137) contributed
47 percent of the weighted sum organ dose equivalent and 63 percent of
the fatal cancer risk.
2.4-5
-------
For the 300-400 Areas, the organ dose equivalent rate for the five
highest organs ranged from 1.4 mrem/y to the thymus to 1.5 mrem/y to
the endosteal cells. The radionuclide of greatest significance in the
area was krypton-88 which contributed 95 percent of the weighted sum
dose equivalent rate and 99 percent of the fatal cancer risk.
The pathway of greatest significance dosimetrically from the
100 Area is air immersion which contributes 2.1 of the 2.1 mrem/y
weighted sum dose. For the 200 Area, the majority of the 1.7E-2 mrem/y
weighted sum dose equivalent rate is divided almost equally between
ground surface (8.0E-3 mrem/y) and the inhalation pathways (6.1E-3
mrem/y). For the 300-400 Area, the maximum pathway is air immersion
which contributes 1.2 mrem/y of the weighted sum dose equivalent rate
of 1.3 mrem/y.
2.4-6
-------
Table 2.4-1. Radionuclide emissions from the Hanford Reservation, 1981
Emissions (Ci/y)
Radionuclides
Argon -41
Arsenic-76
Carbon-14
Barium-Lanthanum-140
Cerium-144
Cobalt-58
Cobalt-60
Cesium-137
Cesium-138
Europium-154
Europium-155
Iron-59
Tritium
Iodine-131
Iodine-132
Iodine-133
Iodine-135
Kryp ton-85m
Krypton-87
Krypton-88
Manganese-54
Manganese -56
Sodium-24
Plutonium-239
Ruthenium-103
Ru thenium-Rhod ium-1 06
Strontium-89
Strontium-90
Strontium-91
Molybdenum-Technetium-99m
Uranium-234
Uranium-238
Xenon-135
100 Area 200 Area
6.5E+4
2.3E-2
3.2
1.1E-1
7 . 9E-2
6.6E-3
1.6E-2
8.9E-3 5.0E-2
1 . 1E+4
1.5E-1
2.5E-2
2.7E-3
1 . 8E+1
9.7E-2
4.3
9.4E-1
1.6
2.5E+2
2.8E+2
5.4E+2
2.8E-3
4.6E-1
1.2E-1
6.4E-5 3.7E-4
3.3E-3
4.2E-3
1 . 5E-3
4.8E-3 3. IE -3
1.8E-1
2.5E-1
4.6E+2
300-400 Area
4.5E-7
3.3E-7
3.0E-4
4.5E+2
2.2E-5
8.8E-5
7.5E-5
7.5E-5
2.4-7
-------
Table 2.4-2. Radiation dose rates from radionuclide emissions from
the Hanford Reservation, 1981
Maximum individual (mrem/y)
rgan 100 Area 200 Area 300-400 Area
Testes
Liver
Red marrow
Endosteal
Pulmonary
Pancreas
Upper large intestine
Thymus
Muscle
Spleen
2.3
2.4
2.2
2.2
2.4
1.2E-2
2.2E-2
2.0E-2
8.4E-2
2.1E-2
1.5
1.4
1.4
1.5
1.4
Weighted sum dose
equivalent rate 2.1 1.7E-2 1.3
- Regional population
r8an (person-rem/y)
Red marrow
Endosteal
Pulmonary
Upper large intestine
Muscle
11.1
13.3
10.7
10.7
10.4
2.4-8
-------
Table 2.4-3. Fatal cancer risks due to radionuclide emissions from
the Hanford Reservation, 1981
Source
100 Area
200 Area
300-400 Area
Lifetime risk
to maximum individual
4E-5
3E-7
3E-5
Regional population
(Fatal cancers/y of operation)
2E-3
6E-5
1E-3
2.4-9
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
DOE82 Department of Energy, Summary of Annual Environmental Reports
for CY 1980, DOE/EP-0038, 1982.
ERDA75 Energy Research and Development Administration, Final
Environmental Impact Statement, Waste Management Operations,
Hanford Reservation, Richland, Washington, ERDA-1538, UC-70,
Volumes 1 and 2, Washington, D.C., 1975.
ERDA77 Energy Research and Development Administration, Final
Environmental Impact Statement, High Performance Fuel
Laboratory, Hanford Reservation, Richland, Washington,
ERDA-1550, UC-2, 11, Washington, D.C., 1977.
Su82 Sula M. J., McCormack, Dirkes R. L., Price K. R., Eddy P. A.,
Environmental Survillance at Hanford for CY-81, PNL-4211, May
1982.
2.4-10
-------
2.5 Idaho National Engineering Laboratory; Upper Snake River Plain
2.5.1 General Description
The Idaho National Engineering Laboratory (INEL) is a large
reactor testing facility located in southeastern Idaho. INEL was
established in 1949 (then called the National Reactor Testing Station)
to provide an isolated station where various kinds of nuclear reactors
and support facilities could be built and tested. The site encompasses
2,314 square kilometers and is situated 35 kilometers west of Idaho
Falls and 37 kilometers northwest of Blackfoot. As of 1976, 51
reactors had been built, 16 of which were still operating or
operational.
Current programs at INEL are conducted at various areas of the
site and are managed for DOE by four contractors: EG&G Idaho, Inc;
Allied Chemical Corporation; Argonne National Laboratory; and
Westinghouse Electric Corporation.
EG&G Idaho, Inc., operates the Power Burst Facility located in the
Special Power Excursion Reactor Test Area (SPERT); the Advanced Test
Reactor and the Engineering Test Reactor, located in the Test Reactor
Area (TRA); the Technical Support Facility (TSF), located in the Test
Area North (TAN); and the Hot Cell located in the Auxiliary Reactor
Area (ARA-1). Programs that require the use of these facilities
include test irradiation services from the two operating high-flux
reactors and light-water-cooled reactor safety testing and research.
Allied Chemical Corporation operates the Idaho Chemical Processing
Plant. One of the activities performed here is the recovery of uranium
from highly enriched spent fuels. Argonne National Laboratory-West
(ANL-W) operates the Experimental Breeder Reactor No. 2 and related
support facilities. Westinghouse Electric Corporation operates the
Naval Reactor Facility at INEL.
2.5.2 Process Description
EG&G Facilities
The Power Burst Facility (PBF) is a high-performance, water-
cooled, uranium-fueled reactor, designed to operate at powers of up to
40 megawatts for time intervals up to 48 hours. The facility is used
to provide operating information in support of DOE's light-water
reactor safety program.
The Test Reactor Area (TRA) contains six reactors (three test
reactors and three low-power reactors). Of the three test reactors,
only two are operating: the Advanced Test Reactor (ATR) and the
Engineering Test Reactor (ETR). The third, the Materials Testing
Reactor (MTR), was placed on standby in 1970. The ATR and the ETR
2.5-1
-------
facilities provide research data on the performance of reactor
materials and equipment components under conditions of high neutron
flux. This research is in support of DOE's reactor development
program. Also, the facilities at TRA have occasionally been made
available to private organizations and other government agencies for
research purposes.
TSF, part of TAN, is used in a support role for materials
examination and repair, fabrication and assembly of the Loss of Fluid
Test (LOFT) Mobile Test Assembly, and various reactor safety studies.
Remote disassembly and reassembly of large radioactive components are
performed in the Hot Shop Area. Activities in the Warm Shop at TSF are
limited to the handling of only slightly radioactive materials.
Auxiliary Reactor Area-1 (ARA-1) is presently used for the
operation of research and laboratory facilities and a Hot Cell. The
Hot Cell is used to prepare test specimens for use in the various INEL
reactors.
The Radioactive Waste Management Complex (RWMC) is one of the
three principal waste handling facilities at INEL (the other two are
the ANL-W Radioactive Scrap and Waste Facility and the Idaho Chemical
Processing Plant). Waste from INEL and other DOE facilities, such as
Rocky Flats, is packaged and stored at RWMC.
Allied Chemical Corporation, Idaho Chemical Processing Plant
The three major activities at the Idaho Chemical Processing Plant
(ICPP) are irradiated fuel storage, fuel reprocessing, and waste
calcination. Spent fuel from INEL reactors and other domestic and
foreign research reactors is either stored at ICPP or converted to
uranium oxide powder and shipped to Oak Ridge National Laboratory
(ORNL) or Portsmouth. In addition, the ICPP contains the Waste
Calcining Facility (WCF), which is used to convert high-level
radioactive liquid waste to solid form.
Argonne National Laboratory-West Facilities
The Argonne National Laboratory-West (ANL-W) currently has five
operational complexes: the Experimental Breeder Reactor No. 2
(EBR-II), the Transient Reactor Test Facility (TREAT), the Zero Power
Plutonium Reactor (ZPPR), the Hot Fuels Examination Facility (HFEF),
and the Laboratory and Office (L&O) support complex. All of these
complexes provide support services for DOE's Fast Breeder Reactor (FBR)
research program.
Westinghouse Electric Corporation
The Naval Reactor Facility (NRF), located 22 kilometers west and
north of the ANL-W area, is operated by Westinghouse Electric
Corporation. The facility serves as a testing area for prototype naval
reactors and as a disassembly and inspection area for expended reactor
2.5-2
-------
cores. The prototype reactors are also used as training centers for
naval reactor operators. Three operating reactors and the Expended
Gore Facility (ECF) are located in this area. These include the Large
Ship Reactor (A1W), the Submarine Thermal Reactor (S1W), and the
Natural Circulation Reactor (S5G).
2.5.3 Radionuclide Emissions Measurements and Control Technology
Measurements of airborne releases at INEL have been consolidated
and are presented in Table 2.5-1. The majority of emissions are
attributable to the operation of the ATR and the ETR in the Test
Reactor Area. These releases include argon-41, a majority of reported
isotopes of xenon, cesium-138, barium-139, krypton-85, krypton-85m,
krypton-87, and rubidium-88. TREAT accounts for the xenon-133
emissions, and activities at ICPP are responsible for exhausting
tritium and krypton-85. EBR-II releases 50 percent of the total site
xenon-135 emissions.
Releases from the ETR and ATR facilities are not treated. Other
facilities at INEL, however, use multiple or single HEPA filters and,
occasionally, charcoal absorbers. Areas using such control
technologies include ZPPR, TREAT, NRF facilities, PBF, and ARA-1.
2.5.4 Health Impact Assessment of Idaho National Engineering Laboratory
For the purpose of the dose/health effects assessment, it is
assumed that all particulates released are 1 micron or less in
diameter, and the entire site release is respirable. The assessment is
based on all emissions being combined into one point source midway
between the TRA and GPP areas at a height of 1 meter above the ground.
Actual site boundary distances from the assumed point source were used
in the calculations.
Radiation dose rates are given in Table 2.5-2. The individual
receiving the highest weighted sum dose equivalent rate is located
19500 m north of the assumed release point. The maximum individual
lifetime fatal cancer risk is 5E-7. Air immersion is the major pathway
contributing to the individual dose equivalent rate (68 percent).
The estimated fatal cancer risk to the regional population per
year of operation is 5E-5 (Table 2.5-3). The pathway contributing
primarily to the fatal cancer risk was ingestion. The collective
weighted sum dose equivalent rate is 0.38 person-rems per year.
2.5-3
-------
Table 2.5-1. Radionuclide emissions from Idaho
National Engineering Laboratory, 1981 (DOE81)
Radionuclide
Silver-HOm
Argon-41
Barium-131
Barium- 13 9
Barium-Lanthanum-140
Beryl lium-7
Bromine-82
Carbon- 14
Cerium- 141
Cerium- 144
Cobalt-57
Cobalt-58
Cobalt-60
Cesium-134
Cesium-137
Cesium-138
Chromium-51
Europium-152
Europium-154
Europium-155
Tritium-3
Hafnium-181
Iodine-129
Iodine-131
Iodine-133
Krypton-85
Krypton-85m
Krypton-87
Krypton-Rub id ium-88
Manganese-54
Niobium-95
Promethium-144
Plutonium-238
Plutonium-239
Ruthenium-103
Emissions
(Ci/y)
8.5E-7
2.5E+3
2.2E-9
1.6E+2
3.4E-5
1.3E-5
9.0E-1
1.7E-1
1.7E-6
3.9E-4
1.6E-8
3.6E-5
2.3E-4
6.0E-5
8.6E-3
1.7E+1
2.8E-5
6.0E-7
7.7E-6
1.5E-6
4.0E+2
1.1E-5
3.7E-2
5.5E-2
2.0E-6
5.9E+4
2.2E+2
8.7E+2
8.0E+2
7.3E-6
2.5E-5
3.7E-4
7.4E-5
1.8E-5
1.4E-6
2.5-4
-------
Table 2.5-1. Radionuclide emissions from Idaho
National Engineering Laboratory, 1981 (DOE81)
Radionuclide Emissions
(Ci/y)
Ruthenium-Rhodium-106 7.7E-2
Antimony-122 1.2E-7
Antimony-125 1.9E-1
Strontium-90 4.1E-3
Tantalum-182 1.9E-7
Tellurium-132 1.6E-7
Technetium-99m l.OE-4
Tin-113 1.8E-7
Xenon-133 1.6E+2
Xenon-135 8.0E+2
Xenon-135m 4.2E+2
Xenon-138 2.5E+3
Zirconium-95 1.9E-6
2.5-5
-------
Table 2.5-2. Radiation dose rates from radionuclide emissions
from Idaho National Engineering Laboratory
Maximum individual Regional population
rSan (mrem/y) (person-rem/y)
Pulmonary
Endosteal
Red marrow
Upper large intestine
Muscle
Thyroid
3.1E-2
3.1E-2
2.6E-2
2.5E-2
2.4E-2
1.2E-1
1 . 7E-1
2.6E-1
1.9E-1
1.9E-1
1.4E-1
5.5
Weighted dose equivalent 2.9E-2 3.8E-1
Table 2.5-3. Fatal cancer risks due to radionuclide emissions from
Idaho National Engineering Laboratory
s Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
INEL 5E-7 5E-5
2.5-6
-------
REFERENCES
DOE81
DOE82a
DOE82b
ERDA77a
ERDA77b
TRI79
Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
Department of Energy, Idaho Operations Office, 1981
Environmental Monitoring Program Report for Idaho National
Engineering Laboratory Site IDO-12082 (81), May 1982.
Department of Energy, Summary of Annual Environmental Reports
for CY1980, DOE/EP-0038, 1982.
Energy Research and Development Administration, Final
Environmental Impact Statement, Waste Management Operations,
Idaho National Engineering Laboratory, Idaho, ERDA-1536,
Washington, D.C., September 1977.
Energy Research and Development Administration, Environmental
Monitoring at Major U.S. Energy Research and Development
Administration Contractor Sites, Calendar Year 1976, Volumes 1
& 2, ERDA 77-104/1 & /2, Washington, D.C., 1977.
Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, In., McLean, Virginia,
1979.
2.5-7
-------
2.6 Lawrence Livermore National LaborafrO£yj_J1ivermore> California
2.6.1 General Description
The Lawrence Livermore National Laboratory is located about 64
kilometers east of San Francisco, California, in the Livermore Valley
of eastern Alameda County, approximately 5 kilometers east of the City
of Livermore. The site covers an area of 2.54 km2 and is surrounded
by open agricultural areas on the north, east, west, and part of the
south side. Sandia Laboratories, Livermore, is located on adjoining
property to the south. Materials testing and high-explosives
diagnostic work is conducted at a remote site, Site 300, located on a
27 km2 site 16 kilometers southeast of Livermore.
In addition to its primary role of .nuclear weapons research and
development, Lawrence Livermore National Laboratory conducts research
programs in the areas of magnetic fusion, nonnuclear energy, laser
fusion, laser isotope separation and biomedical research.
2.6.2 Process Description
There are five principal facilities that release radioactivity
into the air at Lawrence Livermore Laboratory.
Light Isotope Handling Facility (Building 331)
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.
Insulated Core Transfer Accelerator (ICT) (Building 212)
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) (Building 194)
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.
2.6-1
-------
Decontamination Facility (Building 419)
HEPA filters are used to reduce particulate radioactivity from
exhaust air. The radioactivity in air effluents originate^ from
various decontamination operations. Stack effluents are conteinuously
sampled.
Solid Waste Disposal Facility (Building 612)
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.
2.6.3 Radionuclide Emissions
Table 2.6-1 identifies radioactive emissions from the facilities
at Lawrence Livermore Laboratory in 1981. For the purpose of this
analysis, the total emissions are assumed to be released from
Building 194 from a 30-meter stack. Radioactive emissions for the
period 1979 to 1981 are shown in Table 2.6-2.
2.6.4 Health Impact Assessment of Lawrence Livermore National Laboratory
The estimated annual radiation doses resulting from radionuclide
emissions from Lawrence Livermore National Laboratory are listed in
Table 2.6-3. The maximum individual is located 590 meters
east-northeast of the assumed release point (Building 194). The
predominant exposure pathway is ingestion and primarily from tritium.
The total population within an 80-km radius of the site is 4.6 million.
(
Table 2.6-4 shows the estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses. The lifetime risk to the maximum individual is estimated
to be 3E-5 and the total number of fatal cancers per year of operation
is 2E-3.
2.6-2
-------
Table 2.6-1. Radionuclide emissions from
Lawrence Livermore National Laboratory, 1981
(Ci/y)
Tritium 2.6E+3
Nitrogen- 13
Oxygen-15
Plutonium-239(a)
Strontium-90(b)
Building
292 212 194
4.4E+1 2.3E+1
1.7E+2
1.7E+2
4.2E-6
5.5E-5
419 Totals
2.6E+3
1.7E+2
1.7E+2
9.0E-7 5.1E-6
1.7E-5 7.2E-5
(a)Reported as "Unidentified Alpha."
(^Reported as "Unidentified Beta + Gamma."
Table 2.6-2. Radionuclide emissions from the
Lawrence Livermore National Laboratory,
1979 to 1981 (Ci/y)
Radionuclide
1979
1980
1981
Argon-41
Tritium
Nitrogen-13
Oxygen-15
Plutonium-239(a)
Strontium-90(b)
3.8E+2
4.5E+3
5.0E+2
3.3E+2
7.2E-10
6.0E-5
1 . 6E+2
2.3E+3
9. 9E+2
6.6E+2
NR
4.7E-5
NR
2.6E+3
1.7E+2
1.7E+2
5.1E-6
7.2E-5
(^Reported as "Unidentified Alpha."
(^Reported as "Unidentified Beta + Gamma."
NR=None reported.
2.6-3
-------
Table 2.6-3. Radiation dose rates from radionuclide emissions
from the Lawrence Livermore National Laboratory
_ Maximum individual Regional population
Organ / i \ t i \
(mrem/y; vperson-rem/y;
Lower large
Upper large
Stomach
intestine
intestine
1.
1.
1.
9
5
5
Small intestine
Red marrow
Kidneys
1.
1.
3
3
8.
7.
6.
6.
5.
7
0
9
0
8
Weighted sum 1.3 5.7
Table 2.6-4. Fatal cancer risks due to radioactive
emissions from the Lawrence Livermore National Laboratory
g Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
LLNL 3E-5 2E-3
2.6-4
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, In., McLean, Virginia,
1979.
UCRL82 University of California, Environmental Monitoring at the
Lawrence Livermore National Laboratory - 1981 Annual Report,
Publication No. UCRL-50027-81, University of California,
Livermore, California, 1982.
2.6-5
-------
2.7 Los Alamos National Laboratory; Los Alamos, New Mexico
2.7.1 General Description
The Los Alamos National Laboratory (LAND is a multidisciplinary
facility located in north-central New Mexico. The site is about 100
kilometers north-northeast of Albuquerque and 40 kilometers northwest
of Santa Fe. LANL is one of the prime research and development
facilities in DOE's nuclear weapons program. In addition to national
defense programs, activities at Los Alamos include research in the
physical sciences, energy resources (both nuclear and nonnuclear) and
applied programs, and biomedical and environmental studies. Facilities
for these programs are dispersed widely over the site which is
separated into a number of technical areas (TAs).
A substantial portion of LANL's reported emissions may be
attributed to operations at the Meson Physics Facility (TA-53), the
South Mesa Site (TA-3), the Omega Site (TA-2), and several other
technical areas. Programs at these sites include the operation of an
800 MeV proton accelerator, laser and magnetic fusion activities, the
operation of two research reactors—one of which is a 10-MW reactor—at
the Omega site, and experiments using a tandem Van de Graaff
accelerator.
2.7.2 Process Description
During 1981, effluents were released from more than 75 stacks
located in 13 Technical Areas. A brief description of the activities
conducted in these areas follows.
TA-2, Omega Site
Omega West Reactor, an 8 megawatt nuclear research reactor, is
located here. It serves as a research tool in providing a source of
neutrons for fundamental studies in nuclear physics and associated
fields.
TA-3, South Mesa Site
In this main technical area of the Laboratory is the
Administration Building that contains the Director's office and
administrative offices and laboratories for several divisions. Other
buildings house the Central Computing Facility, Personnel
Administration Department offices, Materials Department, the science
museum, Chemistry and Metallurgy Division, Physics Division, technical
shops, cryogenics laboratories, a Van de Graaff accelerator, and
cafeteria.
2.7-1
-------
TA-21, DP-Site
This site has two primary research areas, DP West and DP East. DP
West is concerned with tritium research. DP East is the high
temperature chemistry site where studies are conducted on the chemical
stability and interaction of materials at temperatures up to and
exceeding 3300° C.
TA-33 ,_HP-Si. te
Design and development of nuclear and other components of weapon
systems are conducted here. A major tritium handling facility is
located here. Laboratory and office space for Geosciences Division
related to the Hot Dry Rock Geothermal Project are also here.
TA-35, Ten Site
Nuclear safeguards research and development, which is conducted
here, is concerned with techniques for nondestructive detection,
identification, and analysis of fissionable isotopes. Research in
reactor safety and laser fusion is also done here.
TA-41, W-Site
Personnel at this site are engaged primarily in engineering design
and development of nuclear components, including fabrication and
evaluation of test materials for weapons. Also located here is an
underground laboratory that is used for physics experiments.
TA-43, Health Research Laboratory
The Biomedical Research Group does research here in cellular
radiobiology, molecular radiobiology, biophysics, mammalian
radiobiology, and mammalian metabolism. A large medical library,
special counters used to measure radioactivity in humans and animals,
and animal quarters for dogs, mice, and monkeys are also located in this
building.
TA-46, WA Site
Here applied photochemistry, which includes development of
technology for laser isotope separation and laser-enhancement of
chemical processes, is investigated. Solar energy research,
particularly in the area of passive solar heating for residences, is
done.
2.7-2
-------
TA-48, Radiochemistry Site
Laboratory scientists and technicians at this site study nuclear
properties of radioactive materials by using analytical and physical
chemistry. Measurements of radioactive substances are made and "hot
cells" are used for remote handling of radioactive materials.
TA-50, Waste Management Site
Personnel at this site have responsibility for treating and
disposing of most contaminated liquid wastes received from Laboratory
technical areas, for development of improved methods of waste treatment,
and for containment of radioactivity removed by treatment. Radioactive
waste is piped to this site for treatment from many of the technical
areas.
TA-53, Meson Physics Facility
The Los Alamos Meson Physics Facility (LAMPF), a linear particle
accelerator, is used to conduct research in the areas of basic physics,
cancer treatment, materials studies, and isotope production.
TA-54. Waste Disposal Site
This is a disposal area for radioactive and toxic wastes.
TA-55. Plutonium Processing Facilities
Processing of plutonium and research in plutonium metallurgy are
done here.
2.7.3 Radionuclide Emissions
Radioactive airborne releases at Los Alamos are summarized in
Table 2.7-1. Emissions from all stacks within a Technical Area were
summed, and the curie quantities of each radionuclide discharged within
an Area are listed (DOE82). Emissions include various isotopes of
uranium and plutonium, americium-241, and activation products
(beryllium-7, carbon-11, nitrogen-13, oxygen-15, phosphorus-32,
argon-41, and tritium).
The Los Alamos site covers approximately 111 square kilometers and
is nestled between several residential areas. Except for TA-33, the
major source of tritium, all areas that contributed radioactive airborne
contaminants are grouped along and within a few kilometers of the
northern site boundary. Thus, it was decided to treat all emissions as
two point sources; one is tritium from TA-33, and the other consists of
all the remaining effluents and is located roughly central to all among
the other 12 TAs. The effluents listed in Table 2.7-1 were summed to
2.7-3
-------
provide the radioactive source terms for the two point sources. These
quantities are listed in Table 2.7-2. All effluents are released from
stacks with assumed heights of 30 meters.
The effluent control devices at LANL are determined by the type of
activity conducted at the facility. Facilities in which transuranics
are handled are equipped with glove boxes and hot cells and use negative
pressure zonation to ensure containment of accidental releases. Exhaust
streams from these facilities are prssed through particulate filters
(usually HEPA units, although bag filters and cyclones are also used)
prior to discharge from building stacks.
Activated gases produced at facilities conducting fusion beam
research are held up to allow the decay of short-lived isotopes. There
are no effluent controls fitted to the test reactors at the Omega Site.
2.7.4 Health Impact Assessment of Los Alamos National Laboratory
The health risk assessment performed for this facility is
summarized in Tables 2.7-3 and 2.7-4. The assessment was based on all
emissions being combined into two point sources: those from the TA33
site, and those from a hypothetical stack that was considered the source
for all other site emissions. The health effects are reported
separately for these two emission sources. The individual receiving the
maximum exposure from the TA33 source could be located 930 m southwest
of the stack, while the individual receiving the maximum exposure from
the combined area source could be located 2100 m south-southwest of the
hypothetical stack. The population within the 80 km radius assessment
area is 100,000 people.
The dose equivalents received by five organs of the maximum
individual exposed to emissions from the hypothetical (combined
emissions) stack were about 15 to 20 times those resulting from exposure
to emissions from TA33 (see Table 2.7-3). The dose equivalents to the
higher exposed individual ranged from 9 to 11 mrem/y, with the red
marrow and endosteal cells receiving the largest dose equivalents.
External exposure (immersion) to carbon-11 and oxygen-15 contributed
over 80 percent of these dose equivalents.
The collective dose equivalents listed in Table 2.7-3 were summed
for the two sources yielding a total collective dose equivalents for the
regional population that ranged from a maximum of 62 person-rem/y to the
endosteal cells to 51 person-rem/y to the muscle. Carbon-11, oxygen-15,
and nitrogen-13 were responsible for over 95 percent of the collective
dose equivalents. The principal exposure pathway to the population was
immersion.
The risks of having fatal cancer as a result of exposure to the
radioactive emissions from this facility are listed in Table 2.7-4. The
highest individual lifetime risk is 2E-4 (20 cancers in 105 people),
while the risk within the regional population for the combined sources
is 1E-2 fatal cancers each year of facility operation.
2.7-4
-------
Table 2.7-1. Radionuclide emissions (Ci) from
Los Alamos National Laboratory, 1981
Radionuclide
Tritium
Beryllium-7
Carbon-11
Nitrogen-13
Technical Area
2 3 21 33 35 41
9.0E+2 1.1E+2 6.1E+3 1.3E+2
43
Oxygen-15
Phosphorus-32
Argon-41
Iodine-131
3.0E+2
Oxygen-15
Phosphorus-32
Argon-41
Iodine-131
Uranium-235
Uranium-238
Uranium-235/238
Plutonium-239
Plutonium-238/239
Americium—241
MFP
4.4E-5
2.0E+5
1.4E-5
2.3E-6
1.3E-6 1.6E-6
1.2E-7
1.4E-3 2.3E-5
2.0E-5
Uranium-235
Uranium- 238
Uranium-235/238
Plutonium-239
Plutonium-238/239
Ame r ic ium- 24 1
MFP
Radionuclide
Tritium
Beryllium-7
Carbon-11
Nitrogen-13
1.8E-6
1.6E-4
5.3E-5
4.0E-5
1.7E-4
46
l.OE-3
6.2E-6
5.9E-6
2.9E-7
2.8E-6
48
2.7E-7
Technical Area
50 53
6.6
3.9E-H
1.3E+5
2.5E+4
3.7E-7
54 55
9.0E-9 4.9E-8
4.8E-8
MFP Mixed fission products.
2.7-5
-------
Table 2.7-2. Radionuclide emissions (Ci) from Los Alamos
National Laboratory, 1981
Radionuclide
Technical Area
33
All other
Tritium 6.1E+3
Berry Ilium- 7
Carbon-11
Nitrogen-13
Oxygen-15
Phosphorus- 32
Argon-41
Iodine-131
Uranium-235
Uranium-238
Uranium-235, -238
Plutonium-239
Plutonium-238, -239
Americium-241
MFP
1.1E+3
3.9E-H
1.3E+5
2.5E+4
2.0E+5
2.0E-5
1 . 4E+3
4.4E-5
l.OE-3
1.7E-4
5.3E-5
9.8E-6
4.6E-5
2.9E-7
1.6E-3
(a>Technical Areas: 2, 3, 21, 35, 41, 43, 46, 48, 50, 53-55. Quanti-
ties summed from Table 2.7-1.
MFP Mixed fission products.
Table 2.7-3. Annual radiation dose rates from radionuclide
emissions from the Los Alamos National Laboratory
From TA33 Source
Organ
Endosteal
Red marrow
Testes
Spleen
Muscle
Maximum
individual
(mrem/y)
5.4E-1
6.8E-1
6.8E-1
6.8E-1
6.8E-1
Regional
population
(person-rem/y)
1.4
1.8
1.8
1.8
1.8
From all other sources
Maximum
individual
) (mrem/y)
1.1E+1
1.1E+1
l.OE+1
9.6
9.1
Regional
population
(person-rem/y)
6.2E+1
5.9E+1
5.7E+1
5.5E+1
5.1E+1
Weighted sum 6.9E-1
1.8
8.5
4.8E+1
2.7-6
-------
Table 2.7-4. Fatal cancer risks due to radioactive emissions from
the Los Alamos National Laboratory
Maximum individual Regional population
(lifetime risk) (total cancers/y of operation)
TA33 IE-5 5E-4
All other sites 2E-4 1E-2
2.7-7
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
LANL82 Los Alamos National Laboratory, Environmental Surveillance at
Los Alamos During 1981, Los Alamos National Laboratory Rept.,
LA-9349-ENV (UC-41), April 1982.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean, Virginia,
1979.
2.7-8
-------
2.8 Oak Ridge Reservation; Oak Ridge, Tennessee
2.8.1 General Description
The Oak Ridge Reservation is located in northeastern Tennessee,
approximately 35 kilometers west of Knoxville, Tennessee. It is in a
valley between the Cumberland and Great Smokey Mountains and consists
of approximately 150 square kilometers.
With the exception of the City of Oak Ridge (located on the
northeastern boundary), the land within 8 kilometers of the Reservation
is predominantly rural used mainly for residences, small farms and
pasture. Approximately 675,000 people live within an 80 kilometer
radius of the site.
2.8.2 Process Description
The three major facilities at the Oak Ridge Reservation are the
Oak Ridge National Laboratory (ORNL), the Oak Ridge Gaseous Diffusion
Plant (ORGDP), and the Y-12 Plant. Also located on the reservation are
the Comparative Animal Research Laboratory and the Oak Ridge Associated
Universities.
Oak Ridge National Laboratory
ORNL is a multipurpose research laboratory involved in basic and
applied research in all areas related to energy. These research
facilities consist of nuclear reactors, chemical pilot plants, research
laboratories and support facilities.
The central radioactive gas disposal facilities release tritium,
iodine-131, and krypton and xenon from radioisotope separations,
reactor operations, and laboratory procedures. The gases undergo HEPA
filtration at their source prior to discharge. 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.
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.
2.8-1
-------
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 hexafluoride (UF) in the uranium-235 isotope.
The principal sources of release from ORGDP 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.
Oak Ridge Associated Universities
The Oak Ridge Associated Universities (ORAU) conduct 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.
2.8.3 Radionuclide Emissions
The principal radioactive atmospheric emissions are uranium-234
and uranium-238 (depleted) from the Y-12 Plant and tritium, krypton-85,
and xenon-133 from the Oak Ridge National Laboratory. Table 2.8-1
summarizes the radioactive airborne emissions from the Oak Ridge
facilities for 1981. For this analysis the total radioactive emissions
are assumed to be released from a vent 10 meters in height at the Y-12
plant. One-half the uranium-234 is assumed to be Class Y and one-half
Class W; all uranium-238 is assumed to be Class Y. Table 2.8-2
compares the radioactive emissions from Oak Ridge for the years
1979-1981.
2.8-2
-------
Table 2.8-1. Radionuclide emissions from Oak Ridge Reservation (Ci/y)
Radionuclide
Carbon-14
Tritium
Iodine-125
Iodine-131
Krypton-85
Plutonium-239(a>
Technetium-99
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Xenon-133
ORAU
1.2E-3
5.2E-3
2.5E-4
2.0E-4
—
-
-
-
-
-
-
2.0E-3
ORGDP
_
-
-
-
2.5E+1
-
3.6E-2
3.7E-3
1.2E-4
2.4E-5
8.1E-4
-
ORNL
_
1 . 1E+4
—
6.0E-1
6.6E+3
7.8E-8
—
-
-
-
-
3.2E+4
Y-12
_
—
—
_
—
—
1.2E-1
-
-
4.0E-2(b)
-
1981
Total
1 . 2E-3
1 . 1E+4
2.5E-4
6.0E-1
6.6E+3
7.8E-8
3.6E-2
1.2E-1
1.2E-4
2.4E-5
4.0E-2(b)
3.2E+4
(a>Reported as "Unidentified Alpha."
( ^Preliminary estimate.
ORAU Oak Ridge Associated Universities.
ORGDP Oak Ridge Gaseous Diffusion Plant.
ORNL Oak Ridge National Laboratory.
Table 2.8-2. Radionuclide emissions from the Oak Ridge Reservation,
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Carbon-14
Tritium
Iodine-125
Iodine-131
Kryp ton-85
Plutonium-239(a)
Technecium-99
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Xenon-133
2.6E-4
5.1E+3
-
3.0E-1
1 . 1E+4
4.8E-6
1.4
1.1E-1
1.4E-3
2.1E-4
7.0E-3
5.1E+4
1.6E-4
1 . 5E+4
2.9E-4
2.3E-1
8.8E+3
4.9E-6
8.8E-1
1.9E-1
8.3E-4
1.2E-4
4.1E-3
4.2E+4
1.2E-3
1 . 1E+4
2.5E-4
6.0E-1
6.6E+3
7.8E-8
3.6E-2
1.2E-1
1.2E-4
2.4E-5
4.0E-2(b)
3.2E+4
(a)Reported as "Unidentified Alpha".
(b)preliminary estimate.
2.8-3
-------
2.8.4 Health Impact Assessment of Oak Ridge Reservation
The estimated annual radiation doses resulting from radionuclide
emissions from the Oak Ridge Reservation are listed in Table 2.8-3.
The maximum individual is located 980 meters north of the assumed
release point location at the Y-12 plant. The predominant exposure
pathway is inhalation. The doses are primarily due to uranium-234 and
tritium.
Table 2.8-4 lists the estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses. The lifetime risk to the maximum individual is estimated
to be 4E-5 and the total number of fatal cancers per year of operation
to be 1E-2.
Table 2.8-3. Radiation dose rates from radionuclide
emissions from the Oak Ridge Reservation
Q Maximum individual Regional population
r8an (mrem/y) (person-rem/y)
Pulmonary 49.8 212
Thyroid 9.3 15
Endosteal 7.6 22
Kidney 5.4 15
Lower large intestine 4.9 13
Weighted sum 17.3 69.6
Table 2.8-4. Fatal cancer risks due to radionuclide
emissions from the Oak Ridge Reservation
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Oak Ridge Reservation 4E-5 1E-2
2.8-4
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States, EPA
520/7-79-006, Environmental Protection Agency, Washington,
D.C., 1979.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean,
Virginia, 1981.
UC82 Union Carbide Corporation, Environmental Monitoring Report,
United States Department of Energy Oak Ridge Facilities,
Calendar Year 1981, Report No. Y/UB-16, Union Carbide
Corporation, Oak Ridge, Tennessee, 1982.
2.8-5
-------
2.9 Paducah Gaseous Diffusion Plant; Paducah. Kentucky
2.9.1 General Description
The Paducah Gaseous Diffusion Plant (PGDP) is a uranium enrichment
cascade plant with a uranium hexafluoride (UFg) manufacturing plant
and various other support facilities (UC82). The plant is located in
McCracken County, Kentucky, about 6 kilometers south of the Ohio River
and 32 kilometers east of the confluence of the Ohio and Mississippi
Rivers. The Paducah uranium enrichment cascade consists of 1812 stages
housed in five buildings with a total ground coverage of about 0.3
km2. Including support facilities, the plant has a total complement
of about 30 permanent buildings.
Except for the large raw water treatment plant, all buildings are
within a 3 km2 fenced area. A buffer area of at least 365 meters in
depth exists on all sides of the fenced area. Beyond the DOE-owned
buffer is an extensive wildlife management area leased or deeded to the
Commonwealth of Kentucky. There are no residences within 900 meters of
any of the process buildings. The nearest incorporated towns are
Metropolis, Illinois, located 8 kilometers to the northeast; and
LaCenter, Kentucky, located 18 kilometers southwest. Paducah,
Kentucky, a city of 35,000, is located 19 kilometers east of the
plant. The population within a 80 km radius is 450,000.
2.9.2 Process Description
The primary plant, the diffusion cascade, contains a physical
process in which UFg is fed into the system, pumped through the
diffusion stages, and eventually is removed as UF6. The product is
enriched in the fissionable uranium-235 isotope and the "tails" are
withdrawn at the bottom as UFg depleted in uranium-235. The process
pumps require electric power, lubrication, and air for cooling. The
compressed gases are cooled by heat exchange fluid which is, in turn,
cooled by recirculating cooling water.
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 of Feed Plant uses hydrogen, anhydrous
hydrogen fluoride (HF), and uranium oxide (1103) to produce the UFg
that is fed into the diffusion cascade. Gaseous emissions, from
fluorination operations of UF4 to UF5, are passed through a series
2.9-1
-------
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 UFg from the Cascade is reacted
with HF to convert it to UF4 which is more easily stored.
2.9.3 Radionuclide Emissions
Radioactive material emissions are from two discharge points,
C-310 stack and vent C-400 (Table 2.9-1) (DOE81). Releases for 1981
have increased when compared to the average for 1979-1981, except for
technetium which has decreased (Table 2.9-2). All releases were
assumed to be at ground level from vent C-400 (for calculation
purposes). Releases for 1982 from the C-400 stack are expected to be
an order of magnitude smaller due to recent improvements in emission
controls. Also a new 200-ft stack will be used for releases from the
former C-310 stack. All uranium emissions are assumed to be Class W.
Table 2.9-1,
Radionuclide emissions from the Paducah Plant
(Ci/y)
Radionuclide
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Technetium-99
C-310
5.5E-4
2.9E-5
3.6E-7
4 . 7E-4
6.1E-3
C-400
1 . OE-2
5.0E-4
3.0E-5
3.9E-2
Total
1981
1 . OE-2
5.3E-4
3.0E-5
3.9E-2
6.1E-3
2.9-2
-------
Table 2.9-2. Radionuclide emissions from the Paducah Plant,
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Technetium-99
Uranium-234
Uranium-235
Uranium-236
Uranium-238
6. IE -2
2.7E-3
1.7E-4
3.9E-5
7.7E-3
5.3E-2
6 . 5E-4
3.5E-5
4.2E-7
5.5E-4
6.1E-3
1.1E-2
5.3E-4
3.0E-5
4.0E-2
2.9.4 Health Impact Assessment of Paducah Plant
The estimated annual radiation dose from plant emissions are
listed in Table 2.9-3. The maximum individual is located 1100 meters
north of the release location. The predominant exposure pathway is
that of inhalation. The annual radiation dose is primarily from
uranium-234 and uranium-238.
Table 2.9-4 list the estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses. The lifetime risk to the maximum individual is estimated
to be 2E-5 and the total number of fatal cancers per year of operation
to be 2E-4.
2.9-3
-------
Table 2.9-3. Radiation dose rates from radionuclide emissions
from Paducah Plant
Maximum individual Regional population
r&an (mrem/y) (person-rem/y)
Pulmonary 4.7 3.4
Lower large intestine 1.2E-1 2.4E-1
Endosteal 7.1 1.3E+1
Thyroid 2.0E-1 4.3E-1
Kidney 3.6 6.7
Red marrow 5.1E-1 9.3E-1
Weighted sum 1.7 1.5
Table 2.9-4. Fatal cancer risks due to radioactive
emissions from Paducah Plant
Lifetime risk to Regional population
Source maximum individual (Fatal cancers/y of operation)
Paducah Plant 2E-5 2E-4
2.9-4
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, DC, 1981.
UC82 Union Carbide Corporation, Environmental Monitoring Report,
Department of Energy, Paducah Gaseous Diffusion Plant,
Paducah, Kentucky, May 1982.
2.9-5
-------
2.10 Portsmouth Gaseous Diffusion Plant; Piketon. Ohio
2.10.1 general Description
The Portsmouth Gaseous Diffusion Plant is operated by Goodyear
Atomic Corporation, a subsidiary of the Goodyear Tire and Rubber
Company. Support operations include the feed and withdrawal of
material from the primary process; treatment of water for both sanitary
and cooling purposes; decontamination of equipment removed from the
plant for maintenance or replacement; recovery of uranium from various
waste materials; and treatment of sewage wastes and cooling water
blowdown.
The Portsmouth Gaseous Diffusion Plant is located in sparsely
populated, rural Pike County, Ohio, on a 16.2-km2 site about 1.6 km
east of the Scioto River Valley at an elevation approximately 36.6 m
above the Scioto River flood plain. The terrain surrounding the plant,
except for the Scioto River Flood Plain, consists of marginal farm land
and densely forested hills. The Scioto River Valley is farmed
extensively, particularly with grain crops.
Several small communities, such as Piketon, Wakefield, and Jasper,
lie within a few kilometers of the plant. The nearest community with a
substantial population is Piketon (population: 1700), which is
approximately 8 km north of the plant on U.S. Route 23. Population
centers within 50 km of the plant are Portsmouth (population: 26,000),
32 km south; Chillicothe (population: 23,000), 34 km north; Jackson
(population: 7,000), 29 km east; and Waverly (population: 5,000, 11
km north. The total population of the area lying within an 80 km
radius of the plant is approximately 600,000.
2.10.2 Process Description
A cold recovery system is used in the recovery of UFg from
comparatively large volumes of purge gases collected from locations
throughout the plant. The purge gases have low UFg concentrations
with assays of less than 27 percent uranium-235. The purge gases are
passed through refrigerated cold traps to freeze out UFg and then
through NaF traps for removal of remaining traces of UFg prior to
being discharged to the atmosphere by means of air-jet exhausters.
When the traps are full they are valved to holding drums and heated to
vaporize the UF6. After assay determination, the material is fed
back to the cascade at the proper location.
2.10.3 Radionuclide Emissions
The gaseous radioactive discharges for 1981 representing all
cold-recovery activities for the plant are shown in Table 2.10-1.
The total air emission of radioactive material has decreased for most
radionuclides from 1979 to 1981. The most significant release point
2.10-1
-------
for 1981 appears to be X326 Top Purge Vent. This release point
discharged approximately 84 percent of the total plant release. This
is shown in Tables 2.10-1 and 2.10-2. Uranium emissions are assumed to
be Class W.
2.10.4 Health Impact Assessment of Portsmouth Plant
The estimated annual radiation doses resulting from emissions at
the Portsmouth Plant are listed in Table 2.10-3. The maximum
individual is located 1300 meters west-northwest of the release
location. The predominant exposure pathway is that of inhalation. The
doses are primarily from uranium-234.
Table 2.10-4 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population from these doses. The lifetime risk to the maximum
individual is estimated to be 3E-5, and the total number of fatal
cancers per year of operation of the Portsmouth Plant to be 8E-4.
2.10-2
-------
Table 2.10-1. Atmospheric emissions of radionuclides from
the Portsmouth Plant, 1981
Emissions
Source/Radionuclide (Ci/y)
Top Purge Cascade
X326 Top Purge Vent
Protactinium-234M 3.7E-2
Technetium-99 1.OE-1
Thorium-234 3.7E-2
Uranium-234 8.5E-2
Uranium-235 2.5E-3
Uranium-236 3.4E-5
Uranium-238 1.4E-4
X330 Cold Recovery System Vent
Protactinium-234M 2.0E-2
Technetium-99 2.8E-3
Thorium-234 2.0E-2
Uranium-234 9.7E-4
Uranium-235 4.7E-5
Uranium-236 1.1E-6
Uranium-238 5.5E-4
X-333 Cold Recovery
X-333 Cold Recovery System Vent
Protactinium-234M 9.9E-4
Technetium-99 1.2E-3
Thorium-234 9.9E-4
Uranium-234 5.7E-4
Uranium-235 3.3E-5
Uranium-236 1.1E-6
Uranium-238 5.6E-4
X-744-G Oxide Sampling Facility
Hood exhaust vent
Protactinium-234M l.OE-5
Thorium-234 l.OE-5
Uranium-234 4.6E-6
Uranium-235 2.3E-7
Uranium-236 4.5E-9
Uranium-238 2.4E-8
2.10-3
-------
Table 2.10-2. Radionuclide emissions from the Portsmouth Plant
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Protactinium-234M
Technetium-99
Thorium-234
Uranium-234
Uranium-235
Uranium-236
Uranium-238
6.2E-2
1.7E-1
6.2E-2
8.2E-2
2.4E-3
5.6E-4
1 . 9E-3
4.0E-2
2.1E-1
4.0E-2
2.2E-1
6.7E-3
1.1E-4
1.4E-3
5.8E-2
1.1E-1
5.8E-2
8.7E-2
2.6E-3
3.6E-5
1.3E-3
Table 2.10-3. Radiation dose rates from radionuclide
emissions from the Portsmouth Plant
o Maximum individual Regional population
° (mrem/y) (person-rem/y)
Pulmonary
Thyroid
Lower large intestine
Endo steal
Upper large intestine
Kidney
Red marrow
7
2
0.6
11
0.2
5
0.8
11
8
2
35
0.8
17
3
Weighted sum
Table 2.10-4. Fatal cancer risks due to radioactive emissions
from the Portsmouth Plant
Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Portsmouth Plant 3E-5 8E-4
2.10-4
-------
REFERENCES
DOE81 Department of Energy Effluent Information System, EPA Report,
Department of Energy, Washington, DC, 1981.
EPA79 Environmental Protection Agency, Radiological Impact Caused
by Emission of Radionuclides into Air in the United States,
(Preliminary Report), EPA 520/7-79-006, Washingon, D.C.,
August 1979.
GA82 Goodyear Atomic Corporation, Portsmouth Gaseous Diffusion
Plant Environmental Monitoring Report for Calendar Year 1981,
Acox, Anderson, Hary, Klein, and Vausher, Piketon, Ohio,
April 1982.
2.10-5
-------
2.11 Rocky Flats Plant; Jefferson County, Colorado
2.11.1 General Description
The Rocky Flats Plant (RFP) is the prime DOE facility for the
fabrication and assembly of plutonium and uranium components for nuclear
weapons. The two programs at RFP that involve the handling of
significant quantities of plutonium are component fabrication and
assembly and plutonium scrap recovery. Fabrication operations use the
metallurgical processes of casting, milling, machining, cleaning, and
etching. These mechanical processes for producing weapons components
generate plutonium scrap. The scrap is collected and recovered on the
site.
Uranium in both the enriched and depleted forms is handled at RFP.
Depleted uranium is utilized in component fabrication and is treated by
many of the same metallurgical processes as plutonium. Enriched uranium
is recovered from decommissioned weapons and is returned to DOE's
enrichment facility at Oak Ridge for recycling.
The Rocky Flats Plant is located in Jefferson County, Colorado,
approximately 26 kilometers northwest of Denver. The facilities are
located within a 1.55 km2 security area which is situated on 26.5
km2 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.
2.11.2 Process Description
The processes conducted at the plant use plutonium and uranium.
Plutonium 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. Smaller pieces of plutonium are drilled or
broken to provide samples for the Analytical Laboratory and for casting
operations. The formed pieces are 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 plutonium-related operations is
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.
2.11-1
-------
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.
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.
2.11.3 Radionuclide Emissions
Atmospheric emissions from the Rocky Flats Plant are listed in
Table 2.11-1. Manufacturing operations at the site are reportedly
responsible for 85 to 95 percent of the plutonium and uranium emissions
and 55 percent of the tritium released. All particulates are assumed
to be 1 micron in diameter.
Releases from the buildings at RFP are from short stacks and
building vents. Given the relatively small size of the production
area, the 26.5 km^ site is considered to be a ground-level point
source. For the purpose of our analysis, we have assumed that releases
are from a point 2.5 kilometers from the southeastern site boundary.
Several of the release points are similar in release quantities.
For comparison purpose and calculations, Building 771 - Main Plenum was
selected. This point releases 54 percent of the plutonium-239, -240 and
3 percent of the uranium-233, - 234, -235. The most significant
release site for uranium is Building 883, Duct B, which has
approximately 19 percent of the total uranium emission.
A comparison of the source term for 1979 to 1981 is shown in Table
2.11-2.
2.11-2
-------
Table 2.11-1. Atmospheric emissions of radionuclides from
the Rocky Flats Plant, 1981
Source/Radionuclide
Emissions
(Ci)
Plutonium Analytical Laboratory
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Fabrication Assembly Building
Building 707-106 Plenum
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Building 707-108
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Building 707-105
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Building 707-107
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Building 707-101/103
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Building 707-102/104
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Manufacturing
371 Nl + N2
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
2.0E-2
4.4E-7
4.1E-7
3.9E-3
4.7E-8
1.6E-7
2.5E-3
5.5E-8
9.2E-8
4.6E-3
1.6E-7
2.8E-7
1.4E-2
5.5E-8
2.0E-7
2.6E-3
5.0E-8
3.8E-8
6.4E-3
1.2E-8
1.1E-8
4.3E-3
5.7E-8
8.7E-8
2.11-3
-------
Table 2.11-1. Atmospheric emissions of radionuclides from
the Rocky Flats Plant, 1981—continued
„ ,„ , . ,., Emissions
Source/Radionuclide
Manufacturing (continued)
371 South
Tritium 1.6E-3
Plutonium-239, -240 1.6E-8
Uranium-233, -234, -238 1.7E-8
Building 771-Main Plenum
Tritium 8.0E-2
Plutonium-239, -240 4.5E-6
Uranium-233, -234, -238 l.OE-6
Building 77lC-Main Plenum
Tritium 4.5E-5
Plutonium-239, -240 3.8E-7
Uranium-233, -234, -238 7.4E-8
Building 77lC-Room Plenum
Plutonium-239, -240 8.9E-7
Uranium-233, -234, -238 5.6E-8
374 Waste Treatment Facility
374 Spray Dryer
Tritium 7.6E-4
Plutonium-239, -240 5.0E-9
Uranium-233, -234, -238 5.2E-8
Building 774-202
Tritium 1.8E-3
Plutonium-239, -240 7.8E-8
Uranium-233, -234, -238 2.0E-8
Manufacturing Building
Building 776-250
Tritium 1.5E-2
Plutonium-239, -240 1.2E-7
Uranium-233, -234, -238 2.0E-7
Building 776-206
Tritium 1.2E-1
Plutonium-239, -240 5.0E-8
Uranium-233, -234, -238 1.9E-7
2.11-4
-------
Table 2.11-1. Atmospheric emissions of radionuclides from
the Rocky Flats Plant, 1981—continued
„ /r, j • i-j Emissions
Source/Radionuclide . ..
(Ci)
Manufacturing Building (continued)
Building 776-201/203
Tritium 8.4E-4
Plutonium-239, -240 3.1E-9
Uranium-233, -234, -238 1.8E-8
Building 776-205
Tritium 3.8E-2
Plutonium-239, -240 l.OE-8
Uranium-233, -234, -238 2.8E-8
Building 776-204
Tritium 1.5E-2
Plutonium-239, -240 1.1E-7
Uranium-233, -234, -238 5.6E-7
Building 776-251
Tritium 1.7E-8
Plutonium-239, -240 4.8E-8
Uranium-233, -234, -238 1.7E-8
Building 776-252
Plutonium-239, -240 2.7E-8
Uranium-233, -234, -238 1.9E-8
Building 776-202
Plutonium-239, -240 4.1E-8
Uranium-233, -234, -238 2.9E-8
Plutonium Development Building
Building 779-729 Plenum
Tritium 2.1E-3
Plutonium-239, -240 3.1E-8
Uranium-233, -234, -238 l.OE-7
Building 779-782 Plenum
Tritium 4.2E-2
Plutonium-239, -240 2.5E-7
Uranium-233, -234, -238 4.6E-7
2.11-5
-------
Table 2.11-1. Atmospheric emissions of radionuclides from
the Rocky Flats Plant, 1981—continued
,„ , . ,., Emissions
Source/Radionuclide
Laundry
Building 778 Laundry
Plutonium-239, -240 7.4E-8
Uranium-233, -234, -238 4.5E-7
Waste Treatment Facility
Building 374-Main
Tritium 1.9E-2
Plutonium-239, -240 5.8E-8
Uranium-233, -234, -238 1.6E-7
Manufacturing Building
Building 444-Ducts 2 and 3
Uranium-233, -234, -238 9.2E-7
Building 444-Duct 1
Uranium-233, -234, -238 l.OE-6
Building 444-Duct 5
Uranium-233, -234, -238 2.0E-7
Building 447 Main
Uranium-233, -234, -238 1.2E-6
Materials and Process Development Laboratory
Building 865-East
Uranium-233, -234, -238 1.8E-7
Building 865-West
Uranium-233, -234, -238 7.0E-7
Manufacturing Building
Building 881-Ducts 1, 2, 3 and 4
Tritium 4.2E-2
Plutonium-239 3.6E-7
Uranium-233, -234, -238 2.6E-6
Building 881 (Ducts 5 and 6)
Plutonium-239, -240 2.3E-7
Uranium-233, -234, -238 4.2E-6
2.11-6
-------
Table 2.11-1. Atmospheric emissions of radionuclides from
the Rocky Flats Plant, 1981—continued
Source/Radionuclide Emissions
(Ci)
Manufacturing Building (continued)
Building 883-Duct A
Uranium-233, -234, -238 7.0E-6
Building 883-Duct B
Uranium-233, -234, -238 5.8E-6
Nuclear Safety Facility
Building 886-875
Plutonium-239, -240 1.2E-8
Uranium-233, -234, -238 2.3E-7
Equipment Decontamination Building
Building 889-Main
Plutonium-239, -240 1.5E-8
Uranium-233, -234, -238 8.8E-7
Assembly Building
Building 991-985
Plutonium-239, -240 8.8E-9
Uranium-233, -234, -238 1.6E-7
991 Main
Plutonium-239, -240 3.2E-8
Uranium-233, -234, -238 8.2E-8
2.11-7
-------
Table 2.11-2. Radionuclide emissions from the Rocky Flats Plant
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Tritium
Plutonium-239, 240
Uranium-234
8.0E-1
5.4E-6
9.0E-6
7.8E-1
1.2E-5
4. 3E-1
7.8E-6
Uranium-238
Uranium-234, 235,
and 238
Uranium-233, 234
Uranium-238
2.5E-5
1.5E-5
1.4E-5
3.0E-5
2.11.4 HealthImpact Assessment of Rocky Flats Plant
The estimated annual radiation doses resulting from radionuclide
emission from the Rocky Flats Plant are listed in Table 2.11-3. The
maximum individual is located 2260 meters north northwest of the release
location. The predominant exposure pathway is that of inhalation. The
doses are primarily from uranium-233, -234, -238 (70 percent); and
plutonium-239 and -240 (30 percent).
Table 2.11-4 lists the estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses. The lifetime risk to the maximum individual is estimated
to be 4E-8 and the total number of fatal cancers per year of operation
of the Rocky Flats Plant to be 6E-6.
2.11-8
-------
Table 2.11-3. Radiation dose rates from radionuclide
emissions from the Rocky Flats Plant
Organ Maximum individual Regional population
(mrem/y) (person-rem/y)
Endo steal 1.5E-4 1.6E-1
Pulmonary 1.2E-2 1.3E-1
Liver 2.8E-3 2.9E-2
Red Marrow 1.2E-4 1.2E-2
Testes 2.6E-4 2.5E-3
Weighted sum 4.3E-3
Table 2.11-4. Fatal cancer risks due to radioactive emissions from
the Rocky Flats Plant
Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Rocky Flats Plant 4E-8 6E-6
2.11-9
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, DC, 1981.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emission of Radionuclides into Air in the United States,
(Preliminary Report), EPA 520/7-79-006, Washington, DC, August
1979.
2.11-10
-------
2.12 Savannah River Plant; Aiken. South Carolina
2.12.1 General Description
The Savannah River Plant (SRP) is located in South Carolina on the
Savannah River, approximately 35 kilometers southeast of Augusta,
Georgia, and 150 kilometers north-northwest of Savannah, Georgia. The
site occupies an area of approximately 770 square kilometers and lies
within portions of Aiken, Barnwell, and Allendale Counties of South
Carolina.
The facilities at SRP are used primarily in the production of
plutonium and tritium, the basic materials for the fabrication of
nuclear weapons. Additional activities at Savannah River include the
production of special nuclear materials for medical and space
applications.
2.12.2 Process Description
SRP facilities are grouped into five major areas according to
their operational function in the plutonium recovery process. These
areas and the major activities performed there include:
100 Area - three nuclear production reactors;
200 Area - plutonium and uranium separations, waste management;
300 Area - fuel and target fabrication;
400 Area - heavy water recovery and production;
700 Area (Savannah River Laboratory) - research and process
development and pilot-scale demonstration projects.
100 Area - Nuclear Production Reactors
Of the five production reactors at SRP, only three (the P, K, and
C reactors) are currently used for plutonium production. The other
two, R and L, have been on standby status since 1964 and 1968,
respectively. The L reactor is being upgraded and will be restarted in
the fall of 1983. The impact of the L reactor restart is discussed in
a later section. The three operating reactors are used to produce
plutonium and tritium by the irradiation of uranium and lithium using
heavy water (D£0) as both primary coolant and neutron moderator. The
heavy water is circulated in a closed system through heat exchangers.
200 Area - Separations and Waste Management Facilities
Nuclear fuel reprocessing takes place in the 200 Area, where the F
and H Separations Facilities are sited. Plutonium is recovered in the
2.12-1
-------
F Area, and uranium and other special nuclear materials are recovered
in the H Area.
Plutonium is recovered from irradiated uranium in the F-Canyon
Building using the Purex solvent-extraction process. The recovery of
enriched uranium from reactor fuel and the recovery of plutonium-238
from irradiated neptunium are done in the H-Canyon Building. Both
activities are performed using a procedure similar to the Purex
process. Tritium is recovered from irradiated lithium/aluminum targets
in three other H Area buildings.
Solid and liquid wastes from this and other DOE facilities are
stored between the F and H Separation Areas.
300 Area - Fuel and Target Fabrication
Fuel and target fabrication operations are conducted in three
facilities: the Alloy Extrusion Plant, the Uranium Metal Element
Fabrication Plant, and the Target Extrusion Plant. Support facilities
include two test reactors and the Metallurgical Laboratory.
Tubular fuel and target elements are produced at the two target
extrusion plants. Coextrusion is used to clad depleted uranium (0.2
percent uranium-235) fuel and target elements with aluminum or a
mixture of lithium and aluminum. A low-power reactor and a subcritical
test reactor are then used to test the fabricated reactor elements for
cladding defects. These elements are then shipped to the production
reactors in Area 100 for irradiation.
Once the elements have been irradiated by the SRP reactors, they
are inspected in the Metallurgical Laboratory. The Metallurgical
Laboratory facilities are also used to test materials produced in the
300 Area.
400 Area - Heavy Water Production and Recovery
Activities in the 400 Area include both the production and the
recovery of heavy water (D20). These operations are performed in two
distillation plants and one extraction plant. The Drum Cleaning
Facility and Analytical Laboratory are used as support facilities.
Heavy water is produced from river water and recovered from
contaminated reactor coolant. The D£0 is then shipped to the 100
Area where it is used both as moderator and primary coolant in the
production reactors.
2.12-2
-------
700 Area - The Savannah River Laboratory
Research and process development work supporting the overall
mission of SRP is performed at the Savannah River Laboratory (SRL).
Major activities in this area include:
- fabrication of fuel element and target prototypes,
- fabrication of radioisotopic sources for medical, space,
and industrial applications,
- R&D on separations processes at the pilot-scale level,
- thermal and safety studies on reactor operations, and
- applied research in the areas of physics and the
environmental sciences.
2.12.3 Radionuclide Emissions and Control Technology
Annual emissions for all facilities at SRP are summarized by
operational area in Table 2.12-1. Airborne releases and controls for
each SRP area are described below.
100 Area - Nuclear Production Reactors
Carbon-14, argon-41, tritium, and various isotopes of krypton and
xenon are the major radionuclides released from the three production
reactors. Discharges range from tens of curies to hundreds of
thousands of curies per year (Table 2.12-1).
All of the releases from the production reactors are from 60-meter
stacks. All air exhausted from the reactor containment buildings is
filtered through moisture separators, particulate filters, and carbon
beds prior to release. Although these treatments are effective for
particulates and radioiodine, they have little effect on the discharge
of noble gases and tritium.
200 Area - Separations and Waste Management Facilities
Airborne releases from the 200 Area are from the separations
facilities (the waste management facilities reportedly emit no
radionuclides). Operations generating pollutants include the use of
evaporators and furnaces and leakage in the process system. Major
releases include tritium and activation and fission products (Table
2.12-1). Control technologies employed include either scrubbers,
fiberglass filters, high-efficiency sand filters, or oxidation and
moisture trapping.
2.12-3
-------
300 Area - Fuel and Target Fabrication
Airborne effluents released from the 300 Area consist of natural
uranium, unidentified alpha-emitters, and tritium. In 1981, there were
no reported tritium or uranium releases. Off-gases from the Alloy
Extrusion Plant and the Metallurgical Laboratory are passed through
HEPA filters prior to discharge. Exhaust streams from the Uranium
Metal Element Fabrication Plant, the Target Extrusion Plant, and the
test reactors are vented directly from the buildings to ambient air
without filtration. Discharges from the area are made from a variety
of stacks and building vents, and release heights vary from 10 to 31
meters.
400 Area - Heavy Water Production and Recovery
Radioactive discharges from the 400 Area are composed entirely of
tritium. The tritium released is from tritiated reactor coolant waters
and represents less than 1 percent of the total tritium released at SRP
during 1981. Releases from the 400 Area are monitored for some
facilities and estimated for others. The releases are not treated
prior to discharge. Discharges are from building vents and stacks;
release heights range from 10 to 30 meters.
700 Area - Savannah River Laboratory
Airborne releases from SRL include cobalt-60, tritium, and
iodine-131. The amount of tritium and iodine-131 released at the 700
Area accounts for less than 1 percent of the total site release of each
nuclide. The cobalt-60 is the only release of this nuclide reported
for the site. All discharges from processing areas are filtered
through at least two stages of HEPA filtration and a multilayered sand
trap before discharge from a 50-meter stack.
Summary of Radioactive Emissions at SRP
The separations facilities and the reactor areas are responsible
for the majority of radioactive releases at SRP. Releases from the
Savannah River Laboratory and the 300 and 400 Areas account for less
than 5 percent of the total SRP releases. The production reactors
release virtually all of the noble gases discharged at SRP and
one-third of the tritium (see Table 2.12-1). Separations activities in
the 200 Area result in the release of two-thirds of the tritium. Fuel
reprocessing activities in the separations areas result in significant
releases of activation products, fission products, and the
transuranics. The size of all particles released is assumed to be 1
micron. Table 2.12-2 indicates the releases for 1979 to 1981.
SRP occupies a large area of 770 square kilometers. Population
densities in the vicinity of the site are relatively low. For these
reasons, SRP is considered to be a point source. The single stack from
which releases are emitted is assumed to be 60 meters high and to be
located in the center of the facility.
2.12-4
-------
2.12.4 Health Impact Assessment of Savannah River Plant
The estimated annual radiation doses resulting from radionuclide
emissions from the Savannah River Plant are listed in Table 2.12-3.
The maximum individual is located 10,500 m east of the assumed release
location (center of site). The predominant exposure pathway is that of
ingestion. The doses are primarily from tritium and argon-41.
Table 2.12-4 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population from these doses. The lifetime risk to the maximum
individual is estimated to be 4E-5 and the total number of fatal
cancers per year of operation of the Savannah River Plant to be 3E-2.
2.12-5
-------
Table 2.12-1. Radionuclide emissions from the
Savannah River Plant, 1981
(Ci/y)
Radionuclide
Americium-24l(a)
Argon-41
Carbon-14
Cerium-141
Cerium-144
Curium-244
Cobalt-60
100
4.4E-6
6 . 2E+4
4.1E+1
-
-
-
—
Area
200 300 400
4.9E-4 3.6E-7
_
2.8E+1
3.2E-4
2.6E-2
1.6E-4
_ _ -
Total
700
5.0E-4
6.2E+4
6.9E+1
3.2E-4
2.6E-2
1.6E-4
8.9E-5 8.9E-5
Cesium-134
Cesium-137
Tritium
Iodine-129(b)
Iodine-131
Krypton-85
Kryp ton~85m
Krypton-87
Krypton-88
Niobium-95
Plutonium-238
Plutonium-239(a)
Ruthenium-103
Ruthenium-106
Strontium-90(b)
Uranium-234
Uranium-238
Xenon-131m
Xenon-133
Xenon-135
Zirconium-95
-
—
1.2E+5
4.5E-4
7.0E-3
-
1 . 3E+3
8 . 7E+2
1 . 5E+3
-
—
4.4E-6
-
-
4.5E-4
-
-
-
3.9E+3
2.5E+3
"
6.4E-4
3.1E-3
2 . 3E+5
1.6E-1
3.7E-2
8.4E+5
—
-
-
6.4E-2
4.57E-3
2.8E-3
1.3E-2
7.8E-2
3.1E-3
6 . 1E-3
6.1E-3
6.4
-
-
1.7E-2
_
_
2.0E+3 1.5E+1
5.0E-6
3.2E-3
_
— — —
_ _ _
_ _ _
_ _ _
_ _ _
3.6E-7
_ _ _
_ _ _
5.0E-6
- -
_ _ _
— — _
- - _
- - _
~ — —
6.4E-4
3.1E-3
3.5E+5
1.6E-1
4.7E-2
8.4E+5
1.3E+3
8.7E-T:
1.5E+3
6.4E-2
4.6E-3
2.8E-3
1.3E-2
7.8E-2
3.5E-3
6.1E-3
6.1E-3
6.4
3.9E+3
2.5E+3
1.7E-2
(a)lncludes one-half that activity designated as "Unidentified Alpha."
(b)lncludes one-half that activity designated as "Unidentified Beta +
^ ••
Gamma.
2.12-6
-------
Table 2.12-2. Radionuclide emissions from the Savannah River Plant,
1979 to 1981 (Ci/y)
Radionuclide
Amer i c ium-24 1 ( a )
Argon-41
Carbon-14
Cerium-141
Cerium-144
Curium-244
Cobalt-60
Cesium-134
Cesium-137
Tritium
Iodine-129
-------
Table 2.12-3. Radiation dose rates from radionuclide
emissions from the Savannah River Plant
Maximum individual Regional population
(mrem/y) (person-rem/y)
Thyroid
Lower large intestine
Upper large intestine
Stomach
Endosteal
Pulmonary
4.9
2.9
2.5
2.4
2.2
2.1
1 . 2E+2
1.4E+2
1 . 1E+2
1 . 1E+2
9.7E+1
9.8E+1
Weighted sum 2.2 9.8E+1
Table 2.12-4. Fatal cancer risks due to radioactive emissions
from the Savannah River Plant
„ Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
SRP 4E-5 3E-2
2.12-8
-------
REFERENCES
DOESla Department of Energy, Environmental Monitoring in the Vicinity
of the Savannah River Plant, Annual Report for 1981,
DPSPU-82-30-1, E.I. du Pont de Nemours and Company, Aiken,
South Carolina, 1981.
DOESlb Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc. McLean, Virginia,
1979.
2.12-9
-------
2.13 Feed Materials Production Center; Fernald. Ohio
2.13.1 General Description
The Feed Materials Production Center (FMPC), operated by the
National Lead Company of Ohio, is located on 425 hectares in
southwestern Ohio in Hamilton and Butler counties. The facility is 1.6
kilometers north of Fernald and 32 kilometers northwest of Cincinnati.
The population within an 80 kilometer radius of FMPC is 2.6 million.
2.13.2 Process Description
The Feed Materials Production Center produces purified uranium
metals, uranium rod and tubing extrusions, uranium compounds, and some
thorium compounds for use by other Department of Energy (DOE)
facilities. Uranium may be natural, depleted, or enriched with respect
to uranium-235; the average uranium-235 content is that of natural
uranium. Feed stock may be ore concentrates, recycled uranium, or
various uranium compounds.
Impure feedstock is dissolved in nitric acid, and the uranium is
separated by organic liquid extraction. It is then reconverted to
uranyl nitrate, heated to form a trioxide powder, reduced with hydrogen
to uranium dioxide, and reacted with anhydrous hydrogen fluoride to
produce uranium tetrafluoride. Purified metal is produced by reacting
uranium tetrafluoride with metallic magnesium in a refractory-lined
vessel, remelted with scrap uranium metal, and cast into ingots. From
these ingots uranium rods and tubing are extruded, cut, machined, and
finally sent to other DOE facilities for fabrication into nuclear
reactor fuel elements.
The facility periodically purifies small quantities of thorium
through production steps similar to those outlined above for uranium.
Finished products include thorium metal, thorium nitrate solution, and
solid thorium compounds.
There are eight buildings at FMPC for these production
activities. Exhausted air from these buildings is passed through
scrubbers or cloth type bag filters prior to release to building
stacks. The processes associated with each of the eight buildings are
as follows:
Plant 1 Material sampling and grinding;
Plant 2 Dry feeds digestion;
Plant 4 Uranium tetrafluoride production and
repackaging;
Plant 5 Metal production and slag grinding;
Plant 6 Metal machining;
2.13-1
-------
Plant 8
Plant 9
Pilot Plant
Dumping and milling;
Metal production, remelting, and machining;
Uranium and thorium metal and compound
production.
2.13.3 Radionuclide Emissions
Table 2.13-1 summarizes the radionuclide emissions from FMPC in
1981 for each of the eight stacks and an on-site incinerator. Only
natural uranium was released during 1981; no thorium was released
during the year.
Table 2.13-1. Radionuclide emissions from
Feed Materials Production Center, 1981 (Ci/y)
Source
Total
Natural uranium emissions
(Ci/y)
Plant 1
Plant 2
Plant 4
Plant 5
Plant 6
Plant 8
Plant 9
Pilot Plant
Incinerator
3.3E-4
0.
6.26E-2
4.46E-2
0.
5.33E-3
0.
0.
4.15E-4
0.113
2.13.4 Health Impact Assessment
For the health impact assessment, all releases were assumed to
originate from a single 10-meter stack at the center of the production
area. The nearest site boundary is 680 meters. Since only natural
uranium was released during 1981, the assumption was made that the
release consisted of one-half uranium-234 and one-half uranium-238 in
equilibrium with its daughters, thorium-234 and protactinium-234m.
The estimated annual radiation doses from radionuclide emissions
from FMPC are shown in Table 2.13-2. These estimates are for a
regional population of 2.6 million. The maximum individual is located
810 m northeast of the release point at the site boundary. The major
pathway of exposure is inhalation, and the critical organ is the
pulmonary, with a dose equivalent of 88 mrem/y.
2.13-2
-------
The individual lifetime risk and the number of fatal cancers per
year of operation are shown in Table 2.13-3. The lifetime risk to the
maximum individual is estimated to be 2E-4 and the total number of
fatal cancers per year of operation is estimated to be 2E-2.
Table 2.13-2. Radiation dose rates from radionuclide
emissions from the Feed Materials Production Center
Orean Maximum individual Regional population
(mrem/y) (person-rem/y)
Pulmonary 88 440
Endosteal 26 114
Kidney 12 56
Lower large intestine 0.4
Red marrow 1.8 8
Weighted sum 26 132
Table 2.13-3. Fatal cancer risks due to radionuclide
emissions from the Feed Materials Production Center
Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
FMPC 2E-4 2E-2
2.13-3
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
ERD77 Energy Research and Development Administration, Feed Materials
Production Center, Environmental Monitoring Annual Report for
1976, NLCO-1142, Bobach, M. W., et al., National Lead Company
of Ohio, Cincinnati, Ohio, 1977.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States,
Preliminary Report, EPA 520/7-79-006, Environmental Protection
Agency, Washington, D.C., 1979.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean, Virginia,
1979.
2.13-4
-------
2.14 Ames Laboratory; Ames, Iowa
2.14.1 General Description
Ames Laboratory is operated by Iowa State University for the
Department of Energy. The principal facility is the Ames Laboratory
Research Reactor, located 2.4 km northwest of the Iowa State University
campus and 4.8 km northwest of Ames, Iowa. The site occupies 16.2
hectares in Story County.
2.14.2 Process Description
The Ames Laboratory Research Reactor (ALRR) was used until 1978 as
a neutron source for the production of byproduct materials and the
neutron irradiation of various materials for research. The reactor was
fueled with enriched uranium, was moderated and cooled by heavy water
(D20), and was operated continuously at 5000 watts thermal. Operation
of the ALRR was terminated on December 1, 1977. Decommissioning began
January 3, 1978, and was completed on October 31, 1981. Also located
at the site is the waste disposal processing facility, serving both the
reactor and the research laboratories located on campus.
2.14.3 Radionuclide Emissions
Prior to decommissioning, the major airborne releases were tritium
and argon-41 from the ALRR. Tritium was the major radionuclide
released during the 1981 decommissioning activities. Table 2.14-1
contains the release data for 1981. These releases are from the 30-
meter reactor stack, located 215 meters from the nearest boundary, with
an annual exhaust volume of 2.5E+14 ml. No airborne emissions have
been found from the research laboratories on the main campus.
2.14.4 Health Impact Assessment of Ames Laboratory
The estimated annual radiation doses from radionuclide emissions
from ALRR are listed in Table 2.14-2. These estimates are based on a
regional population of 630,000. The maximum individual is located 750
meters north of the facility. The major pathway of exposure was
ingestion.
Table 2.14-3 presents estimates of the maximum individual lifetime
risks and the number of fatal cancers per years of operation from these
doses. The lifetime risk to the maximum individual is estimated to be
2E-8 and the total number of fatal cancers per year of operation is
estimated to be 1E-6.
2.14-1
-------
Table 2.14-1. Radionuclide emissions from Ames Laboratory, 1981
Emissions
Radionuclide CCi/v}
Cobalt-60 2.2E-7
Tritium 4.5
Unidentified alpha 1.6E-7
Unidentified beta + gamma 2.7E-6
Zinc-65 2.4E-7
Table 2.14-2. Radiation dose rates from radionuclide emissions
from Ames Laboratory for 1981
0 Maximum individual Regional population
^an (mrem/y) (person-rem/y)
Lower large intestine 1.4E-3 5.5E-3
Endosteal 1.1E-3 4.8E-3
Upper large intestine 1.1E-3 4.5E-3
Stomach 1.1E-3 4.4E-3
Pulmonary 9.6E-4 4.0E-3
Weighted Sum 9.6E-4 3.9E-3
Table 2.14-3. Fatal cancer risks due to radionuclide emissions
from Ames Laboratory, 1981
_ Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
ALRR Stack 2E-8 1E-6
2.14-2
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
DOE82 Department of Energy, Environmental Monitoring Summary for
Ames Laboratory, Calendar Year 1981, IS-4798, Milo D. Voss,
Ames Laboratory, Ames, Iowa, 1982.
ERDA77 Energy Research and Development Administration, Environmental
Monitoring at Ames Laboratory, Calendar Year 1976, IS - 4139,
Milo D. Voss, Ames Laboratory, Ames, Iowa, 1977.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States,
Preliminary Report, EPA 520/7-79-006, Environmental Protection
Agency, Washington, D.C., 1979.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean, Virginia,
1979.
2.14-3
-------
2.15 Battelle-Columbus Laboratory; Columbus, Ohio
2.15.1 General Description
The Battelle-Columbus Laboratory (BCL) is located in the Greater
Columbus area and conducts activities for the Nuclear Regulatory
Commission as well as the Department of Energy (under Contract No.
W-7405-ENG-92).
BCL operates two complexes, the King Avenue Site in Columbus and
the West Jefferson Site (Nuclear Sciences Area) which is located
approximately 27 kilometers west of the King Avenue Laboratories.
King Avenue Site
The King Avenue Site is located on a four-hectare track which is
bounded on the north by the Ohio State University intramural sports
practice field, on the west by the Olentangy River and on the south and
east by two- and four-family dwellings.
West Jefferson Site (Nuclear Sciences Area)
The Nuclear Sciences Area occupies a four-hectare-fenced security
area on 405 hectares. A 16-kilometer radius circle whose center is at
the site includes a small portion of Columbus having a population of
about 60,000. The only other significant population center near the
site is West Jefferson, Ohio, located about 3.2 kilometers to the
southwest with a population of 5,700.
2.15.2 Process Description
King Avenue Site
The Uranium-235 Processing Facility, located on the first floor of
Building 3, is the management point for all transactions involving
nuclear materials at the King Avenue Site. Building 3 also houses the
Melting Facility and the Power Metallurgy Laboratory.
Activities involving contract and license materials were very
limited during 1981; therefore, effluent monitoring at this site was
limited to liquid discharges only. There were no reported airborne
releases for 1981.
West Jefferson Site (Nuclear Sciences Area)
Facilities in this area include the JN-1 Hot Cell Facility, the
JN-2 Vault Facility, and the JN-4 Plutonium Laboratory which is being
decontaminated. The JN-4 Plutonium Laboratory was conducting research
on uranium-235/plutonium-239 nitride reactor fuel. The Nuclear
Sciences Area also houses a decommissioned Research Reactor.
2.15-1
-------
The JN-1 Hot Cell Facility the JN-2 Vault Facility and are
presently the only facilities at the Nuclear Sciences Area where
contract materials are handled.
Irradiated reactor fuel element studies are conducted at the JN-1
Hot Cell Facility, and materials accountability and storage operations
are conducted at the JN-2 Vault Facility.
2.15.3 Radionuclide Emissions and Control Technology
Radionuclide emissions from the West Jefferson Site are presented
in Table 2.15-1. Emissions for 1979 through 1981 are listed in
Table 2.15-2. All particulates are 1 micron or less in diameter and
are thus respirable. For health impact assessments, we assume Battelle
Laboratories to be a point source with a stack height of 10 meters.
There were no reported airborne releases for the King Avenue
Facility for 1981.
Control Technology
Radionuclide 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.
2.15-2
-------
Table 2.15-1. Radionuclide airborne emissions
Battelle Columbus, West Jefferson Site (Ci/y)
Radionuclide
Bismuth-214
Cerium-141
Cerium-144
Cobalt-57
Cobalt-60
Cesium-134
Cesium-137
Europium-152
Potassium-40
Lead-212
Lead-214
Rhodium-106
Antimony-125
Uranium-235
Plutonium-239
Unidentified
Alpha
Unidentified
Beta and Gamma
JN-1
3.1E-6
3.1E-8
6.6E-7
1.1E-8
3.0E-7
2.5E-6
7.4E-6
1.1E-7
9.0E-6
6.3E-7
3.1E-6
1.4E-7
8.6E-6
9.8E-7
-
-
~
JN-2 JN-4 I***1
3.1E-6
3.1E-8
6.6E-7
1.1E-8
3.0E-7
2.5E-6
7.4E-6
1.1E-7
9.0E-6
6.3E-7
3.1E-6
1.4E-7
8.6E-6
9.8E-7
1.4E-8 1.5E-7 1.6E-7
1.4E-8 - 1.4E-8
1.1E-7 - 1.1E-7
2.15-3
-------
Table 2.15-2. Radionuclide emissions from
Battelle-Columbus, West Jefferson Site (Ci/y)
Radionuclide
Actinium-228
Americium-241
Barium-133
Barium-140
Bismuth-244
Cadmium-109
Cerium-134
Cerium-139
Cerium-141
Cerium-144
Cobalt-57
Cobalt-60
Cesium-134
Cesium- 136
Cesium-137
Chronium-51
Dysprosium-159
Europium-152
Europium-154
Europium-155
Iodine-129
Iodine-131
Potassium-40
Krypton-85
Manganese-54
Neptunium-239
Lead-210
Lead-212
Lead-214
Rhodium-101
Rhodium-106
Antimony-125
Samarium-145
Tellurium-125
Tantalum-182
Terbium- 160
Thorium-228
1979
7.6E-7
4.9E-8
1.7E-6
1.3E-7
9.6E-7
5.7E-7
-
2 . 1E-8
4.2E-8
7.3E-7
1.1E-7
2.5E-6
3.0E-8
3.9E-7
9.8E-8
—
4.0E-8
-
1.5E-8
1.8E-7
5.1E-8
-
-
3.9E-1
1.1E-7
-
3.7E-8
4.1E-7
2.6E-7
6.7E-9
-
5.6E-7
2.3E-7
-
4.4E-8
l.OE-7
"""
1980
1.5E-7
-
2.1E-8
-
5.6E-7
-
4.8E-8
2.9E-8
1.4E-7
5.6E-7
5.9E-8
8.5E-6
-
-
6.9E-7
3.1E-7
—
-
8.8E-11
-
-
8.0E-7
-
2.8E+2
-
8.9E-6
-
1.5E-7
2.3E-7
1.7E-8
_
8.7E-5
-
3.8E-6
—
-
1 . 7E-6
1981
-
-
-
-
3.1E-6
—
-
—
3.1E-8
6.6E-7
1.1E-8
3.0E-7
2.5E-6
-
7.4E-6
—
—
1.1E-7
-
-
-
-
9.0E-6
_
-
-
_
6.3E-7
3.1E-6
-
1.4E-7
8.6E-6
_
_
_
_
-
2.15-4
-------
Table 2.15-2. Radionuclide emissions from
Battelle-Columbus, West Jefferson Site (Ci/y) (Continued)
Radionuclide 1979 1980 1981
Thorium-234
Thallium-208
Uranium-235
Plutonium-238
Plutonium-239
Xenon-138
3.6E-8
3.4E-7
2.3E-8
1.1E-7
2.5E-7
2.1E-7
_
5. IE -8
2.4E-7
—
4,4E-7
—
_
-
9.8E-7
—
1.6E-7
—
Unidentified
Alpha 1.2E-7 - 1.4E-8
Unidentified
Beta and Gamma 1.9E-6 - 1.1E-7
2.15.4 Health Impact Assessment of Battelle-Columbus Laboratory
The estimated annual radiation doses resulting from radionuclide
emissions from the Battelle Laboratories are listed in Table 2.15-3.
The maximum individual is located 150 meters north of the release
location and the predominant exposure pathway is inhalation.
Table 2.15-4 list the estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses. The lifetime risk to the maximum individual is estimated
to be 4E-8 and the total number of fatal cancers per year of operations
of the Battelle-Columbus Laboratory is estimated to be 7E-7.
2.15-5
-------
Table 2.15-3. Radiation dose rates from radionuclide
emissions from Battelie-Columbus, West Jefferson Site
_ Maximum individual Regional population
^an (mrem/y) (person-rem/y)
Endosteal
Pulmonary
Liver
Spleen
Testes
Red marrow
Weighted sum
7.7E-3
7.6E-3
2.0E-3
9.5E-4
9.4E-4
3.1E-3
9.9E-3
l.OE-2
2.7E-3
1.4E-3
2.1E-3
4.2E-3
Table 2.15-4. Fatal cancer risks due to radionuclide
emissions from Battelle-Columbus, West Jefferson Site
s Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
West Jefferson Site 4E-8 7E-7
2.15-6
-------
REFERENCES
DOESla Department of Energy, Environmental Monitoring Report for
Battelle Columbus Laboratories, Annual Report for CY 1981,
Battelle Columbus Laboratories, Columbus, Ohio, 1981.
DOESlb Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (Unpublished), Teknekron Research Inc., McLean, Virginia,
1979.
2.15-7
-------
2-L6 Bettis Atomic Power Laboratory; West Mifflin, Pennsylvania
2-16.1 General Description
The Bettis Atomic Power Laboratory is sited on an 0.8 square
kilometer tract in West Mifflin, Pennsylvania, approximately 12 km
south of Pittsburgh. The facility designs and develops of nuclear
power reactors. Currently, the most significant program at Bettis is
the fabrication of the fuel assemblies used in DOE's light-water-
breeder reactor program. The population within 80 kilometers of the
release is 3.2 million.
2.16.2 Process Description
Bettis facilities, which include both development laboratories and
fabrication facilities, are clustered in the northwest corner of the
site. There is no information available which identifies the
activities conducted within specific buildings at the site. Emissions
data for the site are reported only in aggregate form; therefore, it is
impossible to attribute releases to a specific activity.
2.16.3 Radionuclide Emissions and Existing Control Technology
Airborne emissions data for Bettis are presented in Table 2.16-1.
Reported airborne releases are primarily krypton-85 with much lesser
amounts of antimony-125 and iodine-131.
Gaseous effluent streams from activities at Bettis are treated
with wet scrubbing and passed through charcoal absorbers and HEPA
filtration units prior to release.
2.16.4 Health Impact Assessment of Bettis Atomic Power Laboratory
The entire site is treated as a ground level point source located
centrally within the facility. For purposes of the dose/health effects
assessment, it is assumed that all particulates released are 1 micron
or less in diameter and all releases are respirable. Actual site
boundary distances were used in the calculations.
Table 2.16-2 lists the estimates of the annual radiation doses
resulting from radionuclide emissions. The individual (offsite)
receiving the maximum dose equivalent rate is located 410 meters north
of the release point. The major pathway contributing to the individual
dose equivalent rate is inhalation (76 percent).
Table 2.16.3 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population from these doses. The lifetime risk to the maximum
individual is estimated to be 2E-8. The estimated collective fatal
cancer risk per year of operation is 2E-6. Inhalation is the
predominant pathway contributing to the fatal cancer risk (64 percent.
2.16-1
-------
Table 2.16-1. Radionucllde emissions from
Bettis Atomic Power Laboratory, 1981
Radionuclide
Tritium
Iodine-129
Iodine-131
Krypton-85
Antimony-125
Emissions
(Ci/y)
3.0E-5
2.51-7
8.4E-7
1.6E-1
5.8E-5
Unidentified alpha 1.8E-6
(assumed equally uranium-234
and uranium-238)
Unidentified beta-gamma 1.52E-5
(assumed equally cesium-137,
cobalt-60, and strontium-90)
Table 2.16-3. Radiation dose rates from radionuclide emissions
from the Bettis Atomic Power Laboratory
Maximum individual Regional population
*an (mrem/y) (person-rem/y)
Pulmonary 3.9E-3
Thyroid 1.5E-3
Endosteal 8.6E-4
Red marrow 5.5E-4
Spleen 2.6E-4
Weighted sum 1.4E-3 1.3E-2
Table 2.16-4. Fatal cancer risks due to radioactive
emissions from the Bettis Atomic Power Laboratory
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
BAPL 2E-8 2E-6
2.16-2
-------
REFERENCES
BAPL82 Effluent and Environmental Monitoring Report for Calendar Year
1981, Bettis Atomic Power Laboratory, WAPD-RC/E(ESE)-576, 1982.
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
DOE82 Department of Energy, Summary of Annual Environmental Reports
for CY1980, DOE/EP-0038, 1982.
ERDA77 Energy Research and Development Administration. Environmental
Monitoring at Major U.S. Energy Research and Development
Administration Contractor Sites, Calendar Year 1976, Volumes 1
& 2, ERDA 77-104/1 & /2, Washington, D.C., 1977.
TRI81 Teknekron Research, Inc. Information Base for the Evaluation
and Control of Radioactive Materials to Ambient Air, 1981.
2.16-3
-------
2'17 Knolls Atomic Power Laboratory; Knolls. Kesselring. and Windsor
Sites; Schenectady, New York
2.17.1 General Description
The Knolls Atomic Power Laboratory (KAPL) is operated for the
Department of Energy (DOE) by the General Electric Company. The
facilities of KAPL are located on three separate sites: Knolls,
Kesselring, and Windsor.
The primary missions at KAPL are the development of nuclear
reactors and the training of operating personnel.
Knolls and Kesselring Sites
The Knolls and Kesselring sites are both located in east central
New York State. The Knolls facilities are located on a 0.69 square
kilometer tract about 8 kilometers east of Schenectady. The Kesselring
site is about 27 kilometers north of Schenectady, and occupies an area
of almost 16 square kilometers. Schenectady, Albany, and Troy to the
south, and Saratoga Springs to the north-northeast are the major
population centers in the vicinity. Land use in the vicinity of the
two sites is typical low density residential, with numerous small truck
and dairy farms. The population within 80 kilometers is 1.2 million.
Windsor Site
The Windsor site, which occupies a 0.04 square kilometer tract, is
located just northwest of the town of Windsor, Connecticut. Hartford,
lying 12 kilometers south, and Springfield, Massachusetts, 20
kilometers north, are the major population centers in the vicinity of
the facility. Land in the immediate area (0-10 km) is a mixture of low
density residential and small scale agriculture. The principal crop is
shade-grown wrapper tobacco. Population within 80 kilometers of the
site is 3.1 million.
2.17.2 Process Description
Facilities at the Knolls site are utilized in the development of
nuclear power plants. Nuclear power plant operators are trained at the
Kesselring and Windsor sites. Pressurized water reactors are located
at both the Kesselring and Windsor site.
2.17.3 Radionuclide Emissions and Control Technology
The chemistry, physics, and metallurgy laboratories at the Knolls
site are the only potential emitters of radionuclides to the
atmosphere, while effluents from react »r operations are the only source
of radioactive emissions at the Kesselring and Windsor sites.
2.17-1
-------
All releases at the Knolls site are from elevated stacks (assumed
height 20 meters) and all exhaust streams carrying radioactive
effluents are passed through HEPA filters or activated carbon filters.
The exhaust systems of the reactors at both the Kesselring and
Windsor sites are fitted with HEPA filtration systems to control
particulate emissions. There are no controls for gaseous effluents.
Releases at both sites are from elevated stacks.
Combined airborne emissions for 1981 from the KAPL sites are given
in Table 2.17-1.
2.17-2
-------
Table 2.17-1. Radlonuclide emissions from
Knolls Atomic Power Laboratory
Radionuclide
Argon-41
Bromine-82
Carbon- 14
Cobalt-60
Cesium-137
Iodine-131
Kryp ton-83m
Krypton-85
Krypton-85m
Krypton-87
Krypton-88
Mangane se-54
Plutonium-239
Sulfur-35
Antimony-125
Strontium-90
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Xenon-131m
Xenon-133m
Xenon-133
Xenon-135
Xenon-138
Knolls and
Kesselring
sites
3.8
3.3E-4
1.8E-1
2.3E-6
4.0E-5
4.05E-6
1.1E-3
1.4E-1
3.7E-3
3.4E-3
7.8E-3
2 . 3E-6
1.7E-8
1.8E-6
9.1E-6
4.0E-5
2.9E-5
8.7E-7
5.7E-8
9.0E-10
2.5E-4
1 . 4E-3
4.2E-2
4.0E-2
1.3E-3
Emissions (Ci/y)
Windsor
site
l.OE-4
5.7E-3
4.0E-7
2.4E-4
l.OE-5
8.5E-4
5.9E-4
1.6E-3
5.4E-5
3.7E-4
l.OE-2
9.5E-3
2.17-3
-------
2.17.4 Health Impact Assessment of KAPL
All airborne particles released are assumed to be 1 micron in
diameter and respirable. The assessment is based on all releases for
the Knolls and Kesselrings sites being combined at a central point at
the Knolls site. A release height of 10 meters was assumed for all
effluents. Actual site boundary distances were used for the Knolls
site and the Windsor site. Table 2.17-2 presents the dose rates from
radionuclide emissions at these sites.
Knolls and Kesselring Sites
For the Knolls and Kesselring sites, the maximum individual was
located 300 meters north of the release point. Ingestion was the major
pathway of exposure.
Windsor Site
For the Windsor site, the maximum individual was located 110 m south
of the release point. Inhalation was the major pathway of exposure.
Table 2.17-2. Radiation dose rates from radionuclide emissions
from the Knolls and Kesselring Sites
Maximum individual (mrem/y)
- - *—£•
Knolls and Kesselring sites Windsor site
Endosteal
Red marrow
Muscle
Pulmonary
Lower large intestine
S tomach
8.0E-2
7.8E-2
5.0E-2
4.7E-2
3.8E-2
3.4E-3
3.6E-3
2.4E-3
1.8E-3
1.7E-3
Weighted sum 4.9E-2 2.1
Regional population (person-rem/y)
Knolls and Kesselring sites Windsor site
Weighted sum 1.1E-1 2.1E-3
The lifetime risk to the maximum individual and the total number of
fatal cancers per year of operation of these sites are listed in Table
2.17-3. Air immersion is the major pathway of exposure for these
estimates.
2.17-4
-------
Table 2.17-3. Fatal cancer risks due to radionuclide emissions
from Knolls Atomic Power Laboratory
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Knolls and Kesselring
sites 1E-6 3E-5
Windsor site 4E-8 3E-7
2.17-5
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
DOE82 Department of Energy, Summary of Annual Environmental Reports
for CY1980, DOE/EP-0038, 1982.
ERDA77 Energy Research and Development Administration, Environmental
Monitoring at Major U.S. Energy Research and Development
Administration Contractor Sites, Calendar Year 1976, Volumes 1
and 2, ERDA 77-104/1 & 2, Washington, D.C., 1977.
TRI Teknekron Research, Inc. Information Base for the Evaluation
and Control of Radioactive Materials to Ambient Air, 1981.
2.17-6
-------
2.18 Lawrence Berkeley Laboratory; Berkeley. California
2.18.1 General Description
Lawrence Berkeley Laboratory (LBL) is operated for the Department
of Energy by the University of California-Berkeley. The Laboratory is
located in the Berkeley Hills, above the University of California-
Berkeley campus. The site is three kilometers from downtown Berkeley,
about 20 kilometers from downtown Oakland, and 30 kilometers from
downtown San Francisco. The population within a 50-mile radius of the
Laboratory is 4.5 million. This includes most of the residents of the
greater metropolitan San Francisco Bay area.
Lawrence Berkeley Laboratory is a large multifaceted research
laboratory conducting programs of pure and applied research in
physical, biological, and environmental sciences.
2.18.2 Process Description
LBL research facilities include four large accelerators, several
small accelerators, a number of radiochemical laboratories, and a
tritium labeling laboratory. The large accelerators include the
Bevatron, the Super HILAC, the 224-centimeter Sector-Focused Cyclotron,
and the 467 centimeter Cyclotron.
The Tritium Facility was designed to accommodate kilocurie
quantities of tritium as a labeling agent for chemical and biotnedical
research. Radiochemical and radiobiological studies in many
laboratories typically use millicurie quantities of various
radionuclides.
2.18.3 Radionuclide Emissions
Radionuclide emissions during 1981 at Lawrence Berkeley Laboratory
are shown in Table 2.18-1.
Table 2.18-1. Radionuclide emissions from Lawrence Berkeley Laboratory
Radionuclide Emissions
(Ci/y)
Carbon-14 3.6E-2
Cobalt-60 4.0E-5
Tritium 70.4
Iodine-125 5.7E-4
Plutonium-239 2.5E-9
Strontium-90 4.0E-5
2.18-1
-------
2.18.4 Health Impact Assessment of Lawrence Berkeley Laboratory
Table 2.18-2 lists the estimates of the annual radiation doses
resulting from radionuclide emissions. The maximum individual is
located 100 meters east of the assumed release point. The predominant
exposure pathway is ingestion (80 percent).
Table 2.18-3 gives the estimates of the maximum individual
lifetime risk and the number of fatal cancers per year of operation.
Ingestion is the major pathway for population exposure (74 percent).
Table 2.18-3. Radiation dose rates from radionuclide emissions
from the Lawrence Berkeley Laboratory
Maximum individual Regional population
(mrem/y) (person-rem/y)
Thyroid
Lower large intestine
Upper large intestine
S tomach
Small intestine
Weighted sum
1.6
6.8E-1
5.5E-1
5.4E-1
4.7E-1
5.0E-1
7.1E-1
Table 2.18-4. Fatal cancer risks due to radioactive
emissions from the Lawrence Berkeley Laboratory
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
LBL 1E-5 2E-4
2.18-2
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
LBL81 Lawrence Berkeley Laboratory, Annual Environmental Monitoring
Report of the Lawrence Berkeley Laboratory, Report No.
LBL-19553, University of California, Berkeley, California,
1981.
TRI79 Technekron Research, Inc., Technical Support for the
Evaluation and Control of Emissions of Radioactive Materials
to Ambient Air (unpublished), Technekron Research, Inc.,
McLean, Virginia, 1979.
2.18-3
-------
2.19 Mound Facility; Miamisburg, Ohio
2.19.1 General Description
Mound Facility is located in Miamisburg, Ohio, approximately 16
kilometers southwest of Dayton. Mound Facility has extensive programs
in research and development (R&D), recovery and handling of tritium
from solid waste, and development, fabrication, and testing of weapons
components for the Department of Defense (DOD). Specific programs in
these areas include the separation, purification, and sale of stable
isotopes of noble gases and fabrication of chemical and radioisotopic
heat sources for space and military applications.
2.19.2 Process Description
Nine buildings at the Mound Facility released radioactivity into
the atmosphere in 1981. 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 transferred 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 H, PP, R,
SM, WD, WDA, and 41 Buildings. Contaminated clothing is laundered at
the H Building. Plutonium processing and other related activities are
conducted at the PP Building. At the R Building plutonium heat source
production is the principal activity. The SM Building is an idle
contaminated facility. Operations at the WD, WDA, and 41 Buildings
involve radioactive waste disposal processes. At all these facilities,
particulate radioactivity is removed from process air streams by HEPA
filters.
2.19.3 Radionuc1ide Emissions
Table 2.19-1 identifies radioactive emissions from nine buildings
at the Mound Facility in 1981.
2.19-1
-------
Table 2.19-1. 1981 Radionuclide emissions from the Mound Facility
(Ci/y)
Source
Tritium
H Building stack
HH Building stack
PP Building stack
R Building stack
SM Building stack
SW Building
SW stack
NCDPF stack
HEFS stack
WD Building
WD sludge solidification stack
WDA low risk stack
WDA Building
WDA low risk stack
WDA high risk stack
Building 41 stack
Total curie release
5.26E+1
(Ci/y_)_
"Plutonium-238
1.1E-10
1.21E-6
3.55E-7
6.49E-6
6.13E+2
3.80E+2
3.24E+3
4.29E+3
4.20E-8
4.14E-8
1.07E-7
2.50E-8
2.31E-9
8.28E-6
Total emissions are assumed to be released from the SW Building
with an effective stack height of 61 meters. Table 2.19-2 compares the
radioactive emissions from Mound for the years 1979 to 1981.
2•19.4 Health Impact Assessment of the Mound Facility
The estimated annual radiation doses resulting from radionuclide
emissions from the Mound Facility are listed in Table 2.19-3. The
manimum individual is located 1,500 meters north-northeast of the
assumed release point (SW Building). Ingestion is the major pathway of
exposure (78 percent).
Table 2.19-4 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population from these doses. The regional population within an 80
kilometer radius of the site is 2.9 million. Ingestion is the major
pathway for population exposure (68 percent).
2.19-2
-------
Table 2.19-2.
Radionuclide emissions from the Mound Facility
1979 to 1981 (Ci/y)
Radionuclide
Tritium
Plutonium-238
1979
3.83E+3
1.17E-5
1980
3.80E+3
1.52E-5
1981
4.29E+3
8.28E-6
Table 2.19-3. Radiation dose rates from radionuclide emissions
from the Mound Facility
Organ
Lower large intestine
Upper large intestine
S tomach
Endosteal
Kidneys
Weighted sum
Maximum individual
(mrem/y)
2.9E-1
2.3E-1
2.3E-1
2.0E-1
1.9E-1
1.9E-1
Regional population
(person-rem/y)
8.9
Table 2.19-4. Fatal cancer risks due to radioactive
emissions from the Mound Facility
Source
Lifetime risk
to maximum individual
Regional population
(Fatal cancers/y of operation)
Mound Facility
4E-6
3E-3
2.19-3
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emission of Radionuclides into Air in the United States,
(Preliminary Report), EPA 520/7-79-006, Washington, B.C.,
August 1979.
FA82 Farmer B. M. and Carfagno D. G., Annual Environmental
Monitoring Report:Calendar Year 1981, Report No. MLM-2930,
Monsanto Research Corporation, Mound Facility, Miamisburg,
Ohio, 1982.
2.19-4
-------
2020 Nevada Test Site; Nye County, Nevada
2.20.1 General Description
The Nevada Test Site (NTS) is located in Nye County, Nevada. The
site is approximately 100 kilometers northwest of Las Vegas and covers
an area of about 3,500 square kilometers.
NTS is part of DOE's nuclear weapons research and development
complex. Programs at NTS include the development, redesign and
maintenance of nuclear weapons, nuclear explosion effects studies,
high-energy physics research, and seismic studies. Primary activities
at NTS are centered around the testing of weapons. Tests are conducted
at the site for DOE contractors (e.g., Lawrence Livermore Laboratories,
Los Alamos Scientific Laboratory, Reynolds Electrical Engineering, and
for the Department of Defense).
Additional research includes: testing non-nuclear high-
explosives; and studies of nuclear energy systems (including gas core
reactor testing), disposal of commercial radioactive waste, and storage
of unreprocessed spent fuel.
2.20.2 Process Description
The Nevada Test Site is divided into six operational areas.
Non-weapons programs are conducted in Area 27 and at the NTS
experimental test farm (ERDA77a). Support facilities for most NTS
activities are found in the Mercury vicinity. Underground test sites
include Mesa vicinity (the NTS experimental farm is also located in
this area) and Pahute Mesa vicinity (used for higher yield underground
tests).
2.20.3 Radionuclide Emissions and Existing Control Technology
Radionuclides are released primarily from underground test sites.
Activities responsible for these releases are conducted after
underground nuclear detonations and include re-entry drilling
operations and tunnel ventilations.
Reported releases for drill-back operations and tunnel
ventilations are presented in Table 2.20-1. In addition to the
monitored releases, the source terms from NTS should include the
continuing release (due to leakage) of krypton and tritium. These
releases have not been measured but are estimated to be several hundred
curies per year. Plutonium also contributes to the source term because
of resuspension of soil from contaminated areas, but there are no data
quantifying such emissions. Energy research using the Super Kukla
Reactor and experiments with waste disposal and fuel storage may
possibly release radionuclides, but no releases have been reported for
these operations.
2.20-1
-------
During drill-back operations and tunnel ventilations, emissions
are controlled by passing the air streams through HEPA filters to
control particulates and through charcoal absorbers to control
radioiodine (ERDA77a). There are no applicable controls for the
continued leakage of noble gases and tritium. Although it is possible
to reduce the quantities of plutonium in contaminated areas, these
areas are being used for research into the behavior of plutonium in the
environment (ERDA77a).
2.20.4 Heal_th__Imj>act Assessment of th^J^evgdaJTe^t ^Site
The estimated annual individual radiation dose equivalents from
radionuclide emissions from the Nevada Test Site are shown in Table
2.20-2. The maximum individual is located 34,000 meters south of the
assumed release point located near the center of the test site, and the
weighted sum dose equivalent rate is 1.5E-3 mrem/y. Air immersion (57
percent) is the major pathway for the individual dose equivalent rate.
Table 2.20-3 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population. The individual lifetime fatal cancer risk is 3E-8. The
risk for the regional population per year of operation is 3E-7.
Ingestion (77 percent) is the major pathway contributing to the fatal
cancer risk.
Table 2.20-1. Radionuclide emissions from Nevada Test Site in 1981
Radionuclide Emissions
(Ci/y)
Tritium 534
Iodine-131 0.05
Xenon-133 2700
Xenon-133m 29
Xenon-135 142
2.20-2
-------
Table 2.20-3. Radiation dose rates from radionuclide emissions
from the Nevada Test Site
Organ Maximum individual Regional population
(mrem/y) (person-rem/y)
Edosteal 2.1E-3
Red marrow 2.1E-3
Thyroid 1.8E-3
Testes 1.6E-3
Muscle 1.5E-3
Weighted sum 1.5E-3 l.OE-3
Table 2.20-4. Fatal cancer risks due to radioactive
emissions from the the Nevada Test Site
Lifetime risk Regional population
naximum individual (Fatal cancers/y o
NTS 3E-8 3E-7
ource to maximum individual (Fatal cancers/y of operation)
2.20-3
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
ERDA77a U.S. Energy Research and Development Administration. Fianl
Environmental Statement: Nevada Test Site, Nye County,
Nevada, ERDA-1551, September 1977.
ERDA77b U.S. Energy Research and Development Administration.
Environmental Monitoring at Major U.S. Energy Research and
Development Administration Contractor Sites, Calendar Year,
1976, Volumes 1 and 2, ERDA 77-104/1 & 2, Washington, D.D.,
1977.
2.20-4
-------
2.21 Pantex Plant; Amarillo, Texas
2.21.1 General Description
The Pantex Plant is operated for the Department of Energy (DOE) by
Mason & Hanger - Silas Mason Company, Inc. Pantex is a weapons testing
and surveillance facility. Primary objectives of the plant include:
- fabrication and test firing of chemical high explosives,
- assembly of nuclear weapons,
- surveillance of atomic weapon stockpiles, and
- retirement of atomic weapons.
The Pantex Plant is situated on a 37 square kilometer site in the
Texas panhandle, approximately 30 kilometers northeast of Amarillo,
Texas.
The Pantex Plant is split into numerous areas and some areas are
only 250 meters from the boundary. Land in the vicinity of Pantex is
almost exclusively rural, with agricultural activities having the most
significant impact on the area economy. Principal crops are wheat and
grain sorghums. Cattle ranching and feeding are also of importance.
There is almost no industry in the area.
The population within 80 kilometers of Pantex is approximately
218,000. This includes Amarillo, located 30 kilometers to the
southwest with a population of 185,000, and Pampa, 65 kilometers to the
northeast with a population of 21,000.
2.21.2 ^rocessJDescripJtion
The primary mission at Pantex involves assemblying, monitoring,
and retiring atomic weapons. Significant quantities of plutonium,
uranium, and tritium are handled during these activities. However,
with few exceptions, these materials are handled only in sealed
containers which are not opened at the site. Therefore, normal
emissions at Pantex are limited, although the potential of an accident
involving significant releases does exist.
Pantex conducts explosive test fires of chemical high explosives
as a regular part of its operations. These test fires occur on an
irregular basis, and vary in number from year to year. In recent
years, all such tests were conducted at Firing Site 5, and the only
radioactive material released was depleted uranium-238. The estimated
annual releases have averaged 120 microcuries/year during the years
surveyed.
2.21-1
-------
2.21.3 Ra^dio^nucJL^ide Emis s io ns and ^xi^s^ting^ Cqn.t_ro 1 Technology.
Airborne emissions from Pantex for 1981 are given in Table
2.21-1. Tritium is emitted from the Assembly Area, and depleted
uranium is the only radionuclide released from activities at Firing
Site 5. The emissions for 1979 through 1981 are summarized in Table
2.21-2.
Reports issued by Pantex indicate that no control technology is
being used in the assembly areas since all radioactive materials are
handled in sealed containers. No control technologies are appropriate
to the releases which result from the test firings, so atmospheric
dilution is relied upon.
2.21.4 Health Impact Assessment f9_£ _th?_JPantex_Plant
For the purposes of dose/health effects assessment, it is assumed
that all particles released are 1 micron or less in diameter and that
all are respirable. The assessment is based on all emissions being
combined into one central point on the site. Actual site boundary
distances were used in the calculations.
The estimated annual radiation doses resulting from radionuclide
emissions from the Pantex Plant are listed in Table 2.21-2. The
off site individual receiving the highest weighted sum dose equivalent
rate is located 1,350 meters north of the release point. The major
pathway contributing to the individual dose equivalent rate is
inhalation (97 percent). The collective weighted sum dose equivalent
rate is 7.9E-4 person-rem per year.
Table 2.21-3 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population. The maximum individual fatal cancer risk is 1E-8. The
estimated collective fatal cancer risk per year of operation is 1E-7.
The pathway contributing primarily to the fatal cancer risk is
inhalation.
Table 2.21-1. Radionuclide emissions from Pantex Plant
1979 to 1981 (Ci/y)
Radionuclide
Tritium
Uranium-238
1979
2.0E-2
3.0E-5
1980
l.OE-1
5.0E-5
1981
9.5E-2
l.OE-5
2.21-2
-------
Table 2.21-3. Radiation dose rates from radionuclide emissions
from the Pantex Plant
Orean Maximum individual Regional population
(mrem/y) (person-rem/y)
Pulmonary
Lower large intestine
Upper large intestine
Kidneys
Weighted sum
4.6E-3
5.2E-5
3.8E-5
3.9E-5
1.4E-3
7.9E-4
Table 2.21-4. Fatal cancer risks due to radioactive
emissions from the Pantex Plant
Lifetime risk Regional population
Source fco maximum individual (Fatal cancers/y of operation)
Pantex Plant 1E-8 1E-7
2.21-3
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
DOE82 Department of Energy, Summary of Annual Environmental Reports
for CY1980, DOE/EP-0038, 1982.
ERDA77 Energy Research and Development Administration. Environmental
Moniroring at Major U.S. Energy Research and Development
Administration Contractor Sites, Calendar Year 1976, Volumes 1
& 2, ERDA 77-104/1 & /2, Washington, D.C., 1977.
MHSMP82 Environmental Monitoring Report for Pantex Plant Covering
1981, MHSMP-82-14, 1982.
TRI81 Teknekron Research, Inc., Information Base for the Evaluation
and Control of Radioactive Materials to Ambient Air, 1981.
2.21-4
-------
2.22 Pinellas j»lant; Pinellas County. Florida
2.22.1 General Description
The Pinellas Plant is operated by the Neutron Devices Department
of the General Electric Company. The plant is located on a 39-hectare
site in the center of Pinellas County, Florida, approximately 10
kilometers northwest of St. Petersburg. Pinellas is an integral part
of the nation's weapons program. Major operations include the design,
development, and manufacture of special electronic and mechanical
nuclear weapons components.
2.22.2 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.
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.
2.22.3 Radionuclide Emissions and Control Technology
The principal releases of radioactivity reported are tritium gas,
tritium oxide, krypton-85, and carbon-14. Locations and quantities of
releases reported are in Table 2.22-1.
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 silical gel.
2.22.4 Health Impalet Assessment^ o^f Pinellas Plant
The estimated annual individual radiation dose equivalents from
radionuclide emissions from the Pinellas Plant are shown in Table
2.22-2. The maximum exposed individual is located 470 meters west of
the release point. Ingestion (78 percent) is the major contributor to
the individual dose equivalent rate.
The risks of fatal cancer are shown in Table 2.22-3. The risk for
the regional population per year of operation is 3E-4. The individual
lifetime fatal cancer risk is 5^-6. Inhalation was the pathway
contributing 51 percent of the fatal cancer risk; ingestion was the
pathway contributing 49 percent of the risk.
2.22-1
-------
Table 2.22-1. Radionuclide emissions from Pinellas Plant
Radionuclide
Tritium gas
Tritium oxide
Krypton-85
Carbon-14
Main Stack
129.2
115.3
3.7
Emissions (Ci/y)
Laboratory Stack
89.7
75.4
8.5E-5
Building 800
2.81
4.63
—
Table 2.6-3.
Organ
Radiation dose rates from radionuclide emissions
from the Pinellas Plant
Maximum individual
(mrem/y)
Lower large intestine
Upper large intestine
Stomach
Small intestine
Kidneys
Weighted sum
3.8E-1
3.1E-1
3.0E-1
2.6E-1
2.5E-1
2.5E-1
Regional population
(person-rem/y)
8.8E-1
Table 2.6-4. Fatal cancer risks due to radioactive
emissions from the Pinellas Plant
Source
Lifetime risk
to maximum individual
Regional population
(Fatal cancers/y of operation)
Pinellas Plant
5E-6
3E-4
2.22-2
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
EPA79 Environmental Protection Agency Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States,
Preliminary Report, EPA 520/7-79-006.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean, Virginia,
1979.
2.22-3
-------
2 . 23 BoctoweU_Jnternati£na^^
2.23.1
Rockwell International, a division of Rockwell International
Corporation, has two nuclear energy research and development sites in
the Los Angeles area. Current programs at these two facilities include
the fabrication of test reactor fuel, decontamination, and the design,
production, and testing of components and systems for central power
plants.
Canoga Park, the headquarters site, is approximately 37 kilometers
northwest of downtown Los Angeles. Facilities at Canoga Park are used
for administrative activities and for NRC- and State-licensed
programs. The Santa Susana site (SSFL) is situated in the Simi Hills
of Ventura County, approximately 48 kilometers northwest of Los
Angeles. Facilities owned by the Department of Energy (DOE), as well
as Rockwell -owned NRC- and State-licensed facilities, are located at
SSFL.
2.23.2 Process Description
NRC- and State-licensed activities at Canoga Park include uranium
fuel production (Building 001), research in analytical chemistry
(Building 004), and cobalt-60 gamma irradiation studies. Non-DOE
facilities at the Santa Susana site include the Rockwell International
Hot Laboratory (RIHL) (Building 020), the Nuclear Materials Development
Facility (NMDF) (Building 055), a neutron radiography facility
containing the L-85 nuclear examination and research reactor (Building
093), and several X-radiography inspection facilities.
DOE operations at the Santa Susana site that release radioactive
materials into the atmosphere are conducted at the Radioactive Material
Disposal Facility (RMDF). The two buildings (021-022) that constitute
this facility are used for processing wastes generated by a program for
the decontamination and disposition of DOE facilities. HEPA filters
are in use at RMDF.
2.23.3 Rad io nu c 1 id e End ss ions
Table 2.23-1 compares radioactive releases for the years
1979-1981. The 1981 release information is used in the health impact
assessment section.
2.23-1
-------
Table 2.23-2. Radionuclide emissions from the SSFL
(DOE facilities only), 1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
2.8E-6 1.8E-6 4.1E-6
'a'Mixed fission products; assumed to be strontium-90 for health
impact assessment.
The total emissions are assumed to originate from Buildings 21 and
22 with an effective stack height of 30 meters.
2.23.4 Health Impact Assessment of Rockwell International
The estimated annual radiation doses resulting from radionuclide
emissions Rockwell International Plant are listed in Table 2.23-2. The
maximum individual is located 180 meters north of the assumed release
point (Buildings 21 and 22). Ingestion is the predominant exposure
pathway and is responsible for 71 percent of the dose.
Table 2.23-3 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers per year of operation.
Ingestion is the primary pathway for population exposure. The regional
population within 80 kilometers of the site is 8 million.
2.23-2
-------
Table 2.6-3. Radiation dose rates from radionuclide emissions
from the Rockwell International Plant
Maximum individual Regional population
(mrem/y) (person-rem/y)
Endosteal cells 4.1E-5
Red marrow 2.1E-5
Lower large intestine 1.5E-6
Upper large intestine 3.5E-7
Thyroid 2.8E-7
Weighted sum 4.0E-6 1.1E-4
Table 2.6-4. Fatal cancer risks due to radioactive
emissions from the Rockwell International Plant
Lifetime risk Regional population
Source individual (Fatal cancers/y of operation)
Rockwell 7E-11 3E-8
2.23-3
-------
REFERENCES
DOE81 Department of Energy. Effluent Information System, Department
of Energy, Washington, D.C., 1981.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States,
(Preliminary Report), EPA 520/7-79-006, Washington, D.C.,
August 1979.
ESG82 Energy Systems Group, Environmental Monitoring and Facility
Effluent Annual Report 1981, ESG-82-21, Rockwell
International, Canoga Park, California, 1982.
2.23-4
-------
2<24 Sand i a jjationalJLaboratoriesj^jklbuquar;que^Jjew_Mexico_
2. 24.1 Ggneral_ JDescription
Sandia National Laboratoraies (SNL) is a nuclear ordinance
laboratory with locations in Albuquerque, New Mexico, and Livermore,
California. The Livermore site is discussed in Section 2.6 under the
discussion of the Lawrence Livermore Laboratory. Sandia Laboratories
serves as an interface between the nuclear weapons developed at the Los
Alamos and Livermore Laboratories and military delivery systems. The
Sandia site is located within the limits of Kirkland Air Force Base, 10
kilometers south of Albuquerque. Facilities at Albuquerque are grouped
in five Technical Areas (TAs).
2.24.2 Pro cess De sc rip t ion
The operations at SNL involve testing weapons for quality
assurance and safeguards, arming, and fusing nuclear weapons, and
developing modifications to delivery systems. The major facilities
include the Sandia Pulsed Reactor and the Annular Core Pulsed Reactor,
which are used to irradiate test materials, and the Relativistic
Electron Beam Accelerator. Support facilities include the Neutron
Generator Facility, the Tube Loading Facility, the Fusion Target
Loading Facility, the Tritium Laboratory, and the Nondestructive Test
Facility. These facilities are located at Technical Areas I and V.
TA-I, located in the northwest corner of the site, also houses research
and design laboratories. TA-III is the location for the Sandia
low-level radioactive waste dump.
2.24.3 Radioactive Emissions and Control Technology
Airborne releases from operation at SNL, Albuquerque, are
summarized in Table 2.24-1.
Table 2.24-1. Radionuclide emissions from
Sandia National Laboratories, 1981
Radionuclide Emissions
(Ci/y)
Argon-41 6-84
2.24-1
-------
2.24.4 Health Impact Assessment of Sandla National Laboratories
The entire site is treated as a single ground level point source
release located centrally within the facility. For the purpose of
dose/health assessment, it is assumed that all particulates released
are 1 micron or less in diameter and all releases are respirable.
Actual site boundary distances were used in the calculations.
Table 2.24-2 lists the estimated annual radiation doses from
radionuclide emissions from Sandia National Laboratories at
Albuquerque. The offsite individual receiving the high dose equivalent
rate was 3200 meters west-northwest of the source; the weighted sum
dose equivalent rate was 7.8E-4 mrem per year. Air immersion
contributed essentially 100 percent to the observed dose equivalent
rate and fatal cancer risk.
Table 2.24-3 lists the estimates of the maximum individual
lifetime risks and the number of fatal cancers to the regional
population. The individual lifetime fatal cancer risk was 2E-8 and the
estimated collective fatal cancer risk per year of operation was
10E-7. Air immersion contributed essentially 100 percent of the fatal
cancer risk.
Table 2.24-3. Radiation dose rates from radionuclide emissions
Sandia National Laboratories
n Maximum individual Regional population
(mrem/y) (person-rem/y)
Spleen
Endosteal
Muscle
Red marrow
Upper large intestine
8.7E-4
8. 5E-4
8.1E-4
8.0E-4
7.9E-4
Weighted sum 7.8E-4 3.2E-3
Table 2.24-4. Fatal cancer risks due to radioactive
emissions from Sandia National Laboratories
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
SNL 2E-8 1E-6
2.24-2
-------
REFERENCES
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
DOE82 Department of Energy, Summary of Annual Environmental Reports
for CY1980, DOE/EP-0038, 1982.
ERDA77 Energy Research and Development Administration. Environmental
Moniroring at Major U.S. Energy Research and Development
Administration Contractor Sites, Calendar Year 1976, Volumes 1
& 2, ERDA 77-104/1 & /2, Washington, D.C., 1977.
SNL82 1981 Environmental Monitoring Report, Sandia National
Laboratories, SAND-82-0833, 1982.
TRI Teknekron Research, Inc., Information Base for the Evaluation
and Control of Radioactive Materials to Ambient Air, 1981.
2.24-3
-------
2'25 ShiPPingPort Atomic Power Station; Beaver County. Pennsylvania
2.25.1 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 2.8 square kilometer 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.
2.25.2 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. 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.
2.25.3 Radionuclide Emissions and Control Technology
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.
Table 2.25-1 summarizes the emissions from Shippingport Atomic
Power station in 1977 and 1981. Since the plant was not in operation
the entire year in 1977, the 1977 releases are estimated emissions to
the atmosphere for that year.
2.25-1
-------
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 transferred to one
of four gas storage drums. After a long decay period, the decayed gases
are sampled again before release. In addition, the exhaust from the
containment is equipped with high efficiency filters to prevent release
of radioactive particulates. Protective devices will 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
particulates by fixed filter monitors.
Table 2.25-1. Radionuclide emissions from
Shippingport Atomic Power Station for 1977 and 1981
D ., .., Emissions (Ci/y)
Radionuclide 197?(<0 1981
Argon-41 2.4
Carbon-14 - 7.2E-2
Cobalt-60 3.7E-8 3.7E-8
Kryp ton-83m 1.1E-2
Krypton-85m 1.8E-2
Krypton-85 2.5E-6 6.0E-7
Krypton-87 3.3E-2
Krypton-88 5.7E-2
Manganese-54 3.8E-9
Iodine-130 7.0E-7
Iodine-131 6.4E-5
Iodine-132 4.6E-3
lodine-133 l.OE-3
Iodine-134 1.1E-2
Iodine-135 2.4E-3
Tritium - 8.9E-1
Xenon-13lm 3.0E-8
Xenon-133m 6.1E-4
Xenon-133 6.0E-2 2.4E-4
Xenon-135 1.9E-1
Xenon-135m 4.9E-2
Xenon-137 1.8E-1
Xenon-138 1.5E-2
(a)Estimated for entire year.
2.25-2
-------
2.25.4 Health Impact Assessment of Shippingport Atomic Power Station
No health impact assessment was made with the 1981 emission
data. The health impacts reported in TRI79 (using 1977 data) are
summarized in Tables 2.25-2 and 2.25-3.
Table 2.25-2. Radiation dose rates from radionuclide
emissions from the Shippingport Atomic Power Station, 1977 (TRI79)
Organ
Bone
Red marrow
Muscle
Lung
Liver
Weighted sum
Maximum individual
(mrem/y)
2.8E-4
2.7E-4
2.7E-4
2.5E-4
2.2E-4
2.6E-4
Regional population
(person-rem/y)
5.4E-3
5.1E-3
5.0E-3
4.8E-3
4.3E-3
5.3E-3
Table 2.26-3. Fatal cancer risk due to radionuclide emissions
Shippingport Atomic Power Station, 1977 (TRI79)
Lifetime risk to Regional population
Source maximum individual (Fatal cancers/y of operation)
Shippingport 5E-9 IE-6
2.25-3
-------
REFERENCES
DLC78 Duquesne Light Company, 1978, 1977 Environmental Report,
Radiological - Volume #2, Duquesne Light Company, Beaver
Valley Power Station and Shippingport Atomic Power Station.
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)
DOE81 Department of Energy, Effluent Information System, 1981
Emissions Data, 1981.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emission of Radionuclides into Air in the United States,
(Preliminary Report), EPA 520/7-79-006, Washington, D.C.,
August 1979.
ERDA76 Energy Research and Development Agency, 1976, Final
Environmental Statement, Light Water Breeder Reactor Program,
ERDA-1541, Washington, D.C.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (unpublished), Teknekron Research, Inc., McLean, Virginia,
1979.
2.25-4
-------
2.26 Stanford Linear Accelerator Center; Stanford. California
2.26.1 General Description
The Stanford Linear Accelerator (SLAG) is located in the San
Francisco Bay Area roughly halfway between San Francisco and San Jose.
The total length of the accelerator and the experimental area is
approximately 4.8 kilometers, oriented almost east-west, on about 1.7
square kilometers of Stanford University land. There are 4.2 million
people living in the six counties of the San Francisco Bay Area.
SLAG is a large research laboratory devoted to theoretical and
experimental research in high energy physics and to the development of
new techniques in high energy accelerator particle detectors. The main
tool of the laboratory is a linear accelerator which is used to
accelerate electrons and positrons.
2.26.2 Process Description
The linear accelerator is approximately 3.2 kilometers long and
produces beams of electrons with energies up to 31 billion electron
volts (31 GeV). It can also accelerate positrons, up to 20 GeV. These
beams can be used directly for experiments or they can be transported
into either of two storage-ring facilities-SPEAR or PEP. These
storage-rings are major laboratory facilities, roughly circular in
shape, in which electrons and positrons brought from the accelerator
are stored and circulated continuously in opposite directions. The
energies are 4.5 and 18 GeV per beam for SPEAR and PEP, giving total
collision energies of 9 and 36 GeV, respectively. SPEAR has been in
operation since 1972 and PEP was first filled with beam on
April 13, 1980.
With colliding beam storage rings, such as SPEAR and PEP, the beam
particles are truly 'recycled'; the same particles are brought into
collision over and over again, rather than striking a target only
once. For this reason colliding beam devices produce much less
radiation and residual radioactivity than do conventional accelerators.
2.26.3 Radionuclide Emissions and Control Technology
Airborne radioactivity produced as a result of SLAG operations
and respective half-lives of the radionuclides are listed in Table
2.26-1. During 1981 only 1.1 curies of gaseous radioactivity were
released. For calculational purposes the total release is assumed to
be argon-41. No measurable particulate radioactivity was released.
SLAG does not routinely vent the facility while the beam is on.
There is a waiting period to allow all isotopes, with the exception of
argon-41, to decay before exhausting the facility. The release of
radioactivity is, therefore, infrequent and limited to argon-41 for
brief periods of 30 to 60 minutes.
2.26-1
-------
If personnel entry must be made during an operating cycle, the
facility is vented for 10 minutes prior to entry and after the primary
beam has been shut off. This practice may result in the release of
small quantities of radionuclides other than argon-41.
Control Technology
The primary control of airborne radioactive emissions is hold-up
confinement.
The accelerator, SPEAR and PEP do not represent measurable sources
of gaseous or particulate radioactivity due to low activating potential.
2.26.4 Health Impact Assessment of Stanford Linear Accelerator
The estimated annual radiation doses resulting from radionuclide
emissions from Stanford Linear Accelerator are listed in Table 2.26-2.
The maximum individual is located 250 meters south of the release
location and the predominant exposure pathway is air immersion.
Table 2.26-3 lists the estimates of the maximum individual
lifetime risk and the number of fatal cancers to the regional
population from these doses. The lifetime risk to the maximum
individual is estimated to be 1E-7 and the total number of fatal
cancers per year of operations of the accelerator is estimated to be
9E-6.
Table 2.26-1. Radionuclide emissions from
Stanford Linear Accelerator, 1981
Radionuclide Half-life
Oxygen-15
Nitrogen-13
Carbon-11
Argon-41
2.1 minutes
9.9 minutes
20.5 minutes
1.8 hours
Total activity 1.1 curies
2.26-2
-------
Table 2.26-2.
Radiation dose rates from radionuclide emissions
from Stanford Linear Accelerator
Organ
Maximum individual
(mrem/y)
Regional population
(person-rem/y)
Spleen
Endosteal
Muscle
Red Marrow
ULI Wall
Weighted Sum
5.7E-3
5.6E-3
5.3E-3
5.3E-3
5.2E-3
5.0E-3
3.7E-2
3.6E-2
3.4E-2
3.4E-2
3.3E-2
3.2E-2
Table 2.26-3. Fatal cancer risk due to radionuclide emissions
from Stanford Linear Accelerator
Lifetime risk to
maximum individual
Regional population
(Fatal cancers/y of operation)
1E-7
9E-6
2.26-3
-------
REFERENCES
DOESla Department of Energy, Environmental Monitoring Report for
Stanford Linear Accelerator Center, Annual Report for CY 1981,
Stanford University, Stanford, California, 1981.
DOESlb Department of Energy, Effluent Information System, Department
of Energy, Washington, D.C., 1981.
TRI79 Teknekron Research, Inc., Technical Support for the Evaluation
and Control of Emissions of Radioactive Materials to Ambient
Air (Unpublished), Teknekron Research Inc., McLean, Virginia,
1979.
2.26-4
-------
2.27 Worldwide Impact of Selected Radionuclides
Some radionuclides released from a site may have worldwide health
consequences from their dispersion in the biosphere and their rela-
tively long half-life. The emissions of carbon-14, iodine-129 and
krypton-85 from all Department of Energy sites were considered in this
regard (Table 2.27-1).
Carbon-14
By combining the emission of 67 Ci per year and the dose
equivalent conversion of 700 person-rem per Ci released, a worldwide
dose equivalent of 47,000 person-rem were committed from 1981 emissions
of carbon-14. Similarly, the estimate of fatal cancers committed due
to these emissions (using 0.08 fatal cancers per Ci—Table 2.27-2) is
5. Those effects will be observed during the time it takes carbon-14
to decay away, or over approximately 40,000 years.
Iodine-129 and Krypton-85
The worldwide health impact of emissions of iodine-129 and
krypton-85 are of similar concern. In 1981, 0.19 Ci of iodine-129 and
910,000 Ci of krypton-85 were released from operations at all DOE sites.
The committed collective dose equivalent due to iodine-129 was
50,000 person-rem; for krypton-85, 4000 person-rem.
Health effects conversion factors taken from Table 2.27-2 were
used to calculate estimated fatal cancers committed over the entire
environmental residence time of iodine-129 and krypton-85. For
iodine-129 this was 10 fatal cancers and for the krypton-85 this
yielded an estimated 0.7 fatal cancers. Both of these calculated
values are based on an assumption of 200 fatal cancers per million
person-rem received by the world population.
2.27-1
-------
Table 2.27-1. Emissions of selected radionuclides from DOE facilities
which may lead to worldwide impact
Source(a)
Argonne National Laboratory
Brookhaven National Laboratory
Han ford Reservation
Idaho National Engineering
Laboratory
Oak Ridge Reservation
Savannah River Plant
Combined releases for all DOE
facilities
Emissions (Ci/y)
Carbon-14
0
8.1E-4
3.2
1.7E-1
1 . 2E-3
6.4E+1
6.7E+1
Iodine-129
0
0
0
3.7E-2
0
1.5E-1
1.9E-1
Krypton-85
6.7
0
0
5 . 9E+4
6 . 6E+3
8.4E+5
9 . 1E+5
(a)DOE facility having significant releases of selected radionuclides.
Table 2.27-2. Estimated radiation doses and fatal cancers from
emissions of selected radionuclides from DOE facilities
to the world population
World population
„>. _ -nj. (Fatal cancers/
Kadionuclide , /,,.» „. •, x
(person-rem/Ci) Ci release)
Carbon-14
Krypton-85
Iodine-129
(a)7E+2
(b)4E-3
(e)2.8E+5
(c)8E-2
(d,f)8E-7
(f)6E+l
(a)Dose equivalent recorded by red marrow and endosteal cells, (Un77,
p. 120).
(b)Dose equivalent is received by the skin (Un77, p. 121).
(c)Health effects integrated over all time (Fo79).
-------
REFERENCES
Fo79 Fowler T. W. and Nelson C. B., Health Impact Assessment of
Carbon-14 Emissions from Normal Operations of Uranium Fuel
Cycle Facilities, EPA 520/5-80-004, Office of Radiation
Programs, Environmental Protection Agency, Washington, B.C.,
1979.
Ko81 Kocher, D. C., A Dynamic Model of the Global Iodine Cycle and
J Estimation of Dose to the World Population from Releases to
the Environment, Environment International, Vol. 5, 15-31,
1981.
NCRP75 National Council on Radiological Protection, Krypton-85 in the
Atmosphere, Report No. 44, 1975.
UN77 United Nations Scientific Committee on the Effects of Atomic
Radiation, Sources and Effects of Ionizing Radiation, Annex C,
1977.
2.27-3
-------
2-28 Future operations at DOE Facilities
2.28(A) Resumption of operations at the PUREX Plant
The U.S. Department of Energy has proposed the resumption of fuel
reprocessing in the PUREX plant in the 200 area of the Hanford site.
If the resumption occurs as scheduled, atmospheric releases will be
significantly increased from their present value. For this reason, the
risk from the expected atmospheric emissions have been calculated for
operation of the PUREX plant in the 200 Area of the Hanford site.
Process Description
The PUREX process is based on dissolution, solvent-extraction, and
ion-exchange and is used to recover uranium, plutonium, and neptunium
from the N-Reactor's irradiated fuel elements. Wastes generated during
the process are treated and returned to the process flow or shipped to
the AR Vault for disposal. The PUREX Plant has been operated on an
intermittent schedule, determined by national security needs and the
production of the N-Reactor. The plant has been on standby since 1972,
but a draft Environmental Impact Statement (DOE/EIS-0089D) indicates
that PUREX will be reactivated in 1984 for additional reprocessing of
N-reactor fuel. The PUREX Plant was in operation for 17 years between
1950 and 1972 for separating plutonium from reactor fuel elements
produced by the operating reactors in the 100 Area of Hanford.
The plant is expected to reprocess up to 3000 MT of N-reactor fuel
per year. Estimated releases from PUREX during the forthcoming
operation have been estimated by DOE using experience gained during the
previous operation as well as the effects of improved control
technology which have been added since 1975. A summary of these
estimated atmospheric releases are given in Table 2.28(A)-1.
Radionuclide Emissions and Existing Control Technology at Purex
Table 2.28(A)-1 gives the estimated airborne releases from PUREX
plant assuming a fuel reprocessing rate of 3000 MT per year. Airborne
effluents from all PUREX release points are passed through acid
scrubbers, deentrainers, fiberglass filters, and HEPA filters prior to
release. In addition, emissions from the PUREX plant are passed
through a silver nitrate reactor to remove elemental iodine.
Health Impact Assessment from Operations at the PUREX Plant
The estimated radiation dose rates from resumed operation of the
PUREX Plant are given in Table 2.28(A)-2. The offsite individual
receiving the highest dose equivalent is located 16,000 m south of the
source. The major pathway contributing to the individual dose
equivalent rate is air immersion (43 percent). The five organs
receiving the five highest dose equivalent rate are endosteal cells,
4.8 mrem/yr; red marrow, 2.1 mrem/yr; pulmonary tissue, 2.1 mrem/yr;
thyroid, 1.5 mrem/yr and liver, 1.0 mrem/yr.
2.28-1
-------
The maximum individual lifetime fatal cancer risk is 3E-5. The
estimated collective fatal cancer risk per year of operation is 7E-3
(See Table 2.28(A)-3). Note that this is comparable to the fatal
cancer risk from operation of the other major areas at the Hanford site
(compare with Table 2.4-3).
Table 2.28(A)-1. Estimated radionuclide emssionsions from
resumed operation of the PUREX plant
Radionuclides Emissions
(Ci/y)
Carbon-14 9.0
Tritium 3.0E+3
Iodine-129 5.1E-1
Iodine-131 3.0E-1
Krypton-85 3.3E+6
Plutonium-239 5.7E-3
Strontium-90 1.2
Table 2.28(A)-2 Estimated radiation dose rates from resumed
operation of the PUREX plant
0 Maximum individual Regional population
(mrem/y) (person-rem/y)
Red marrow
Endosteal
Pulmonary
Liver
Thyroid
2.1
4.8
2.1
1.0
1.5
Weighted sum 1.4 3.2E-H
2.28-2
-------
Table 2.28(A)-3. Estimated fatal cancer risks from resumed operation
of the PUREX plant
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
PUREX Plant 3E-5 7E-3
2.28(B) Resumption of L-Reactor Operations at Savannah River Plant
The U.S. Department of Energy has proposed resumption of operation
of the L-Reactor at Savannah River Plant.
Process Description
The L-Reactor has been used to provide raw materials for nuclear
weapons; it has been shut down since 1968. The plant is scheduled to
be capable of operation no later than October 1983.
Radionuclide Emissions From L-Reactor Operations
Table 2.28(B)-1 gives the estimated annual emissions from resumed
operations of L-Reactor. Emissions of tritium, argon-41 and xenon are
the most significant radionuclides based on the quantity released.
Health Impact Assessment from Operations of the L-Reactor
The estimated dose rates from resumption of the L-Reactor are
given in Table 2.28(B)-2 for the individual at the location of highest
risk. This location is onsite; the risk to offsite individuals is an
order of magnitude less. The highest organ doses are received by the
lower large intestine, 5.7 mrem/y; upper large intestine, 5.3 mrem/y;
stomach, 5.0 mrem/y; spleen, 4.6 mrem/y and red marrow, 4.6 mrem/y.
Ingestion is the major pathway for weighted sum dose equivalent rate
(49 percent).
An assessment of the health impact from emissions from resumed
operation of the L-Reactor indicates an estimated onsite individual
lifetime fatal cancer risk of 1E-4 (Table 2.28(B)-3). The estimated
collective cancer risk per year of operation is 5E-3 with 85 percent of
the risk due to tritium (Table 2.28(B)-3). Ingestion is also the major
contributing pathway to health risk (72 percent).
2.28-3
-------
Table 2.28(B)-1 Estimated radionuclide emissions from
resumption of L-Reactor operations at
the Savannah River Plant
Radionuclides Emissions (Ci/yr)
Tritium 5.5E+4
Carbon-14 1.2E+1
Argon-41 2.0E+4
Krypton-85m 6.0E+2
Krypton-87 5.4E+2
Krypton-88 8.0E+2
Xenon-133 1.7E+3
Xenon-135 1.4E+3
Iodine-129 l.OE-4
Iodine-131 4.1E-3
Plutonium-239 5.0E-7
Americium-241 5.0E-7
S trontium-90 1.OE-4
Table 2.28(B)-2. Estimated radiation dose rates
from resumption of the L-Reactor, Savannah River Laboratory
-. Maximum individual'3' Regional population
Organ , , ° / *
(mrem/y) (person-rem/y)
Red marrow
Stomach
Lower large intestine
Upper large intestine
Spleen
4.6
5.0
5.7
5.3
4.6
15.9
18.2
22.2
18.8
15.6
Weighted sum dose equivalet rate 4.5 15.8
(a)At onsite location of highest risk.
2.28-4
-------
Table 2.28(B)-3. Fatal cancer risks due to radionuclide emissions
from resumption of the L-Reactor, Savannah River Laboratory
Source Lifetime risk(a) Regional population
to maximum individual (Fatal cancers/y of operation)
L-Reactor 1E-4 • 5E-3
(a)At onsite location of highest risk.
2.28-5
-------
REFERENCES
DOE82a Department of Energy, Draft Environmental Impact Statement,
Operation of PUREX and Uranium Oxide Plant Facilities,
DOE/EIS-0089D, 1982.
DOE82b Department of Energy, Environmental Assessment, L-Reactor
Operation, Savannah River Plant, DOE/EA-0195, 1982.
2.28-6
-------
Chapter 3: NRC-LICENSED FACILITIES AND
NON-DOE FEDERAL FACILITIES
3.1 Research and Test Reactors
3.1-1 General Description
This category consists of those land-based reactors licensed by
the Nuclear Regulatory Commission that are operated for purposes other
than commercial power production. These uses include basic and applied
research and teaching. There are currently 70 such reactors licensed
to operate in the United States.
3.1.2 Process Description
Research and test reactors are of a wide variety of designs, are
used for different purposes, and operate over a wide range of power
levels. The design types include heavy water, graphite, tank, pool,
homogeneous solid, and uranium-zirconium hydride. Purposes include
testing of reactor designs, reactor components, and safety features;
basic and applied research in the fields of physics, biology, and
chemistry; and education. Power levels range from near zero to 10 MW.
3.1.3 Control Technology
There is no demonstrated treatment technology for control of
emissions of argon-41 from these reactors.
Emissions of tritium are not currently controlled but could be
controlled by use of a catalytic recombiner-
Emissions of both argon-41 and tritium could be reduced by
reducing the amount of time the reactor operates. Argon-41 emissions
could also be controlled by reducing the amount of air that is
irradiated by neutrons, by such techniques as filling voids with an
inert gas and sealing leaks of air into the reactor compartment.
3.1-1
-------
3.1.4 Radionuclide Emissions
Airborne emissions of radioactive materials from research and test
reactors usually contain argon-41 and tritium as the principal
radioactive constituents, and may also contain very small quantities of
other noble gases and some fission products. Some reactors use filters
to remove the small amount of fission products which may be present;
others use no controls.
Research and test reactors are not required to submit data on air
emissions of radionuclides to the Nuclear Regulatory Commission (NRG).
However, some reactor owners do submit these data as part of their
annual operating report. A list of research and test reactors by
design type, which includes their reported radionuclide emissions to
air, is given in Table 3.1-1.
3.1.5 Reference Facility
Table 3.1-2 describes the parameters of a reference reactor used
to estimate the maximum impact on human health. The atmospheric
emissions for the reference facility were chosen as equal to those for
the facility with the highest values shown in Table 3.1-1, and the
actual stack height (50 m) of that facility was used. Other parameters
used in the analysis were assumed to be equal to those for a major
metropolitan area in the northeastern United States.
3.1.6 Health Impact Assessment of Reference Facility
The estimated annual radiation doses from the reference facility
for individuals and population groups are shown in Table 3.1-3.
Individual fatal cancer risks and committed population fatal cancers
are presented in Table 3.1-4. The lifetime fatal cancer risk to the
individual at highest risk is estimated to be 2E-5. The individual at
highest risk is located 1000 meters north of the stack.
The estimated number of potential fatal cancers to the population
living in the region around the reference facility is estimated to be
0.1 per year of reactor operation.
3.1.7 Total Health Impact from Research and Test Reactors
The reference facility emits far more radioactivity than the
average research or test reactor for which data are available. The 47
reactors for which argon-41 data are available report an average annual
release of 291 curies of argon-41. If we assume that each of the 23
reactors for which emission data are not available emits 291 curies of
argon-41 per year, then the total annual emissions from all 70 reactors
is 20,400 curies of argon-41. The reference facility emits 48 percent
of this total. Assuming that the reference facility also causes 48
percent of the total health impact, that impact is 0.2 fatal cancers
per year.
3.1-2
-------
3.1.8 Existing Emission Standards and Air Pollution Controls
Research and test reactors licensed by NRC are subject to the
requirements of 10 CFR 20.106, which places limits on air emissions to
unrestricted areas. Argon-41 is limited to an air concentration of
4 x 10~° microcuries per milliliter above background, and tritium is
limited to an air concentration of 2 x 10"' microcuries per
milliliter.
3.1-3
-------
Table 3.1-1. Radionuclide emissions from research and test reactors
Design
type
1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22.
23.
24.
25.
26.
27.
28.
29.
30.
Heavy water
Tank
Heavy water
Heavy water
Pool
Pool
Pool
Pool
Pool
TRIGA
TRIGA
Pool
TRIGA
TRIGA
TRIGA
Pool
Pool
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
TRIGA
Power
(kW)
10,000
10,000
5,000
5,000
5,000
2,000
2,000
2,000
2,000
1,500
1,500
1,000
1,000
1,000
1,000
1,000
1,000
1,000
1,000
1,000
1,000
1,000
1,000
250
250
250
250
250
250
250
Radionuclide
Argon-41
Tritium
Argon-41
Tritium
Argon-41
Tritium
Argon-41
Argon-41
Noble gas
Radioiodine
Particulate
Argon-41
Argon-41
Argon-41
Argon-4 1
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Particulate
Argon-41
Argon-4 1
Argon-41
Argon-41
Argon-41
Argon-41
none
Argon-41
None
Emissions
(Ci/y)
465.0
155.0
2504
16.3
N/A
9700
8
350
247.0
47.3
0.021
0.01
N/A
6.64
0.09
2.1
N/A
9.2
7.15
2.9
14.03
9.96
41.7
2.41
0.001
2.6
1.8
1.231
1.0
0.003
0.016
0.0
0.06
N/A
N/A
0.0
See footnote at end of table.
3.1-4
-------
Table 3.1-1.
Radionuclide emissions from research and test reactors
(Continued)
Design
type
31. TRIGA
32. TRIGA
33. TRIGA
34. TRIGA
35. Pool
36. Graphite/water
37. Light water
38. TRIGA
39. TRIGA
40. Graphite/water
41. Graphite/water
42. Graphite/water
43. TRIGA
44. Special
45. TRIGA
46. Graphite/water
47. Pool
48. Pool
49. Pool
50. Homogeneous
51. Pool
52. Special
53. Special
54. Tank
55. Homogeneous
56 . Homogeneous
57 . Homogeneous
58. Homogeneous
59. Homogeneous
60. Homogeneous
6 1 . Homogeneous
62 . Homogeneous
63. Homogeneous
64. Tank
65. Homogeneous
66. Homogeneous
67 . Homogeneous
68. Pool
69. Pulse
70. Pulse
Power
(kW)
250
250
250
250
200
100
100
100
100
100
100
100
18
10
10
10
10
10
10
3
1.0
1.0
0.1
0.1
0.015
0.01
0.01
0.006
0.005
0.005
0.0001
'0.0001
0.0001
0.0001
0.0001
0.0001
0.0001
0.0001
N/A
N/A
Radionuclide
Argon-41
Tritium
none
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Argon-4 1
Argon-41
none
none
none
none
none
none
Krypton-85
none
none
none
none
none
none
Argon-41
Emissions
(Ci/y)
0.002
N/A
0.002
0.0
3.1
33
N/A
0.001
N/A
68.2
113
17
0.3
0.0
0.0
N/A
N/A
N/A
N/A
0.0
N/A
N/A
0.0
N/A
0.0
0.0
3E-8
* *
N/A
0.0
0.0
0.0
_ _ / .
N/A
OA
.0
__ / »
N/A
0.0
M / 4
N/A
vr / A
N/A
»T / A
N/A
0.0
13.31
N/A Not available.
3.1-5
-------
Table 3.1-2. Reference facility
mmmiii1i~~il~~*l~^miiii^iiiiimma^m^^^^—^^^^^^^^^f^^^^mi^****^*^*^^^**Hfi^*ii^***—Hiiiiiiiiiiiilv*^mmm
Parameter Value
Type Heavy water reflected
university reactor
Power level 5,000 KW
Stack height 50 meters
Emissions
Argon-41 9700 Ci/y
Tritium 8 Ci/y
Table 3.1-3. Radiation dose rates from radionuclide emissions
from the reference facility
n Maximum individual Regional population
(mrem/y) (person-rem/y)
Whole body 1.0 343
Table 3.1-4. Fatal cancer risks due to radionuclide emissions from
the reference facility
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Research and test
reactor 2E-5 0.1
3.1-6
-------
3.2 Accelerators
3.2.1 General Description
Accelerators are devices for imparting high kinetic energies to
charged particles (such as electrons, alpha particles, protons, and
deuterons) by electrical or magnetic fields. In a typical operation,
the accelerated particles travel in an evacuated tube or enclosure.
The particles impinge on a metallic or gaseous target, producing
secondary radiation.
There are 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.
Constant DC field machines (also called "Potential-drop" machines)
operate at very high voltages, establishing an electric field of
constant strength through which charged particles are accelerated
toward the target. These accelerators are named according to the power
supply used to generate the high DC voltage. The principal design
types are the Van de Graaff, Cockcroft-Walton, Dynamitron, resonant
transformer and insulating core transformer.
Incremental acceleration machines are accelerators whose electric
field strength varies with time. This type of accelerator increases
particle velocity in a nonlinear manner as the particle moves through
the varying field. The principal design types are the linear
accelerator (linac) and the cyclotron.
A magnetic field accelerator uses a time-varying magnetic field to
generate an electric field which accelerates the particles. The only
current example of this category is the betatron, which is used to
accelerate electrons.
Accelerators have a variety of applications, including
radiography, activation analysis, food sterilization and preservation,
industrial processing, radiation therapy, and research. In 1977 the
Bureau of Radiological Health (BRH78) estimated that there were over
1100 accelerators in use in this country, not including Federally-owned
accelerators. All of the very high energy physics research
accelerators are owned by the Department of Energy and are briefly
discussed in Chapter 2.
Of the total number of accelerators in use, the percentage of each
design types is as follows: linacs, 50 percent; neutron generators (of
several different designs), 17 percent; Van de Graaff, 15 percent;
3.2-1
-------
resonant and insulating core transformers, 6 percent; betatrons, 6
percent; cyclotrons, 3 percent; Cockcroft-Walton, 3 percent. Linacs
are the most widely used machines, about 70 percent being used in
medical applications.
3.2.2 Process Description
Radioactive emissions associated with accelerator operation are
produced by two principal mechanisms: (1) the activation of air by
accelerated particles or secondary radiation, resulting in radioactive
carbon, nitrogen, oxygen, or argon; and (2) the loss of radioactive
material (most frequently tritium) from a target into the air.
The principal air activation reactions are shown in Table 3.2-1.
The formation of carbon-11, nitrogen-13, and oxygen-15 requires, at a
minimum, certain threshold energies which are also listed in Table
3.2-1. These products would not be formed by accelerators which
operate at low energies (typically, under 10 MeV).
Carbon-14 and argon-41 are produced by reactions involving the
absorption of a neutron. The amount of radionuclides formed is in
direct proportion to the neutron flux around the accelerator.
3.2.3 Control Technology
Control of air-activation products with short half-lives can be
accomplished by delaying the venting of the room air. Several
accelerators are capable of such holdup, but they do not use holdup as
an emission control during normal operations. There are no controls in
use to reduce tritium emissions.
3.2.4 Radionuclide Emissions
Table 3.2-2 gives estimated annual radioactive emissions from
three reference facilities. These values were taken from a previous
EPA study of these facilities (EPA79).
3.2.5 Reference Facilities
Table 3.2-3 shows the operating parameters of the three reference
accelerator facilities. The three facilities are typical of
accelerators in use today. The reference facility emissions are taken
from Table 3.2-2.
3.2.6 Health Impact Assessment
The estimated annual radiation doses from the three reference
particle accelerators are shown in Table 3.2-4. The individual
lifetime risks and expected fatal cancers are shown in Table 3.2-5.
3.2-2
-------
3.2.7 Total Health Impact
The estimated total number of fatal cancers caused by all non-DOE
accelerators is 7E-5 per year. This was calculated using the
information in Table 3.2-5 and assuming that there are currently 1,500
such accelerators in operation and that 50 percent of them are linacs,
3 percent are cyclotrons, and 47 percent are constant DC field
machines. The three reference facilities were assumed to be
representatives of these three categories.
3.2.8 Existing Emission Standards and Air Pollution Controls
Accelerators are regulated by the individual States. All of the
States have adopted the Radiological Concentration Guides given by the
Nuclear Regulatory Commission in 10 CFR Part 20. These guides do not
cover isotopes with very short half-lives. The guides for carbon-14,
argon-41, and tritium are: 1E-7 microcuries/ml, 4E-8 microcuries/ml
and 2E-7 microcuries/ml, respectively.
3.2.9 Supplemental Control Technology
Emissions of the air activation products could be reduced by the
use of holdup systems. However, tritium, which dominates the total
health effects, cannot be controlled by holdup due to its 12 year
half-life. Experimental tritium control systems include adsorption on
charcoal and cryogenic distillation, but these systems have not been
commercially demonstrated.
3.2-3
-------
Table 3.2-1.
Nuclear reactions responsible for some airborne
radioactivity
Reaction
(Y,n)
(Y,n)
(Y,n)
(n,2n)
(n,2n)
(n,2n)
(n,p)
(P.pn)
(n,a)
(n,Y)
Parent
nuclide
Nitrogen-14
Oxygen-16
Carbon-12
Nitrogen-14
Oxygen-16
Carbon-12
Nitrogen-14
Oxygen-16
Nitrogen-14
Oxygen-17
Argon-40
Radionuclide
produced
Nitrogen-13
Oxygen-15
Carbon-11
Nitrogen-13
Oxygen-15
Carbon-11
Carbon-14
Oxygen-15
Nitrogen-13
Carbon-14
Argon-4 1
Threshold
energy
(MeV)
10.5
15.7
18.7
11.3
18.0
20.0
NA
10.0
10.0
NA
NA
Half-
Life
10 m
2 m
20 m
10 m
2 m
20 m
5730 y
2 m
10 m
5730 y
1.9 h
NA Not applicable.
m minutes
h hours
y years
3.2-4
-------
Table 3.2-2. Estimated annual emissions from typical
particle accelerators (EPA79)
Radio-
nuclide
Carbon-11
Nitrogen-13
Oxygen-15
Tritium
Carbon-14
Argon— 41
100 MeV
Cyclotron
(Ci)
2.0E-3
4.0E-2
1.0
18 MeV
Electron 6 MeV
Linac Van de Graaff3
(Ci) (Ci)
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.
Table 3.2-3. Reference accelerator facilities
Parameter
Value
Type of accelerator;
Emissions control:
Stack characteristics:
Height
6 MeV Van de Graaff with
tritium target—operating
3000 h/y
18 MeV electron linac
operating 2000 h/y
100 MeV research cyclotron
operating 1000 h/y
None
16.8 meters (roof type)
3.2-5
-------
Table 3.2-4. Annual radiation doses due to radioactive
emissions from typical accelerators (EPA80)
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
Population
(person-rem/y)
5.9E-4
3.1E-7
5.1E-6
Table 3.2-5. Individual lifetime risks and number of fatal cancers
due to radioactive emissions from typical accelerators (EPA80)
Type of
accelerator
Individual lifetime risk
Maximum
individual
Expected fatal cancers
per year of operation
(Fatal cancers)
6 MeV
Van de Graaf
18 MeV
Electron linac
100 MeV
Research
Cyclotron
2E-9
6E-13
1E-9
1E-7
6E-11
1E-9
3.2-6
-------
REFERENCES
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.
EPA79 Environmental Protection Agency, A Study of Radioactive
Airborne Effluents from Particle Accelerators, Technical Note,
ORP/TAD-79-12, Washington, D.C., August 1979.
EPA80 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States —
Preliminary Report, EPA 520/7-79-006, Office of Radiation
Programs, EPA, Washington, D.C., Reprinted 1980.
3.2-7
-------
3.3 Radiopharmaceutical Industry
3.3.1 General Description
Increasing medical and research demands for radioactive chemicals
have resulted in the evolution of a large radiopharmaceutical
industry. This industry comprises the suppliers that produce or
package radiopharmaceuticals, the users of radiopharmaceuticals, and
waste-receiving facilities. Suppliers include manufacturers and
nuclear pharmacies. Manufacturers include companies that manufacture
radionuclides for use as raw materials by other radiopharmaceutical
companies, and companies that process radionuclides into radio-
pharmaceuticals and radioimmunoassay (RIA) kits (TRI79). Nuclear
pharmacies obtain bulk amounts of radiopharmaceuticals and repackage
them for distribution.
Users include hospitals and private physicians that dispense
Pharmaceuticals and medical and research laboratories that utilize RIA
materials. Of all users, hospitals contribute the most airborne
radioactivity because most nuclear medicine procedures are performed at
hospitals.
Waste-receiving facilities that receive wastes from suppliers and
users of radiopharmaceuticals have the potential to produce airborne
emissions of radionuclides. These facilities include incinerators and
sewage treatment plants. It is estimated that more than 90 percent of
the airborne radioactive emissions from waste-receiving facilities are
from sewage treatment plants (TRI79).
Suppliers
Industrial suppliers produce 65 different, generally-used
radionuclides (EPA80). Major suppliers of radiopharmaceuticals and
medical isotopes are listed in Table 3.3-1 (TRI79). This list does not
include nuclear pharmacies.
Iodine-131, iodine-125, xenon-133, and technetium-99m have been
identified as the radionuclides having the greatest potential for
release as airborne effluents from radiopharmaceutical suppliers (Le79),
Users
Radionuclides are extensively used for medical diagnosis, therapy,
and research. The number of medical facilities using radioactive
materials has grown from 38 in 1946 to over 10,000 NRC and Agreement
State licensees in 1977. In 1977 alone, it is estimated that there
were 15 million in-vivo and 20 million in-vitro therapeutic and
diagnostic procedures performed at costs of about $48 million for
in-vivo sales and about $105 million for in-vitro sales (TRI79).
Radionuclides used in diagnostic and therapeutic procedures are listed
in Table 3.3-2 (EPA80).
3.3-1
-------
Table 3.3-1. Major suppliers of radiopharmaceuticals and medical
isotopes, excluding nuclear pharmacies (TRI79)
Location
Supplier
Product
California
Emeryville
Glendale
Vallecitos
Van Nuys
San Ramon
Davis
Irvine
Ri chmond
Florida
Miami Lakes
Georgia
East Point
Illinois
North Chicago
Medi-Physics, Inc.
(home office)
Medi-Physics, Inc.
General Electric Company
Nuclear Med. Svcs.,Inc.
Gammaceutics
University of California
ICN Pharmaceuticals
Bio-Rad Laboratories
Medi-Physics, Inc.
Medical Research
Foundation, Inc.
Abbott Laboratories
Arlington Heights Amersham Corporation
Rosemont
Medi-Physics, Inc.
Indium-Ill, Iodine-123,
Gallium-67, Rubidium-81/
Krypton-Sim generators,
Xenon-133, Technetium-99m.
Technetium-99m-
labeled compounds.
Xenon-133.
Groups I, II, & IV
Iodine-123
Iodine-123
RIA
(a)
Iodine-125, Cobalt-57,
RIA kits.
Technetium-99m-
labeled compounds.
Yttrium-90 microspheres.
Molybdenum-99/
Technetium-99m generators.
Kits for preparation of
Tc-99m labeled compounds.
Cobalt-58 as cyanocobalamin,
Selenium-75 as
selenomethionine,
Iodine-125 as fibrinogen.
Technetium-99m as per-
technetate. Kits for
preparation of Tc-99m
labeled material.
See footnotes at end of table.
3.3-2
-------
Table 3.3-1. Major suppliers of radiopharmaceuticals and medical
isotopes, excluding nuclear pharmacies (TRI79)—continued
Location
VBBBBBBBB^HBBBBBBBHB^^
Indiana
Indianapolis
Elkhart
Massachusetts
Billerica
Supplier
^^H«^^^^^*WM
Bio-Dynamics
Miles Laboratories
Ames Company
Cambridge Nuclear Radio-
pharmaceutical Corp.
New England Nuclear Corp.
Attleboro Falls Gamma Diagnostics Lab.
Boston
Bedford
Minnesota
St. Paul
New England Nuclear Corp.
Radiopharmaceutical Div.
CIS Radiopharmaceuticals,
Inc.
Minnesota Mining &
Manufacturing Co.
Product
Kits for preparation
Tc-99m-labeled DTPA^
and pyrophosphate.
Iodine-125 RIA kits.
Kits for preparation of
Tc-99m-labeled
pyrophosphate, DTPA.
Thallium-201, Gallium-67,
Iodine-131, Iodine-125
Selenium-75, Phosphorus-32,
Mo-99/Tc-99m generators.
Tc-99m as pertechnetate,
sulfur colloid, aggregated
albumin.
Organic compounds labeled
with Tritium, Carbon-14,
Phosphorus-32, and Sulfur-35
Kits for preparation of
Tc-99m-labeled DTPA, albumin,
pyrophosphate, sulfur colloid,
and aggregated albumin.
Kits for preparation of
Tc-99m-labeled materials.
Ytterbium-169 as DTPA.
Missouri
St. Louis
Columbia
Mallinckrodt, Inc.
Diagnostic Products Div.
University of Missouri
Kits for preparation of
Tc-99m-labeled materials;
Chromium-51, Iron-59,
Mercury-197, Iodine-125,
Phosphorus-32, Selenium-75,
Mo-99/Tc-99m generators.
Molybdenum-99 (as raw
material).
See footnotes at end of table.
3.3-3
-------
Table 3.3-1. Major suppliers of radiopharmaceuticals and medical
isotopes, excluding nuclear pharmacies (TRI79)—continued
Location
Supplier
Product
New Jersey
Princeton
E.R. Squibb & Sons, Inc. Kits for preparation of
Tc-99m-labeled materials,
Gold-198, Chromium-51,
Mercury-197, Iodine-131,
Iodine-125, Phosphorus-32,
Selenium-75, Strontium-85,
Cobalt-60, Mo-99/Tc-99m
generators.
S. Plainfield
Ohio
Cincinnati
Medi-Physics, Inc.
Iodine-123, Gallium-67, Tc-99m,
Indium-111, Rb-81/Kr-81m
generators.
Procter and Gamble Co. Kits for preparation of
Technetium-99m, disodium
etidronate.
New York
Tuxedo
Virginia
Richmond
Union Carbide Corp.
Va. Commonwealth Univ.
Tc-99m, Xenon-133, Iodine-131,
Iodine-125, Mo-99/Tc-99m
generators.
Kits for preparation of
Tc-99m-labeled materials,
sulfur colloid, aggregated
albumin
(a)
(b)
See 10 CFR 35.100, Schedule A.
RIA Radioimmunoassay.
^C'DTPA Diethylenetriamine pentaacetic acid.
Iodine-131, iodine-125, xenon-133, and technetium-99m have been
identified as having the greatest potential for release as airborne
effluents from medical facilities. Although releases are much more
likely if the nuclide is easily volatilized, technetium-99m is included
because of the large quantities used in hospitals. Xenon is used
primarily in diagnostic procedures with approximately 62 percent used
in large hospitals (over 500 beds).
3.3-4
-------
Table 3.3-2. Major radiopharmaceuticals and
their uses (EPA80)
Radionuclide
Use
Phosphorus-32
Gallium-67
Rubidium-81
Technetium-99m
Iodine-123
Iodine-125
Iodine-131
Xenon-133
Mercury-203
Thallium-201
Bone marrow therapy
Tumor localization
Myocardial imaging
Bone imaging, brain imaging, liver
imaging, lung perfusion, myocardial
imaging, blood pool, renograms,
thyroid imaging, thyroid uptake
renal imaging
Thyroid imaging
Thyroid uptake
Renograms
Renal imaging, renograms, thyroid
imaging, thyroid uptake, tumor
localization and therapy
Lung ventilation
Renograms
Myocardial imaging
3.3-5
-------
Iodine is used for diagnostic and therapeutic procedures with
approximately 60 percent used in large hospitals. Estimated
quantities of radionuclides received and used by hospitals in 1977 are
listed in Table 3.3-4.
Table 3.3-4. Estimated quantities of radionuclides received and
used by hospitals, 1977
Quantity (Ci)
Radionuclide Received Used
Iodine-131 900-1500 300-1350
Xenon-133 2,700-3,300 1,600-2,000
Technetium-99m 26,000-34,000 15,000-30,000
Waste-Receiving Facilities
Most of the radionuclides used at medical facilities are released
via the liquid pathway to the sanitary sewer system. When sewage and
sludge containing this material are treated in a sewage treatment
plant, radionuclides may be emitted into the air.
Iodine-131, iodine-125, and technetium-99m have the greatest
potential for release as airborne effluents from sewage treatment
plants (TRI79).
3.3.2 Process Description
Radionuclides used in the radiopharmaceutical industry are
produced by irradiation of target materials (or fuel) in a reactor or
accelerator, and by radioisotope generators.
Suppliers
Radionuclide manufacturing involves complex chemical processes
that have the potential for releasing radioactive materials to the
environment. Most radionuclides produced for use in the industry are
made in nuclear reactors by one of the reactions shown in Table 3.3-5.
The most common of these is the neutron-gamma reaction because many
elements capture neutrons easily. It is estimated that
reactor-produced isotopes account for 60 to 80 percent of the market
(TRI79).
3.3-6
-------
Table 3.3-5. Nuclear reactions used in radioisotope production
Reaction Examples
(1) Neutron-gamma (n,Y) 59Co + n •* 60Co + y
(2) Neutron-proton (n,p) 32S + n-> 32p + p
(3) Neutron-alpha (n,a) 35C1 + n _,. 32p + a
In a reactor, the main steps in radionuclide production are as
follows (EPA80):
1. A suitable target is prepared and irradiated with
neutrons.
2. The irradiated target is processed by dissolution
or by more complicated separations (including ion
exchange, precipitation, and distillation) to remove
undesirable impurities, or to concentrate the product
nuc1ide.
3. Radionuclides are placed in inventory, dispensed,
and packaged for shipment.
Many radionuclides are produced in particle accelerators, such as
the cyclotron. Amounts of radioactive materials produced in
accelerators are smaller than amounts produced in reactors.
The cyclotron can be used to produce nuc1ides having decay
characteristics that are preferable to other isotopes of the same
element that are produced in reactors and isotopes of elements for which
no reactor-produced nuclides exist. Examples of accelerator-produced
radionuclides are iodine-123, iron-52, mercury-199m, carbon-11,
nitrogen-13, and oxygen-15.
Typical nuclear pharmacy production activities include processing,
mixing or compounding, and distribution of prepared radiopharmaceuticals.
There is a growing trend for nuclear pharmacies to operate
radioisotope generators for the production of certain radionuclides
having short half-lives; for example, technetium-99m. Radioisotope
generators make nuclides with short half-lives available at long
distances from the source of production. These generators consist of a
longer-lived parent nuclide that produces the short-lived daughter as it
decays. In the generator, the daughter nuclide is chemically separated
at intervals, leaving the parent nuclide to generate more of the
daughter.
3.3-7
-------
Users
In hospitals, radionuclides are generally handled in solid or
liquid form, except for some radioactive gases, notably xenon. This
tends to decrease the likelihood of release of airborne effluents.
Therapeutic iodine-131, generally in the form of sodium iodide, is
readily volatilized, and can become an airborne contaminant when used
in some therapeutic procedures.
Xenon-133 can also be released as an airborne effluent. Because
of a low biological half-life, relatively large amounts are
administered for lung-imaging procedures. Following administration,
patients exhale xenon-133 gas into a spirometer. The exhaust from this
instrument exits the hospital through a roof vent, with or without
treatment.
Technetium-99m is used in large quantities in hospitals, and is
obtained directly from the manufacturer or from the nuclear pharmacy
where it is produced in a radioisotope generator from molybdenum-99.
Although not a gaseous or volatile isotope, technetium-99m is a
potential airborne effluent because of the quantities used in nuclear
medicine procedures.
Waste-Receiving Facilities
Radionuclide releases at sewage treatment plants depend upon
several factors. The chemical and physical properties of wastewater
and sludge influence the potential amount of radioactivity released;
e.g., the potential for release is greater at points in the treatment
process where wastewater pH is acidic. Other factors that affect
radionuclide releases include decay losses, evaporative losses, solids
removal, degree of system retention, and dilution.
Sludge treatment processes (drying and incineration) are the
greatest sources of radionuclide emissions from sewage treatment plants
because the high temperatures employed in these processes (typically
725°C) volatilize iodine and technetium. In addition, sludge
incineration has the smallest time delay compared with other sludge
treatment processes, and the greatest potential for release of
particulates caused by mechanical agitation of ash and combustion gases
in the incinerator (TRI79).
It is estimated that approximately 21 percent of the sewage
treatment facilities in the U.S. employ incineration or pyrolysis for
sludge treatment (EPA80). In a treatment facility, sludge is typically
concentrated in settling tanks before it is concentrated further in
another sludge treatment process (e.g., centrifugation). Following
this process, the sludge is conveyed to an incinerator and burned at
temperatures up to 815°C.
3.3-8
-------
3.3.3 Control Technology
Types of effluent controls employed by producers of
radiopharmaceuticals depend on the type and amount of each nuclide
handled in the facility (Le80). All suppliers handling large amounts
of iodine, and some dealing in smaller quantities, handle this material
in hot cells or fume hoods that exhaust through HEPA and/or activated
carbon filters before release through a roof-mounted vent stack. Some
suppliers that handle small amounts of radioiodine, or only nonvolatile
nuclides such as molybdenum and technetium, use.no filters, or only
HEPA filters on fume hoods and building ventilation exhausts. This
exhaust is usually released from a short vent stack (2 to 3 m high) on
top of the building (TRI79). Xenon manufacturers generally use
ventilation controls only. One large producer controls radioactive
xenon emissions by cryogenically liquefying hot cell off-gas, and
holding it for decay.
Small hospitals (less than 300 beds) generally operate with no
effluent controls because the total activity of the principal isotope
used (technetium-99m) is low, and because it is handled in solution.
Hospitals in the medium-size range (300 to 500 beds) generally use
xenon traps and unfiltered fume hoods, but may use controls similar to
those of the larger hospitals if large amounts of activity are handled
daily. Some hospitals capture patient xenon exhalations for holdup in
retention bags before release. Other medium-size hospitals may have no
controls if radiopharmaceuticals are administered infrequently, or if
their emissions meet NRC MFC requirements without controls. Larger
hospitals (over 500 beds) generally use controls similar to those used
by suppliers because of the large amounts of activity handled, and
because of the variety of radioisotopes used. Controls at large
hospitals range from fume hoods with HEPA and activated carbon filters
and xenon traps or retention bags to unfiltered fume hoods and no xenon
controls (TRI79).
3.3.4 Radionuclide Emission Measurements
Suppliers
Data presented in this section are drawn from emissions data
submitted to EPA by medical isotope producers and from reports of
surveys conducted at several radiopharmaceutical manufacturing firms.
The emissions data represent airborne releases from normal operations
as measured by company-owned or contractor monitoring systems. Average
annual emissions of six radiopharmaceutical suppliers are listed in
Table 3.3-6.
The NRC conducted a survey of over 3000 by-product material
licensees in late 1980 to collect annual radioactive effluent emissions
data (NRC81). Table 3.3-7 summarizes emissions data for 385 industrial
facilities that manufacture radionuclides who responded to the survey.
3.3-9
-------
Table 3.3-6. Radionuclide emissions from six major radiopharma-
ceutical producers (Co82, EPA80, Fra82, Frb82, Roa82, Rob82)
Producing
Plant
A
B
C
D
E
F
Iodine-125
1.8E-2
2.2E-6
-
-
l.OE-2
2.6E-3
Emissions
Iodine-131
3 . 9E-4
-
-
-
7.6E-2
3.1E-2
(Ci/y)
Technetium-99m
-
-
4.14E-3
4.5E-3
-
-
Table 3.3-7. Summary of reported atmospheric emissions of
radionuclides from 385 industrial facilities (NRG81)
Number of
„ facilities
Source
using
nuclide
Iodine-131
Iodine-125
Xenon- 13 3
Molybdenum-99
Technetium-99m
11
55
6
4
2
Number of
facilities
reporting
releases
' 4
25
4
4
1
Emissions (Ci/y)
Mean
1 . 8E-4
1.7E-3
7.0
8.3E-6
3.2E-6
Maximum
4.6E-4
2.0E-2
2.3E+1
3.0E-5
3.2E-6
Minimum
3.0E-5
3.0E-8
2.0E-2
1.5E-7
3.2E-6
Users
The survey conducted by the NRC (NRC81) also included radioactive
emissions data for 860 government and public medical facilities. These
data are summarized in Table 3.3-8.
Sewage Treatment Plants
Radioactive airborne emissions resulting from sludge drying and
incineration at a sewage treatment plant were studied (TRI79) and
estimated to be 5.0E-4 Ci/y for iodine-131 and 8.0E-4 Ci/y for
technetium-99m. This report also estimated that about 4000 sewage
treatment plants in the United States employ these sludge treatment
processes.
3.3-10
-------
Table 3.3-8. Summary of reported atmospheric emissions of radio-
nuclides from 860 government and public medical facilities (NRC81)
Number of
Source facilities
using
nuclide
Iodine-131 346
Iodine-125 270
Xenon-133 229
Molybdenum-99 268
Technetium-99m 73
Number of
facilities
reporting
releases
25
19
142
3
2
Emissions (Ci/y)
Mean
2.9E-3
1.7E-3
4.6E-1
1.0
2.8E-1
Maximum
5 . OE-2
9.5E-3
6.4
3.0
5.0E-1
Minimum
2.0E-8
l.OE-8
2.0E-5
l.OE-8
5.2E-2
3.3.5 Reference Facilities
Radiopharmaceutical Supplier Facility
The radiopharamaceutical supply industry can be characterized as
generally urban, with suppliers located near their major users,
hospitals (TRI79). Table 3.3-9 describes the parameters of a typical
radiopharmaceutical production plant. These parameters were used to
estimate health impacts resulting from emissions from the reference
facility.
The typical facility produces technetium-99m, xenon-133,
iodine-131, iodine-125, and molybdenum-99/technetium-99m generators
(EPA80). Airborne releases are discharged from a single stack.
Atmospheric emissions from the reference facility are listed in Table
3.3-10. Emissions from the reference facility were chosen as equal to
emissions from facilities having the highest values listed in Tables
3.3.6 and 3.3.7.
Emissions from the reference facility are controlled by charcoal
beds and HEPA filters.
User Facility
Parameters that describe the reference medical facility are listed
in Table 3.3-9. These parameters represent a typical large hospital.
It is assumed that the hospital has nuclear medicine capabilities, and
administers an average of 0.5 curies per year of iodine-131, 0.05
curies per year of iodine-125, and 25.0 curies per year of xenon-133.
3.3-11
-------
Table 3.3-9. Reference facilities of typical suppliers and
users of radiopharmaceuticals
Parameter
Value
Supply Facility
Product line:
Emission controls:
Stack parameters:
User Facility
Size:
Volume of administrations;
Emission controls:
Sewage Treatment Plant
Process:
Iodine-131, iodine-125, xenon-133,
technetium-99m, molybdenum-99/
technetium-99m generators.
Activated carbon/HEPA filters with
release through a single elevated stack
Height: 15 meters.
500+ beds
Iodine-131, 0.5 Ci/y
Iodine-125, 0.05 Ci/y
Xenon-133, 25.0 Ci/y
Exhaust hoods with carbon and HEPA
filters. Release through building
ventilation roof vents.
Vent height: 10 m.
Sludge drying and incineration
Estimated annual atmospheric emissions from the reference medical
facility are listed in Table 3.3-10. These emission estimates
represent maximum emission levels for 1-131, 1-125, and Xe-133 from
sources described in Table 3.3-8. Although molybdenum-99 and
technetium-99m are used at the reference facility, releases are assumed
to be zero because, as indicated in Table 3.3-8, airborne releases are
rarely observed for these nuclides.
Sewage Treatment Facility
The reference sewage treatment plant dries and incinerates
sludge. Atmospheric emissions from a typical sewage treatment plant
that employs these processes are listed in Table 3.3-10. These
emission estimates are based on a study of airborne emissions from a
sewage treatment plant (TRI79).
3.3-12
-------
Table 3.3-10. Radionuclide emissions from reference
radiopharmaceutical industry facilities
Source/Radionuclide Emissions
(Ci/y)
Supply Facility
Iodine-125 2.0E-2
Iodine-131 7.6E-2
Xenon-133 2.3E+1
Technetium-99m 4.5E-3
User Facility
Iodine-125 9.5E-3
Iodine-131 5.0E-2
Xenon-133 6.4
Sewage Treatment Plant
Iodine-131 5.0E-4
Technetium-99m 8.0E-4
3.3.6. Health Impact Assessment of Reference Radiopharmaceutical
Industry Facilities
The estimated annual radiation doses from radionuclide emissions
from the reference radiopharmaceutical supply facility, medical
facility, and sewage treatment plant are listed in Table 3.3-11. These
estimates are for the near suburbs of a large midwest city with a
regional population of 2.5 million (Reference Site B). The maximum
exposed individuals are located 500 meters from the supply facility,
500 meters from the medical facility, and 500 meters from the sewage
treatment plant.
Table 3.3-12 presents estimates of the maximum individual lifetime
risks and the number of fatal cancers to the regional population from
these doses.
3.3.7 Total Health Impact of the Radiopharmaceutical Industry
For all segments of the radiopharmaceutical industry, the
estimated total health impact is 9E-4 fatal cancers per year. The
following analysis details how this estimate was obtained.
Suppliers
The estimated total health impact caused by all radiopharma-
ceutical suppliers is based on the assumptions that (1) emissions of
3.3-13
-------
Table 3.3-11. Radiation dose rates from radionuclide emissions
from the reference radiopharmaceutical industry facilities
„ Maximum individual Regional population
rgan (mrem/y) (person-rem/y)
Radiopharmaceutical supplier
Thyroid 4.7E-1 3.3
Medical facility
Thyroid 3.6E-1 1.7
Sewage treatment plant
Thyroid 8.0E-4 7.4E-3
Table 3.3-12. Fatal cancer risks due to radionuclide emissions from
the reference radiopharmaceutical industry facilities
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Radiopharmaceutical supplier
2E-7 2E-5
Medical facility
2E-7 6E-6
Sewage treatment plant
2E-10 2E-8
1-125, 1-131, Xe-133, and Tc-99m reported for industrial facilities in
a survey by NRC (NRC81) are from radiopharmaceutical suppliers; and (2)
the number of industrial licensees in non-agreement states, for which
data were available, is approximately equal to the number of licensees
in agreement states.
3.3-14
-------
Data presented in the NEC survey (NRC81) showed that approximately
15% of industrial licensees in the survey handled 1-125, 3% handled
1-131, and less than 2% handled Xe-133 and Tc-99m. Based on these
figures and the above assumptions, the total numbers of suppliers in
the U.S. handling 1-125 and 1-131 are 328 and 66, respectively.
Although the number of suppliers handling Xe-133 and Tc-99m would be
less than 44, this figure will be used for estimation purposes.
Assuming that available average emissions data (Tables 3.3-6 and
3.3-7) are typical of the entire industry, total annual emissions from
all radiopharmaceutical suppliers are as follows: 1-125, 0.82 Ci/y;
1-131, 0.99 Ci/y; Xe-133, 310 Ci/y; and Tc-99m, 0.13 Ci/y.
Based on these emissions, releases from the reference facility
(Table 3.3-10) are 2.4% of the national total for 1-125, 7.7% for
1-131, 7.4% for Xe-133, and 3.5% for Tc-99m. Assuming that the
reference facility also causes equal percentages of total health
impact, the impact from all radiopharmaceutical suppliers is 3E-4 fatal
cancers per year. Contributions to this figure from each radionuclide
are as follows: 1-125, 1E-4; 1-131, 4E-5; Xe-133, 1.7E-4; and Tc-99m,
3E-8.
Users
Assuming that the number of medical facility licensees in
non-agreement and agreement states is approximately equal, data in the
NRC survey (NRC81) indicate that approximately 1,100 facilities in the
U.S. use 1-125, 1,200 facilities use 1-131, and 800 use Xe-133.
If the average emissions listed in Table 3.3-8 are assumed to be
typical of all medical facilities, total annual emissions from all
medical facilities are as follows: 1-125, 1-9 Ci/y; 1-131, 3.5 Ci/y;
and Xe-133. 370 Ci/y-
Emissions from the reference facility contribute 0.5% of the total
1-125 emission, 1.4% of the total 1-131 emission, and 1.7% of the
Xe-133 emission. Assuming that the reference facility contributes
equal percentages to the total health impact, the impact from all
medical facilities is 5E-4 fatal cancers per year. Contributions to
this figure from each radionuclide are as follows: 1-125, 2E-4; 1-131,
1E-4; and Xe-133, 2E-4.
Sewage Treatment Plants
It has been estimated that approximately 4000 sewage treatment
plants in the U.S. employ sludge incineration or pyrolysis (TRI79).
Assuming that emissions from the reference facility are typical of
emissions from all sewage treatment plants that incinerate sludge, the
total annual emissions of 1-131 and Tc-99m are 2.0 Ci/y and 3.2 Ci/y,
respectively.
The total health impact from all sewage treatment plants is 6E-5
fatal cancers per year, which reflects contributions of 6E-5 and 8E-7
from 1-131 and Tc-99m, respectively.
3.3-15
-------
3.3.8 Existing Emission Standards and Air Pollution Controls
Suppliers and users of radiopharmaceuticals are NRC licensees and
are therefore required to limit effluent releases to unrestricted areas
to the maximum permissible concentrations of 10 CFR 20, Appendix B,
Table II. There are no radionuclide emission standards for sewage
treatment plants.
3.3.9 Supplemental Control Technology
Suppliers
Existing emission controls typically employed at supplier
facilities (HEPA and carbon beds/filters) effectively remove
particulates and radioiodines, but not radioactive noble gases.
Supplemental methods for controlling noble gas releases include
cryogenic systems and hold-up tanks. The performance of cryogenic
systems in large commercial facilities has not yet been demonstrated,
nor is there an approved disposal method for the concentrated,
potentially long-lived, high-activity wastes that these systems produce
(TRI79). Hold-up tanks are best suited to effluents with low release
rates which contain short-lived noble gases.
Because the entire volume of effluent must be retained to allow
for decay, hold-up is feasible only at very low release rates. Since
exhaust rates at supplier facilities typically are in the range of
10-* to 10° liters per minute, the tanks required for hold-up would
be too large and too costly to be practical. Implementation of
supplemental controls for noble gas control at supply facilities is,
therefore, not currently practicable.
Users
Xenon retention bags, which are now in use at some medical
facilities, are a feasible means of reducing radioactive emissions
because of low release rates of xenon-133. The costs and risk
reductions achieved by adding supplementary controls to capture patient
xenon exhalations at the reference medical facility are shown in Table
3.3-13.
Sewage Treatment Plants
Sewage treatment plants employing sludge incineration typically
use dry cyclones and wet scrubbers to control gaseous and particulate
emissions. Supplementary controls consist of charcoal filters to
reduce iodine emissions and HEPA filters to reduce particulate
emissions of technetium. HEPA filters are required upstream of the
charcoal filters to prevent plugging.
Costs and risk reductions achieved by adding these supplementary
controls to the incinerator stacks of the reference sewage treatment
plant to reduce iodine-131 and technetium-99m emissions are shown in
Table 3.3-13.
3.3-16
-------
Table 3.3-13. Costs and risk reductions of adding supplemental
controls to reference radiopharmaceutical industry facilities
Type
of
control
Level of
control
Annual
cost
($1000)
(a)
Fatal cancer risks
Individual
lifetime Fatal
risk cancers/y
Medical facility
No xenon
controls^) 0
Add retention
bags or xenon
traps 99.9
Sewage treatment plant
Dry cyclone and
scrubber(b)
Add HEPA filter
with preheater 99
and charcoal filter 90
(c)
(d)
25.0
50.0
2E-7
1E-7
2E-10
2E-11
6E-6
3E-6
2E-8
2E-9
(a)
Does not include capital costs.
^'Typical existing controls.
(cParticulates.
Iodines.
3.3-17
-------
REFERENCES
Co82 Cole L. W. , Environmental Survey of the Manufacturing Facility,
Medi-Physics, Inc., Arlington Heights, Illinois, Oak Ridge
Associated Universities, Oak Ridge, Tennessee, January 1982.
EPA80 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United
States — Preliminary Report, EPA 520/7-79-006, Office of
Radiation Programs, EPA, Washington, D.C., Reprinted 1980.
Fra82 Frame P. W. , Environmental Survey of the New England Nuclear
Corporation, Billerica, Massachusetts, Oak Ridge Associated
Universities, Oak Ridge, Tennessee, April 1982.
Frb82 Frame P. W. , Environmental Survey of the New England Nuclear
Corporation, Boston, Massachusetts, Oak Ridge Associated
Universities, Oak Ridge Tennessee, April 1982.
Le79 Leventhal L. , et al., Radioactive Airborne Effluents from the
Radiopharmaceutical Industry, in Proceedings of the Health
Physics Society, 24th Annual Meeting, Philadelphia, Pa., 1979.
Le80 Leventhal L. , et al., A Study of Effluent Control Technologies
Employed by Radiopharmaceutical Users and Suppliers, in: Book
of Papers, International Radiation Protection Association, 5th
International Congress, Volume II, Jerusalem, Israel, 1980.
NRC81 Nuclear Regulatory Commission, A Survey of Radioactive Effluent
Releases from Byproduct Material Facilities, NUREG-0819, Office
of Nuclear Material Safety and Safeguards, NRG, Washington,
B.C., 1981.
Roa82 Rocco B. P- , Environmental Survey of the Medi-Physics Facility,
South Plainfield, New Jersey, Oak Ridge Associated
Universities, Oak Ridge, Tennessee, January 1982.
Rob82 Rocco B. P., Environmental Survey of the E. R. Squibb and Sons
Facility, New Brunswick, New Jersey, Oak Ridge Associated
Universities, Oak Ridge Tennessee, March
TRI79 Teknekron Research, Inc., Information Base (including Sources
and Emission Rates) for the Evaluation and Control of
Radioactive Materials to Ambient Air, Interim Report, Volume I,
EPA Contract No. 68-01-5142, July 1979.
3.3-18
-------
3.4 Department of Defense Facilities
3.4A Armed Forces Radiobiology Research Institute (AFRRI)
3.4A.1 General Description
The Armed Forces Radiobiology Research Institute (AFRRI) operates
a TRIGA Mark-F pool-type thermal research reactor, and a linear
accelerator (linac) in support of Department of Defense radiation
research. Most of this research involves studies of medical effects of
nuclear radiation and the effects of transient radiation on electronics
and other equipment.
The AFRRI reactor is licensed by the NRC to operate at
steady-state power levels up to 1.0 MW (thermal). This reactor is also
capable of pulse operations, and can produce a 10 msec pulse of about
2500 MW (thermal) at peak power.
AFRRI1s linac typically operates in the 18 to 20 MeV energy range
but is capable of operating at energies up to 30 MeV-
AFRRI is located on the grounds of the National Naval Medical
Center in Bethesda, Maryland, approximately 20 kilometers northwest of
Washington, D.C.
3.4A. 2 Process Description
The AFRRI reactor and accelerator are used for Department of
Defense radiation research. This research includes medical effects of
nuclear radiation, radiobiology, and radioisotope production. AFRRI
facilities have also been used to support Federal criminal
investigations, studies of transient radiation effects on electronics,
and artifact analysis (Sh81).
The reactor core, which is cooled by natural convection, is
located under about 5 m of water, and is movable laterally within an
open cloverleaf-shaped pool. Pool dimensions are 4.2 m across the
major lobes, 3.9 m across the minor lobes, and 5.8 m deep.
Exposure facilities available to users include two separate
exposure rooms, a pneumatic tube transfer system, the pool itself, and
an in-core experiment tube.
Reactor fuel is 8.5 weight percent uranium which has been enriched
to 20 percent uranium-235.
3.4A-1
-------
3.4A.3 Control Technology
Emissions from the AFRRI reactor and accelerator are released to
the atmosphere through a common stack atop the AFRRI building.
Particulate emissions are controlled by a roughing filter, prefilter,
and HEPA filter.
3.4A.4 Radionuclide Emissions Measurements
Annual airborne radionuclide emissions for AFRRI are shown in
Table 3.4A-1. These figures represent average annual emissions for
1981 and 1982.
Table 3.4A-1. Radionuclide emissions from the Armed Forces
Radiobiology Research Institute
Source
AFRRI stack
AFRRI stack
Radionuclide
Argon-41
Nitrogen-13, and
Oxygen- 15
Emissions^
(Ci/y)
1.3
3.5E-2
a)
^a'Average annual emissions for 1981 and 1982.
3.4A.5 Health Impact Assessment of AFRRI
The estimated annual radiation doses resulting from radionuclide
emissions from AFRRI are listed in Table 3.4A-2. The distance from the
AFRRI facility to the nearest residence is approximately 200 meters.
These estimates are for an urban site with a regional population of
2.5E+6 (Reference Site B). The maximum individual is located 500
meters from the AFRRJ facility.
Table 3.4A-3 lists the estimated individual lifetime risks and the
number of fatal cancers to the regional population from these doses.
3.4A.6 Existing Emission Standards and Air Pollution Controls
The AFRRI reactor is licensed by NRC and is therefore subject to
the emission requirements of 10 CFR 20.106. This regulation places
limits on air emissions to unrestricted areas. For argon-41, this
limit is 4 x 10~° microcuries per milliliter above background.
3.4A-2
-------
\
Table 3.4A-2. Radiation dose rates from radionuclide emissions
from the Armed Forces Radiobiology Research Institute
Maximum individual Regional population
(mrem/y) (person-rem/y)
Total body 4.8E-3 1.7E-3
Table 3.4A-3. Fatal cancer risks due to radionuclide emissions from
the Armed Forces Radiobiology Research Institute
Source
Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
AFRRI 1E-7 5E-7
3.4A.7 Supplemental Control Technology
There is no demonstrated treatment technology for control of
emissions of argon-41 from reactors. Reduction of these emissions is
best accomplished by work practice controls; i.e., reducing reactor
operating time.
3.4A-3
-------
REFERENCES
Sh81 Sholtis, J. A. and Moore M. L., Reactor Facility, Armed Forces
Radiobiology Research Institute, AFRRI Technical Report
TR81-2, AFRRI, Bethesda, Md., 1981.
3.4A-4
-------
3.4B U.S. Army Facilities
3.4B.L General Description
The U.S. Army Test and Evaluation Command operates two reactors:
the Army Pulse Radiation Facility (APRF) at Aberdeen Proving Ground,
Maryland, and the Fast Burst Reactor (FBR) at White Sands Missile Range,
New Mexico. These reactors are very similar in design and are used to
support Army and other Department of Defense studies in nuclear radiation
effects.
3.4B.2 Process Description
Both Army reactors are bare, unreflected, unmoderated, and fueled
with enriched uranium. These reactors are capable of self-limiting,
super-prompt-critical pulse operations as well as steady-state operations
at power levels up to 10 kW (EPA80). Operating information for the APRF
and FBR for 1981 is summarized in Table 3.4B-1. The reactors are used
primarily by DOD and defense contractors to study nuclear weapons effects
on electronics and other DOD related equipment.
The White Sands FBR is the principal source of radioactive airborne
emissions from Army reactors. At the FBR, concrete structures around the
reactor reflect and thus lower the energy of neutrons streaming from the
reactor. These low energy neutrons produce airborne radioactivity in the
reactor building by neutron activation of stable argon-40 in air.
Concrete structures at the APRF are farther from the reactor; hence, much
less (essentially zero) argon-41 is produced at this facility (TRI79).
Table 3.4B-1. Number and modes of operations at Army Reactor
Facilities, 1981 (Aa82, ARM81)
Number of operations
Type of operation
Pulse 211 252
Steady State 233 159
Unscheduled Terminations ~ °
Total 444 419
3.4B-1
-------
3. 4B. 3 Control_JTecjmojlogy
Air exhausted from U.S. Army reactor facilities is passed through
HITT>A filters before release to the atmosphere.
3.4B.4 Radionuc1id e Emission Measurements
Radioactive emissions from Army reactors during 1976, 1978, and
1981 are listed in Table 3.4B-2. For the APRF, particulate releases
are reported as gross beta concentrations only. All gaseous releases
from the APRF were below the minimum detectable concentration of 3.0E-3
pCi/m3.
Table 3.4B-2. Radionuclide emissions from Army Pulse Reactors
Radioactive material
Emissions (Ci/y)
APRF
FBR
Gross beta concentration:
1976
1981
Argon-41:
1976
1978
1981
2.8E-6
3.3E-5
11.7
18.0
13.3
Source: (De76, Aaa77, TRI79, AabSl, ARM81).
3.4B.5 Health Impact Assessment from Army Pulse Reactors
The estimated annual radiation doses resulting from radionuclide
emissions from the White Sands FBR are listed in Table 3.4B-3. The
distance to the nearest off site individuals at the APRF and FBR are
approximately 1.6 km and 2.0 km, respectively. The predominant exposure
pathway is that of air immersion. These estimates are for a sparsely
populated southwestern location with a regional population of 3.6E+4
(Reference Site E). The maximum individual is located 500 meters from
the reactor.
Table 3.4B-4 lists the estimated individual lifetime risks and the
number of fatal cancers to the regional population from these doses.
This assessment was made only for the White Sands FBR because nearly
all measured radionuclide emissions from Army reactors originate at the
FBR.
3.4B-2
-------
Table 3.4B-3. Radiation dose rates from radionuclide emissions
from the White Sands Fast Burst Reactor
Organ Maximum individual Regional population
(mrem/y) (person-rem/y)
Endo steal
Spleen
Red Marrow
Muse le
Pulmonary
2.6E-2
2.6E-2
2.4E-2
2.4E-2
2.3E-2
9.2E-2
9.4E-2
8.6E-2
8.7E-2
8.2E-2
Weighted Sum 2.3E-2 8.1E-2
Table 3.4B-4. Fatal cancer risks due to radionuclide emissions from
the White Sands Fast Burst Reactor
Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
FBR 5E-7 2E-5
3. 4B. 6 Existing Emi s sjlo n JS_tatid_ardjs__and_ Air
Because Army pulse reactors are not licensed by NRC, they are not
subject to radionuclide emission standards.
3.4B.7 SupplementalL Control Technology
Emissions from Army pulse reactors consist mainly of argon-41, for
which no demonstrated treatment technology exists. Reduction of argon-41
emissions are best controlled by work practice controls; e.g., reducing
reactor operating time and reducing the amount of air subject to neutron
irradiation by plugging air leaks into the reactor compartment.
3.4B-3
-------
REFERENCES
Aaa77 Aaserude R.A., Dickinson R. W., Dubyoski H. G., and Kazi A. H. ,
APRF, Army Pulse Radiation Facility, Annual Operating Report,
Aberdeen Proving Ground, Md., 1977
Aab82 Aaserude R.A., Dubyoski H. G., Harrell, D.R. and Kazi, A. H.,
Army Pulse Radiation Dividion Reactor, Annual Operating Report,
Materiel Testing Directorate, Aberdeen Proving Ground, Md.,
1982.
ARM81 Army Materiel Test and Evaluation Directorate, White Sands
Missile Range Fast Burst Reactor, Annual Operating Report,
Applied Sciences Division, White Sands Missile Range, N. M.,
1981.
De76 De La Paz A. and Dressel R. W., White Sands Missile Range Fast
Burst Reactor Facility, Annual Operating Report, Army Materiel
Test and Evaluation Directorate, White Sands Missile Range,
N.M., 1976.
TRI79 Teknekron Research, Inc., Information Base (Including Sources
and Emission Rates) for the Evaluation and Control of
Radioactive Materials to Ambient Air, Interim Report, Volume 1,
EPA Contract No. 68-01-5142, July 1979.
3.4B-4
-------
3.4C U.S. Navy Faci1i t ie s
3.4C.1 General Description
Airborne emissions of radionuclides from U.S. Navy facilities are
due, almost entirely, to naval shipyards. Construction, overhaul,
refueling, and maintenance of the 133 submarines and ships of the Navy's
nuclear fleet are performed at nine naval shipyards at the following
locations:
Mare Island Naval Shipyard, Vallejo, California
Electric Boat Division, General Dynamics, Groton, Connecticut
Pearl Harbor Naval Shipyard, Hawaii
Portsmouth Naval Shipyard, Kittery, Maine
Ingalls Shipbuilding Division, Pascagoula, Mississippi
U.S. Naval Station and Naval Shipyard, Charleston, S. C.
Newport News Shipbuilding and Drydock Co., Newport News, Va.
Norfolk Naval Shipyard, Portsmouth, Virginia
Puget Sound Naval Shipyard, Bremerton, Washington
3.4C.2 Process Description
Operations performed at naval shipyards include construction,
startup testing, refueling, and maintenance of the pressurized water
reactors that power the nuclear fleet. Radioactive wastes generated by
these activities are processed and sealed at the shipyards and shipped to
commercial waste disposal sites.
The primary sources of airborne radioactive emissions from naval
shipyards are the support facilities that process and package radioactive
waste materials for shipment to disposal sites. These facilities handle
solid low-level radioactive wastes such as contaminated rags, paper,
filters, ion exchange resins, and scrap materials.
During operation, shipboard nuclear reactors release small amounts
of radioactivity (carbon-14) into the atmosphere; however, most of this
is released at sea, beyond 12 miles from shore (Ri82).
3.4C.3 Control Technology
All air exhausted from radiological support facilities at naval
shipyards is passed through HEPA filters and monitored during discharge.
A comparison of airborne activity measurements in shipyards with
radioactivity concentrations in ambient air indicates that air exhausted
from these facilities actually contains less activity that the intake air
(Ri82).
3.4C-1
-------
3.4C.4 Radionuclide Emission Measurements
Monitoring of effluents from nuclear naval shipyards began in 1963.
To date, this monitoring has shown no concentration of airborne effluents
in excess of naturally occurring background levels (TRI79).
Results of emission measurements taken at Puget Sound Naval Shipyard
in 1974 are shown in Table 3.4C-1. These measurements showed that the
tritium concentration was below the minimum detectable level of 1.0 pCi/1,
and that the level of krypton-85 was within average background levels
(EPA77).
Table 3.4C-1. Radionuclide emissions at Puget Sound Naval Shipyard,
1974
... Emissions
Source Radionuclide , _./,•.
{.pCi/ U
West of Radiological
Support Building Krypton-85 17.4 +_ 10%
Radiological Support
Building Tritium 0.4 _+ 50%
Radiological Support
Building Tritium 0.3 + 66%
3.4C.5 Reference Facility
The typical nuclear shipyard processes, packages, and ships
approximately 85 cubic meters of radioactive solid waste for disposal
annually. The average activity of this material is approximately 6.3
curies. Waste packaging is performed in an enclosed facility, exhaust
from which is passed through HEPA filters before release to the
atmosphere. Air is exhausted from the radiological support facility at a
height of about five meters.
Estimated radioactive emissions from the reference naval shipyard
are listed in Table 3.4C-2. These are conservative, worst-case estimates
used by the Navy in environmental pathways analysis, and are higher than
any measurements made in the past five years at any shipyard (Ri82).
3.4C.6 Health Impact Assessment of the Reference Facility
The estimated annual radiation doses resulting from radionuclide
emissions from the reference shipyard are listed in Table 3.4C-3. The
distance to the nearest offsite individual is approximately one km. The
predominant exposure pathway is that of ground shine. These estimates
3.4C-2
-------
Table 3.4C-2. Radionuclide emissions from
the reference facility (Ri82)
Radio nuclide
Argon-4.1
Cobalt-60
Tritium
Carbon- 14
Krypton- 83m
Krypton-85m
Krypton-85
Krypton-87
Krypton-88
Xenon-131m
Xenon- 133m
Xenon-133
Xenon-135
Emissions
(Ci/y)
4.1E-1
l.OE-3
l.OE-3
l.OE-1
2.0E-2
2.4E-2
l.OE-3
5.0E-2
2.0E-2
5.0E-3
l.OE-2
2.1E-1
2.5E-1
are for an urban site with a regional population of 2.5E+6 (Reference
Site B). The maximum individual is located 500 meters from the
radiological support facility.
Table 3.4C-4 presents estimates of the maximum individual lifetime
risks and the number of fatal cancers to the regional population from
these doses.
Table 3.4C-3. Radiation dose rates from radionuclide
emissions from the reference facility
II •! I _„_ 11.11,1. III! V _._.._• _.«lll> im'-m PI l>* —«—.•— I • IP'— — II II I - - " "I -— • •••!• — - --
Maximum individual Regional population
Organ (mrem/y) (person-rem/y)
Total body 1.6E-2 8.9E-2
3.4C-3
-------
Table 3.4C-4. Fatal cancer risks due to radionuclide emissions from
the reference facility
Source
Lifetime risk
to maximum individual
Regional population
(Fatal cancers/y of operation)
Nuclear naval
shipyard
3E-7
3E-5
3.4C.7 Total Health Impact of U.S. Nuclear Naval Shipyards
The estimated total number of fatal cancers caused by all naval
shipyards is about 3E-4 per year. This estimate was derived from the
ratio of the capacity of the reference shipyard to the capacity of all
nuclear naval shipyards.
3.4C.8 Existing Emission Standards and Air Pollution Controls
There is no demonstrated treatment technology for controlling
emissions of krypton-85 or other radioactive noble gases from
radiological support facilities.
Tritium emissions could be controlled by using a catalytic
recombiner; however, this would be impractical considering the extremely
low levels of tritium emitted from radiological support facilities.
3.4C-4
-------
REFERENCES
EPA77 Environmental Protection Agency, Radiological Survey of Puget
Sound Naval Shipyard, Bremerton, Washington, and Environs, EPA-
520/5-77-001, Office of Radiation Programs, EPA, Washington,
D.C., 1977.
Ri82 Rice, P. D., Sjoblom G. L., Steele J. M. and Harvey B. F.,
Environmental Monitoring and Disposal of Radioactive Wastes
from U.S. Naval Nuclear-Powered Ships and Their Support
Facilities, Report NT-82-1, Naval Nuclear Propulsion Program,
Department of the Navy, Washington, D.C., 1982.
TRI79 Teknekron Research, Inc., Information Base (Including Sources
and Emission Rates) for the Evaluation and Control of
Radioactive Materials to Ambient Air, Interim Report, Volume 1,
EPA Contract No. 68-01-5142, July 1979.
3.4C-5
-------
3.5 Radiation Source Manufacturers
3.5.1 General Description
The term "radiation source" refers to radioactive material which is
enclosed in a sealed container or other nondispersible matrix. Radiation
sources are used in a wide variety of industrial and consumer products
including: (1) radioisotope gauges, which measure the thickness of
industrial products, (2) static eliminators, which are used to reduce
static electricity in industrial machines, (3) nondestructive testing
equipment, (4) self-illuminating signs and watch dials, and (5) smoke
detectors (EPA79).
3.5.2 Process Description v
Radiation source manufacturers process bulk quantities of radioactive
materials received from radionuclide production facilities such as
accelerators or reactors. During the manufacturing process, the
radioactive materials are handled with remote manipulators and custom-made
enclosures, such as glove boxes.
The manufacturers are licensed by NRC to have inventories of
radioactive materials in quantities ranging from ten Ci to as high as
100,000 Ci.
3.5.3 Emission Control Systems
Radiation source manufacturers use many different radionuclides in
their operations. In addition to conventional filtration systems for
removal of particulate matter, manufacturers may use other kinds of
treatment systems which are applicable to their particular emissions. For
example, tritium emissions can be reduced by use of desiccant type
scrubber columns which remove tritiated water; radioiodine releases can be
controlled with charcoal filters; facilities with emissions of krypton or
xenon can use chilled charcoal traps to delay the release of these gases
until radioactive decay has reduced their activity.
3.5.4 Radionuclide Emissions
Each radiation source manufacturer handles a unique combination of
radionuclides; therefore, each site has unique emission characteristics.
Table 3.5-1 shows radionuclide emission data on eighteen manufacturing
sites; these data were taken from reports submitted to NRC.
3.5.5 Reference Facility
For this analysis, a reference facility was created by summing all of
the radionuclides emitted by the eighteen sites listed in Table 3.5-1.
Other parameters used in the analysis were assumed to be those of an
industrial zone in a suburban area adjacent to a major city in the
midwestern United States. Table 3.5-2 describes the parameters of the
reference facility.
3.5-1
-------
3.5.6 Health Impact Assessment of Reference Facility
The estimated annual radiation doses from the reference facility for
individuals and population groups are shown in Table 3.5-3. Individual
fatal cancer risks and committed population fatal cancers are presented in
Table 3.5-4. The lifetime fatal cancer risk to the individuals at highest
risk is estimated to be 5E-6. The individual at highest risk is located
500 meters north of the source.
The estimated number of potential fatal cancers to the population
living in the region around the reference facility is estimated to be 2E-3
per year of facility operation.
Because of the way in which the reference facility was artifically
created, the maximum individual risk estimated for the reference facility
is much higher than the actual maximum individual risk associated with any
individual site. The population risk estimated for the reference facility
is equal to the total population risk for the eighteen sites listed in
Table 3.5-1.
3.5.7 Total Health Impact
The estimated number of fatal cancers caused by all radiation source
manufacturers is 2E-3 per year of operation, the same number as the
reference facility, because of the way in which the reference facility was
created.
3.5.8 Existing Emission Standards and Air Pollution Controls
Radiation source manufacturers licensed by NRC are subject to the
requirements of 10 CFR 20.106, which places limits on air emissions to
unrestricted areas. The particular controls used by a licensee to meet
these requirements will depend on the particular radionuclide(s) involved
and other factors unique to that licensee.
3.5-2
-------
Table 3.5-1,
Radionuclide emisssions from radiation source
manufacturers (Co83)
Site Radionuclide
A none
B Kr-85
C H-3
D Kr-85
E Th-232
F Kr-85
G H-3
Kr-85
H H-3
I none
J 1-125
Kr-85
Cs-137
K H-3
C-14
S-35
L H-3
M H-3
N H-3
0 H-3
P Kr-85
Q Kr-85
Xe-133
R Kr-85
Emissions
(Ci/y)
0.0
1.3
3E-1
5E-1
1.4E-1
IE -3
5.4E+1
5E+1
5E+1
0.0
2E-2
2.5
2E-3
2.14E+2
4.3
1.2E-1
2.5E-1
7.4E+2
3E-1
3E-2
2E-1
2E-3
2E-2
7.3
Table 3.4-2. Reference radiation source manufacturer
Parameter
Value
Fraction of radionuclides released:
Tritium
Krypton-85
Carbon-14
Stack height
1060
61.8
4.3
10 meters
3.5-3
-------
Table 3.5-3. Radiation dose rates from radionuclide emissions
from the reference radiation source manufacturer
0 Maximum individual Regional popultion
(mrem/y) (person-rem/y)
Weighted sum 0.22 8.4
Table 3.5-4. Fatal cancer risks due to radionuclide emissions
from the reference radiation source manufacturer
s Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Reference facility 5E-6 2E-3
3.5-4
-------
REFERENCES
Co83 Corblt C. D., et. al, Background Information on Sources of
Low-level Radionuclide Emissions to Air, PNL-4670, (Draft
report), Pacific Northwest Laboratory, Richland, Washington,
March 1983.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States,
Preliminary Report, EPA 520/7-79-006, August 1979.
3.5-5
-------
Chapter 4: COAL-FIRED UTILITY AND INDUSTRIAL BOILERS
4.0 Introduction
Large coal-fired boilers are used to generate electricity for public
and industrial use, as well as to provide process steam, process hot
water, and space heat. For the purposes of this report, boilers used in
the utility industry are designated utility boilers and those used
primarily to generate process steam/hot water are designated industrial
boilers.
From 1974 to 1977, about 18 percent of the energy needs in the
United States were met by burning coal; 60 percent to generate
electricity, and about 32 percent for industrial uses. More than
600 million tons are burned each year in utility and industrial boilers
(EPA 81).
Coal contains trace quantities of naturally-occurring
radionuclides. Table 4.0-1 lists typical concentrations of uranium and
thorium in U.S. coals. Uranium-238 and thorium-232 are the radionuclides
in coal with the longest half-lives. Other radionuclides in coal are
decay products of either uranium-238 or thorium-232 and are in secular
equilibrium with them. Tables 4.0-2 and 4.0-3 show the major decay
products of uranium-238 and thorium-232, respectively.
As coal is burned, the minerals in the coal melt and then condense
into a glass-like ash. A portion of ash settles to the bottom (bottom
ash) of the boiler and a portion enters the flue gas stream (fly ash).
Both the bottom ash and fly ash contain the radionuclides orginally
present in the coal, but more tends to combine with the fly ash. The
fraction of fly ash that is not trapped by emission control equipment and
that is released into the environment carries with it radionuclides; the
quantity released depends upon the radionuclide content of the coal,
furnace design, and efficiency of the particulate matter control system.
Radionuclides that are contained in fly ash exhausted to the
environment may expose people in several ways: they may be inhaled; they
may settle onto the ground and expose people nearby; and they may settle
onto crops or be taken up through the roots of crops and then be eaten.
Humans exposed to radiation by any of these means have an increased risk
of cancer and other health effects.
4.0-1
-------
Table 4.0-1. Typical uranium and thorium concentrations in coal
Uranium
Reg ion /Coal
Pennsylvania
Anthracite
Appalachian
Bituminous
NR
Bituminous
Bituminous
Illinois Basin
Bituminous
Bituminous
Northern Great Plains
Bituminous-
Subbituminous
S ubb i tumi nou s
Lignite
Western
NR
Rocky Mountain
Bituminous-
Subbituminous
Subbitumnious
Bituminous
Average for all coals
Range
(ppm)
0.3 -
0.2 -
0.4 -
0.1 -
0.3 -
0.2 -
0.2 -
0.2 -
0.1 -
0.2 -
0.3 -
0.2 -
0.1 -
0.1 -
25
11
39
19
5
43
59
3
16
13
2.5
24
76
42
Mean
(ppm)
1.2
1.0
1.3
1.1
1.2
1.3
1.4
1.7
0.7
1.0
1.2
1.0
1.0
1.9
1.4
1.3
Thorium
Range
(ppm)
2.8 -
2 -
1.8 -
-
—
0.7 -
3
0.0 -
2 -
0.1 -
0.3 -
0.6 -
3 -
0.1 -
0.2 -
14
48
94
5
79
79
8
42
14
6
35
54
18
Mean
(ppm)
4.7
2.8
2.0
3.1
1.9
1.6
3
2.4
3.2
2.3
2.0
4.4
3.0
3.2
Source: TRI81.
Note: 1 ppm uranium-238 is equivalent to 0.34 pCi/g of coal.
1 ppm thorium-232 is equivalent to 0.11 pCi/g of coal.
NR Not reported.
4.0-2
-------
Table 4.0-2. Major decay products of uranium-238
Radionuclide
Uranium-238
Thorium-234
Protactinium-234
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
Table
Radionuclide
Thorium-232
Radium-228
Actinium-228
Thorium-228
Radium-224
Radon-220
Polonium-216
Lead-212
Bismuth-212
Thallium-208
Polonium-212
Half-life
4.5xl09 y
24 d
1.2 m
2.4xl05 y
7.7x10* y
1.6xl03 y
3.8 d
3.1 m
27 m
20 m
1.6xlO~4 s
22 y
5.0 d
138 d
4.0-3. Major
Half-life
1.4xl010 y
5.8 y
6.1 h
1.9 y
3.7 d
55 s
0.15 s
10 h
60 m
3.1 m
3.1xlO~7 s
Principal radiation
Alpha
4.20
4.76
4.66
4.77
5.49
6.00
7.69
5.31
decay products
Beta
0.044
0.82
0.22
0.63
0.007
39
of thorium-232
Principal radiation
Alpha
4.00
5.40
5.67
6.29
6.78
8.78
Beta
0.010
0.376
0.100
0.717
0.561
(Mev)
Gamma
0.008
0.011
0.006
0.23
2.03
0.002
(Mev)
Gamma
0.915
0.002
0.009
0.117
0.281
2.37
years
days
hours
m = minutes
seconds
4.0-3
-------
4.1 Utility Boilers
4.1.1 General Description
At the end of 1979, the total capacity of U.S. electric utility
generating units amounted to 593 gigawatts (GW) (TRI81). Table 4.1-1
lists the capacity of the utility industry for 1979 and projections for
1985. Coal-fired steam electic power units accounted for 38 percent of
total capacity and 49 percent of total energy generation in 1979.
Coal-fired steam electric plants will account for 40 percent of total
generating capacity and for 49 percent of total power generation by
1985.
Power plants are designed and operated to serve three load
classes: (a) base-load plants, which operate near full capacity most
of the time (or are dispatched to operate in the most efficient region
of the heat rate curve); (b) intermediate-load (or cycling) plants,
which operate at varying levels of capacity each day (about 40 percent
utilization on an average annual basis); and (c) peaking plants, which
operate only a few hours per day (about 700-800 hours per year).
Fossil-fueled steam electric plants now dominate base-load and
intermediate-load service.
The average national capacity factor dropped from 55 percent in
1970 to 47 percent in 1978; the average base-load capacity factor, from
68 percent in 1970 to 64 percent in 1978. The average capacity factor
for cycling units remained almost constant over this period (TRI81).
Capacity and Age of Coal-Fired Steam Units
There were 1,224 coal-fired units with a total generating capacity
of 225 GW on line in 1979 (the base year). The distribution of these
units by capacity and age is shown in Table 4.1-2. About 50 percent of
coal-fired capacity is less than 10 years old. Most of the units with
capacities of 26 to 100 MW are between 25 and 29 years old, while those
with capacities of 101 to 300 MW are between 20 and 24 years old.
Units larger than 300 MW are 5 to 9 years old. About 21 percent of the
coal-fired units account for 50 percent of total generating capacity.
By 1985 there will be 1,360 coal-fired units on line with a
capacity of 307 GW, an increase over the base year of approximately 36
percent (.TRI81). In 1985, capacity of units less than 5 years old will
account for 22 percent of the total projected capacity and for about 10
percent of the total number of units.
The retirement rates for fossil units of a given capacity and size
will significantly affect system composition by 1985. Seventy-nine
coal units are scheduled for retirement by 1985. No retirements are
scheduled for units greater than 300 MW in capacity.
4.1-1
-------
Table 4.1-1. U.S. electric utility generating capacity
(Gigawatts)
1979
Generating technology
Coal-fired steam electric
Oil-fired steam electric
Gas-fired steam electric
Combined-cycle plants
Combustion gas-turbine,
internal combustion
Nuclear
Hydroelectric
Geo thermal
(GW)
225.1
101.4
59.9
2.5
76.9
51.1
73.3
.9
(% of
total)
(38.0)
(17.1)
(10.1)
(.4)
(13.0)
(8.6)
(12.4)
(.2)
(GW)
306.0
112-5
39.5
5.3
102.4
112.6
77.9
1.9
1985
(% of
total)
(40.0)
(14.7)
(5.2)
(.7)
(13.4)
(14.7)
(10.1)
(.2)
Others
2.0
(.3)
7.9
(1.0)
Total
593.1 (100.D*
766.0
(100.0)
Source: (iTR.181).
*Percentages do not add to 100.0 due to rounding.
Coal consumption by the electric utilities is expected to increase
from 438 million metric tons in 1979 to 633 million metric tons in 1985
(TRI81).
4.1.2 Process Description
In the typical power plant, a mixture of finely ground coal and
air is blown into a combustion chamber at the base of the boiler and
ignited as it passes through a burner. In the upper portion of the
boiler (above the combustion zone), boiler feedwater is simultaneously
pumped through a series of metal tube banks. The heat contained in
combustion gases is transferred to the feedwater which ultimately
leaves the boiler as saturated steam. This high-temperature,
high-pressure steam (540° C at 2.46 kgs/cm2) is used to drive a
turbine that, in turn, drives an electric generator-. Vapor leaving the
turbine is fed to a cooling system that extracts residual heat and
recycles condensate water back to the boiler.
Coal combustion produces an ash that is either retained within the
boiler (bottom ash) or carried out of the boiler with combustion
4.1-2
-------
Table 4.1-2.
Distribution of U.S. coal-fired units by age
and capacity, 1979
Capacity of coal-fired units
Age 1
0
0-4
5-9
10-14
15-19
20-24
25-29
30-34
35-39
40-44
45-49
50-54
55-59
60
Total
3.03-0.1 GW
Units) (GW)
6
9
19
26
36
104
60
32
3
2
2
0
0
299
0.5
0.5
1.3
1.6
2.3
7.1
3.4
1.8
0.1
0.1
0.1
0
0
18.8
0.1-0
.3 GW
(Units) (GW)
21
22
42
73
130
83
4
0
0
0
0
0
0
375
4.6
4.2
8.5
13.8
22.3
11.3
0.5
0
0
0
0
0
0
65.2
0.3-0
.6 GW
(Units) (GW)
54
44
40
18
4
0
0
0
0
0
0
0
0
160
24.8
20.2
18.0
7.1
1.3
0
0
0
0
0
0
0
0
71.4
Greater
than 0.6 GW
(Units) (GW)
30
42
12
2
0
0
0
0
0
0
0
0
0
86
22.4
33.8
8.7
1.3
0
0
0
0
0
0
0
0
0
66.2
Totals(a>
(Units) (GW)
120
139
132
140
204
255
121
58
24
7
16
6
2
1224
52.4
58.9
36.8
24.2
26.3
19.3
4.5
2.0
0.3
0.1
0.2
0.1
0.01
225.1
Source: (TRI81).
(a)Totals include an additional 304 units having a total capacity of
3.5 GW in the 0-0.03 GW range.
gases (fly ash). A portion of the fly ash is removed from the flue gas
before it is released to the atmosphere by a particulate control system.
Fly ash, bottom ash, slag, scrubber sludges, are removed from the
boiler and accumulate in solid waste piles adjacent to the plant.
These waste piles may 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.
Furnace Design
The distribution of particulates between bottom ash and fly ash
depends on the firing method, the ash fusion temperature of the coal,
and the type of boiler bottom (wet or dry).
4.1-3
-------
Fuel-firing equipment (Table 4.1-3) can be divided into three
general categories: stoker furnace (dry bottom), composed of spreader
or non-spreader types; cyclone furnace (wet bottom); and
pulverized-coal furnace (dry or wet bottom) (TRI81).
Table 4.1-3. Classification of coal-fired units by
firing method and type of boiler bottom, 1976
Stoker (all dry bottom)
Cyclone (all wet bottom)
Pulverized (wet bottom)
Pulverized (dry bottom)
Total
Number
of units
165
94
135
837
1231
Generating
capacity
f *k.MT 1 \
(MW)
2,015
24,449
16,440
161,092
203,996
Percent
of total
(1.0)
(12.0)
(8.0)
(79.0)
Source: (TRI81).
Note: Total number of units and generating capacity in Table 4.1-3 are
slightly different from previously-mentioned figures because of unit
retirements, derating, etc.
Stoker-Fired Furnacej. Stoker furnaces are usually small, old
boilers ranging in capacity from 7.3 to 73 MW (thermal). Of the
boilers designed for coal and sold from 1965 to 1973, none exceeded 143
MW(t); 63 percent were stoker-fired; 41 percent, spreader stoker; 9
percent, underfeed stoker; and 13 percent, overfeed stoker. Stokers
require about 3.3 kg of coal per kilowatt-hour and are less efficient
than units handling pulverized coal. Stoker-fired units produce
relatively coarse fly ash. Sixty-five percent of the total ash in
spreader stokers is fly ash.
Cyclone Furnaces. Crushed coal is burned in a high-temperature
combustion chamber called a cyclone. The high temperatures in the
furnace lead to the formation of a molten slag which drains
continuously into a quenching tank. Roughly 80 percent of the ash is
retained as bottom ash. Only 9 percent of the coal-fired utility
boiler capacity in 1974 was of the cyclone type, and no boilers of this
kind have been ordered by utilities in the past seven years (TRI81).
Pulverized-coal Furnaces. Coal is pulverized to a fine powder
(approximately 200 mesh) and injected into the combustion zone in an
intimate mixture with air. Pulverized-coal furnaces are designed to
remove bottom ash as either a solid (dry-bottom boiler), or as a molten
slag (wet-bottom boiler).
4.1-4
-------
The dry-bottom, pulverized-coal-fired boiler, in which the furnace
temperature is kept low enough to prevent the ash from becoming molten,
is now the most prevalent type of coal-burning unit in the utility
sector. About 80 to 85 percent of the ash produced in the dry-bottom,
pulverized-coal-fired boiler is fly ash. The remainder of the ash
falls to the bottom of the furnace, where it is either transported dry
or cooled with water and removed from the boiler as slurry to an
ash-settling pond.
Mode of Operation
The new units have historically been used for base load
generation; cycling capacity has been obtained by downgrading the
older, less efficient, base load equipment as more replacement capacity
comes on line.
In 1979, the average capacity factor^) for coal-fired units
operating in the base load mode was 65 percent; for units operating in
a cycling mode, 42 percent (TRI81). The availability^ of a
coal-fired unit generally declines with increasing generating
capacity. Generating units with capacities of less than 400 MW have
average availabilities of more than 85 percent; those with capacities
of more than 500 MW, only 74 to 76 percent (TRI81). The operating mode
affects the heat rate of the plant; for example, changing the capacity
factor from 42 to 70 percent changed the heat rate from 12.3 to 9.2
MJ/kWh.
4.1.3 Control Technology
Four types of conventional control devices are commonly used for
particulate control in utility boilers: electrostatic precipitators
(ESPs), mechanical collectors, wet scrubbers, and fabric filters.
Comprehensive evaluations of each control device have been given in
several publications (TRI81).
Selection of the particulate control device for a given unit is
affected by many parameters, including boiler capacity and type, inlet
loading, fly ash characteristics, inlet particle size distribution,
applicable regulations, and characteristics of the control device
itself. The location of particulate control devices with respect to
S02 scrubber systems in a plant depends on the type of scrubbers (wet
(^Capacity factor equals the ratio of energy actually produced in a
given period to the energy that would have been produced in the same
period had the unit been operated continuously at its rated power.
(^Availablity refers to the fraction of a year during which a unit
is capable of providing electricity to the utility grid at its rated
power after planned and forced outages have been accounted for.
4.1-5
-------
or dry) installed; these devices are located upstream of a wet
scrubber system or downstream of a spray dryer system.
ESPs with collection efficiencies of more than 99.8 percent have
historically been the control device of choice for utility boilers.
However, as a result of the growing use of low-sulfur western coals,
wet scrubbers and fabric filters have increasingly been chosen.
Table 4.1-4 shows the distribution of control equipment in use in
1976 on coal-fired steam electric boilers (TRI81).
4.1.4 Radionuclide Emissions
The emission of radionuclides in the fly ash generated during
combustion depends on the type of coal used; that is, its mineral
content and the concentrations of uranium, thorium, and their decay
products. Other factors influencing radionucide emissions include
furnace design, capacity, capacity factor, heat rate, ash partitioning,
enrichment factors, and emission control efficiency (Table 4.1-5). The
distribution of ash between the bottom and fly ash depends on the
firing method, coal, and furnace (dry bottom or wet bottom). For
pulverized-coal, dry bottom units, 80-85 percent of the ash is fly ash.
Recent measurements have shown that trace elements, such as
uranium, lead, and polonium, are partitioned unequally between bottom
ash and fly ash (Be78, Wa82). Although the concentration mechanism is
not fully understood, one explanation is that certain elements are
preferentially concentrated on the particle surfaces, resulting in
their depletion in the bottom ash and their enrichment in the fly ash
(Sm80). The highest concentration of the trace elements in fly ash is
found in particulates in the 0.5 to 10.0 micrometer diameter range, the
size range that can be inhaled and deposited in the lung. These fine
particles are less efficiently removed by particulate control devices
than larger particles. Based on measured data, typical enrichment
factors are: 2 for uranium, 1.5 for radium, 5 for lead and polonium,
and 1 for all other radionuclides (EPA81).
Coal storage and waste piles at utility boiler sites are also
potential sources of radon-222. Analyses of fugitive emission data
from these piles indicate, however, that the radon-222 "exhalation
rate" is less than that for soil, as reported by Beck (Be81).
Measured Radionuclide Emissions
EPA has measured radionuclide emissions at nine utility boilers.
Summaries of emissions data from these studies are presented in Tables
4.1-6 and 4.1-7.
4.1-6
-------
Table 4.1-4. Participate emission control equipment
by type of boiler, 1976
Control
equipment
No control
Mechanical (*)
Wet scrubbers
Fabric filters
ESP
Combination'*^
Stoker
Capacity
(GW)
0.7
0.8
-
0.1
0.4
Units
76
63
-
2
24
Pulverized
Capacity
(GW)
3.9
1.2
-
-
19.0
0.4
cyclone
Units
18
7
_
-
62
7
Pulverized
wet bottom
Capacity
(GW)
4.5
0.5
_
-
9.5
2.0
Units
66
11
_
-
44
14
Control
equipment
No control
Mechanical 'a'
Wet scrubbers
Fabric filters
ESP
Combination^
Dry bottom
Capacity
(GW)
26.8
2.4
1.9
0.8
110.1
19.2
Units
266
50
7
3
374
137
Total
Capacity
(GW)
35.9
4.9
1.9
0.9
138.5
22.0
Units
426
131
7
5
480
182
(^Mechanical devices include cyclones and gravitational chambers.
(b)Combination refers to mechanical-electrostatic precipitators.
Source: (TRI81).
4.1-7
-------
Table 4.1-5.
Parameters affecting radionuclide emissions
from coal-fired units
Parameter
Effect
Coal properties
(heating value, mineral
matter, moisture and sulfur
content)
Radionuclide content of ash depends
directly on the amounts of uranium,
thorium, and their daughters
contained in the coal, and the
percentage of mineral matter in
the coal.
Heat rate
Total particulate release is
directly related to coal
consumption, which in turn depends
on heat rate.
Capacity
Total particulate emission is
directly related to unit size.
Mode of operation
(capacity factor)
Mode of operation affects
capacity factor and heat rate,
which in turn influences total
particulate emissions.
Ash partitioning
Partitioning of ash between bottom
and fly ash directly affects
particulate emission rate.
Enrichment of radionuclides
in fly ash
The enrichment of certain
radionuclides in the fly ash
relative to the bottom ash directly
affects the radionuclide emission
rate.
Type of control device
Rate of particulate release
depends on the efficiency of
control devices.
4.1-8
-------
Estimated Radionuclide Emissions Based on Particulate Emissions
An estimate of the radioactivity released by coal-fired utility
boilers may be made by multiplying the particulate release rate (g/y)
by the average concentration of radionuclides in fly ash (pCi/g). The
concentration of radionuclides in fly ash depends on the concentrations
of uranium-238 and thorium-232 and their decay products in the coal,
the ash content of the coal, the partitioning between fly ash and
bottom ash, and the enrichment factors for the radionuclides in the fly
ash. Using typical values of 1.3 ppm uranium-238 and 3.2 ppm
thorium-232 in coal (see table 4.0-1), 7.6 percent ash in coal (TRI81),
a partitioning factor of 0.8 for fly ash (TRI81), enrichment factors of
2 for uranium and 1 for thorium, and the specific activities of uranium
and thorium, the calculated values for uranium and thorium in fly ash
are 9 pCi/g and 4 pCi/g, respectively. Values for the other
radionuclides in the uranium and thorium series may be derived by
applying enrichment factors of 1.5 for radium, 5 for lead and polonium,
and 1 for all other radionuclides.
Particulate emissions from coal-fired boilers have been listed by
EPA's National Emissions Data System (NEDS) (EPASOa). Data from this
system were used to select units releasing the largest amounts of
particulates into the atmosphere. The estimated uranium and thorium
emission rates (assuming 9 pCi of uranium-238 and 4 pCi of thorium-232
per gram of fly ash) for these units are listed in Table 4.1-8.
Table 4.1-6. Radionuclide emission rates (mCi/y) measured at
selected coal-fired steam electric generating stations
Sampling location^3'
Radionuclide
Uranium-238
Uranium— 23 4
Thorium-230
Radium-226
Lead-210
Polonium-210
Thorium-232
Thorium-228
M-l
24
24
1.5
5.3
28
68
0.81
0.72
M-2
5.7
7.2
4.1
4.1
15
14
1.5
1.7
M-3
0.76
0.81
0.29
0.21
1.4
1.1
0.02
0.30
M-4
0.10
0.10
0.08
0.02
0.18
0.16
0.05
0.05
Source: EPASOb
^'Sampling locations:
M-l West North Central Station (874 MW)-
M-2 East North Central Station (450 MW).
M-3 South Atlantic Station (125 MW).
M-4 Mountain Station (12.5 MW).
4.1-9
-------
The particulate emission rate for all the coal-fired utility
boilers in the United States has been estimated by NEDS to be 0.9
million metric tons per year. Assuming 9 pCi U/g fly ash, this results
in an annual emission rate of about 8 curies of uranium-238 for the
industry.
4.1.5 Reference Facility
The annual emissions of radionuclides from coal-fired utility
boilers cover a very large range because of the large range of coal
properties, boiler parameters, and control efficiencies. In order to
assess the health impact of these facilities, a reference coal-fired
utility boiler was defined. The uranium-238 emission rate for the
reference facility of 100 mCi/yr was selected on the basis of EPA's
judgment as representative of the upper range of potential emissions
(see Table 4.1-8J. Such emissions could be obtained from large
well-controlled boilers burning coal with relatively high uranium
content, or from large boilers with fewer controls burning coal of
average or less than average uranium content.
The source term for the reference facility is defined in
Table 4.1-9- The annual emissions for thorium, radium, lead, and
polonium reflect the fly ash enrichment factors noted previously. The
reference facility is assumed to have a stack height of 185 meters and
a plume rise of 50 meters, typical of large utility boilers.
4.1-10
-------
Table 4.1-7. Summary of radionuclide emission rates (mCi/y)
measured at five coal-fired steam electric generating units
Radionuclide
Uranium-238
Thorium-230
Radium-226
Polonium-210
Lead-210
Thorium-232
Sampling location^3)
M-l
100
12
45
1000
220
1.3
M-33
11
8.8
11
30
20
6.5
M-15
0.2
0.2
0.3
<0.6
(b)
0.04
M-34
2
0.5
4
3
3
0.5
M-99
0.03
0.04
0.1
<3.0
(b)
0.01
Source: (Re82).
^'Sampling locations and particulate control devices used at each of
the units are:
M-l West North Central unit (874 MW gross); wet limestone
scrubber.
M-33 South Central unit (593 MW gross); cold side ESP.
M-15 North Central unit (56 MW gross); mechanical collector
followed by a wet venturi scrubber.
M-34 South Central unit (800 MW gross); cold side ESP and
baghouse followed by a wet limestone scrubber.
M-99 North Central unit (75 MW gross); mechanical collector
followed by an ESP.
lead-210 analysis was made on samples collected at these units.
4.1.6 Health Impact Assessment of the Reference Utility Boiler
Because utility boilers may be located near large cites or at
remote sites, the health impact of the reference facility was
determined at four sites classified as urban, suburban, rural, and
remote (see Table 4.1-10).
4.1-11
-------
Table 4.1-8. Estimated uranium-238 emission rates from selected
utility boilers emitting the largest amounts of particulate matter
TT . Control Percent
Unit , ,-,-. .
device efficiency
A
B
C
D
E
F
G
None
Mechanical
Collector
Mechanical
Collector
Mechanical
Collector
ESP
ESP
ESP
0
50
78
90
75
85
91
Particulate
emission rate
(t/y)
23,900
9,300
12,300
6,600
21,100
23,900
18,100
Estimated
U-238
emission rate
(mCi/y)
210
80
100
60
190
210
160
Estimated
Th-232
emission rate
(raCi/y)
95
35
50
25
80
95
70
Source: National Emission Data System (EPASOa).
4.1-12
-------
Table 4.1-9. Radionuclide emissions from the reference facility
Radlonuclide
Emissions
(Cl/y)
Uranium series
Uranium-238
Uranlum-234
Thorium-230
Radium-226
Radon-222
Lead-210
Polonium-210
OE-1
,OE-1
,OE-2
,5E-2
9.6E-1
2.5E-1
2.5E-1
Thorium series
Thorium-232
Radium-228
Actinium-228
Thorium-228
Radium-224
Radon-220
Lead-212
Bismuth-212
Thallium-208
4.3E-2
6.5E-2
4.3E-2
4.3E-2
6.5E-2
8.3E-1
2 . 2E-1
4.3E-2
4.3E-2
Table 4.1-10. Population distribution of the reference facility
Site
Population(a)
Urban
Suburban
Rural
Remote
1.72E+7
2.49E+6
5.89E+5
1.19E+4
(*)Number of people located within a radius of 80 km.
The estimated annual radiation doses from radionuclide emissions
from the reference boiler are presented in Table 4.1-11.
4.1-13
-------
Table 4.1-12 presents estimates of the maximum individual lifetime
risks and the number of fatal cancers to the regional population
resulting from particulate doses at each of the generic sites for the
reference unit. The urban site is a conservative selection, and
estimates for this site represent an upper limit of the potential
health impact to a regional population.
The highest lifetime risk to the maximum individual is estimated
to be 4E-5 at the rural location due to external radiation from
radionuclides deposited on the ground surface. The highest number of
potential fatal cancers per year of operation is estimated to be 2E-1
at the urban location mainly from inhalation of radioactive fly ash
particles.
4.1.7 Health Impact Assessment of Specific Utility Boilers
EPA surveyed emissions from five utility boilers located in areas
similar to the generic rural site. The emission rates for these
boilers are listed in Table 4.1-7. Using the generic rural site data
and the actual emission rates measured by EPA, estimated annual
radiation doses were calculated (Tables 4.1-13 and 4.1-14).
Table 4.1-15 presents estimates of the maximum individual lifetime
risks and the number of fatal cancers to the regional population. The
risk values for the M-l unit are within a factor of three of the risk
values for the reference boiler at the rural site which has similar
emission rates.
Table 4.1-11.
Radiation dose rates from radionuclide emissions
from the reference utility boiler
Urban site
Organ
Lung
Red marrow
Kidney
Bone
Liver
Maximum
individual
(mrem/y)
9-9E-1
l.OE-1
l.OE-1
1.2
5.0E-2
Regional
population
(person-rem/y)
4.2E+3
3.4E+2
2.0E+2
4.8E+3
9.9E+1
Suburban site
Maximum
individual
(mrem/y)
1.3
2.1E-1
2.3E-1
1.8
1.4E-1
Regional
population
(person-rem/y)
3.9E+2
3.4E+1
1.9E+1
4.4E+2
1.1E+1
4.1-14
-------
Table 4.1-11. Radiation dose rates from radionuclide emissions
from the reference utility boiler—continued
Rural site
Remote site
Organ
Lung
Red marrow
Kidney
Bone
Liver
Maximum
individual
(mrem/y)
1.7
2.1
2.4
4.7
1.9
Regional
population
(person-rem/y)
1 . 1E+2
1.7E4-1
1.3E+1
1 . 3E+2
l.OE+1
Maximum
individual
(mrem/y)
1.2
1.4E-1
9.0E-2
1.4
6.0E-2
Regional
population
( person-rem/y)
1.1
l.OE-1
4.0E-2
1.2
3.0E-2
Table 4.1-12. Fatal cancer risks from the reference facility
Site
Lifetime risk
to maximum individual
Regional population
(Fatal cancers/y of operation)
Urban
Suburban
Rural
Remote
4E-6
6E-6
4E-5
5E-6
2E-1
2E-2
7E-3
5E-5
Table 4.1-13. Radiation dose rates to the maximum individual
from radionuclide emissions from five utility boilers
Maximum individual (mrem/y)
Organ
M-l
M-33
M-l 5
M-34
M-99
Lung
Red marrow
Kidney
Bone
Liver
5.4E-1
8 . 7E-1
3.8
2.0
1.2
4.3E-1
3.3E-1
3.3E-1
8.9E-1
2.7E-1
2.5E-2
7.0E-3
6.3E-3
3.7E-2
4.9E-3
5.5E-2
6.6E-2
5.8E-2
1.4E-1
5.4E-2
7.5E-3
3.2E-3
1 . 3E-2
1.4E-2
5.9E-3
4.1-15
-------
Table 4.1-14. Radiation dose rates to the regional population
from radionuclide emissions from five utility boilers
Regional population (person-rem/y)
urgan
Lung
Red marrow
Kidney
Bone
Liver
M-l
7.0E+1
5.2
2.2E+1
2.3E+1
6.6
M-33
2.1E+1
3.2
1.9
3.0E+1
1.4
M-15
7.1E-1
7.9E-2
4. IE -2
8.5E-1
2.9E-2
M-34
2.5
4.0E-1
3.0E-1
1.9
2.7
M-99
2.7E-1
2.6E-2
9.4E-2
2.0E-1
3.9E-2
Table 4.1-15. Fatal cancer risks from radionuclide emissions
from five utility boilers
Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
M-l
M-33
M-15
M-34
M-99
IE -5
6E-6
2E-7
IE -6
6E-8
4E-3
1E-3
4E-5
2E-4
2E-5
4.1.8 Total Health Impact of Utility Boilers
An estimate of the potential health impact of utility boilers
presently in operation may be made by assuming that the health effects
due to emissions from the reference boiler are proportional to the
health effects due to emissions from the whole industry.
About eight curies of uranium-238 per year are emitted by the
whole industry. Most of the U.S. generating capacity from coal-fired
utility boilers is located in areas that would be classified as either
suburban or rural. The estimates of health risks for the reference
facility at these locations are about 1E-1 to 2E-1 potential fatal
cancers per year per curie of uranium-238 released to the atmosphere.
Thus, the health impact from the industry is about eight times this or
about one to two potential fatal cancers per year.
4.1.9 Existing Emission Standards and Air Pollution Controls
There are no radionuclide emission standards for utility boilers.
Particulate emission rates are regulated by EPA and the States,
4.1-16
-------
however. EPA administers New Source Performance Standards (NSPS) that
apply to all utility boilers on which construction began after August
17, 1971, and before September 19, 1978, that have a firing capacity
greater than 73 MW(t) or 25 MW(e). Under these standards, particulate
emissions are limited to 43 ng/J. The 1979 revised New Source
Performance Standards (RNSPS), which apply to all 73 MW(t) or 25 MW(e)
electric utility steam generating units on which construction began
after 19 September 1978, require that particulate emissions be limited
to 13 ng/J (TRI81).
States regulate particulate emissions by State Implementation
Plans (SIPs). These must ensure that emission limitations and
reductions at new power plants are at least as stringent as those
stipulated in the NSPS and RNSPS. The SIPs must also include emission
limits for existing facilities (SIPs relate to National Ambient Air
Quality Standards—
NAAQS) under 40 CFR 50; EPA rules for SIPs are in 40 CFR 60, Subpart B).
All plants that were operating or under construction before
August 17, 1978, must be assigned emission limits by the SIP to ensure
attainment of air quality standards.
In most States, the SIP emission limits for pre-NSPS plants are
considerably less stringent than the NSPS limits. A survey of current
SIP limits shows that values of 43 and 86 ng/J are typical for the
stringent and less stringent states, respectively. SIP-regulated power
plants will continue to be the predominant source of electric utility
emissions thorugh the remainder of this century.
4.1.10 Supplemental Control Technology
Existing boilers can be retrofitted with additional electrostatic
precipitators (ESPs) to reduce emissions to the level prescribed for
new sources (13 ng/J); the number of fatal cancers is reduced also.
EPA's Office of Air Quality Planning and Standards has listed the
reduction in particulate emissions that would result from this action
(RC83). Table 4.1-16 shows how these reductons can be related to
population density.
The number of fatal cancers averted by reducing particulate
emissions can be calculated by converting the amount of particulate
emissions to curies of uranium-238 released and using the relationship
between curies of uranium-238 emitted per year and the number of
potential fatal cancers per year. The results are shown in Table 4.1-17.
Thus, by retrofitting all existing utility boilers to reach levels of
13 ng/J, about one potential fatal cancer may be averted per year.
4.1-17
-------
Table 4.1-16. Relationship of particulate emissions reduction to
population density
Population
density(a)
0-50,000
50,000-100,000
100,000-250,000
250,000-500,000
500,000-1 million
1 million-2.5 million
2.5 million-5 million
5 million-10 million
Generating
capacity
(MW)
8,070
7,040
7,140
43,820
82,840
72,700
31,080
15,430
Reduction in particulates
reach control level of 13
(104 tons/y)
2.5
2.1
2.2
13.3
25.3
22.2
9.5
4.7
to
ng/J
Total 268,100 81.8
(^Population within 80 km of a coal-fired utility boiler.
Source: (RC83).
Cost of Reduced Impact
EPA's Office of Air Quality Planning and Standards has estimated
the costs of retrofitting all existing coal-fired utility boilers with
control devices to reduce particulate emissions (RC83). To reach a
control level of 13 ng/J would result in a capital cost (1982 dollars)
of about $13 billion and an annual cost of about $3.4 billion.
4.1-18
-------
Table 4.1-17.
Number of fatal cancer averted by reducing
particulate emissions
Population
density
0-50,000
50,000-100,000
100,000-250,000
250,000-500,000
500,000-1 million
1 million-2.5 million
2.5 million-5 million
5 million-10 million
Reduction in
uranium-238(a)
emissions
(Ci/y)
0.2
0.2
0.2
1.2
2.3
2
0.9
0.4
Number of fatal
cancers averted
(per Ci/y)
3E-3
5E-3
2E-2
3E-2
6E-2
2E-1
4E-1
8E-1
Total number
of fatal can-
cers averted
5E-4
IE -3
4E-3
4E-2
1E-1
4E-1
4E-1
3E-1
Total
(a)These values are calculated by converting the reduction of
particulates released in tons/year to grams/year, multiplying by the
average concentration of uranium-238 in fly ash (9 pCi/g), and
converting to curies (1 Ci = 1012 pCi).
4.1-19
-------
REFERENCES
Be78 Beck, H.L., et al., 1978, Perturbations of the National
Radiation Environment Due to the Utilization of Coal as an
Energy Source, Paper presented at the DOE/UT Symposium on the
Natural Radiation Environment III, Houston, Texas,
April 23-28, 1978.
EPASOa Environmental Protection Agency, National Emissions Data
System Information, EPA 450/4-80-013, EPA, Office of Air
Quality Planning and Standards, Research Triangle Park, N. C.,
July 1980.
EPASOb Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into the Air in the United States,
EPA 520/7-79-006, EPA, Office of Radiation Programs,
Washington, D.C., 1980.
EPA81 Environmental Protection Agency, The Radiological Impact of
Coal-fired Industrial Boilers, EPA, Office of Radiation
Programs, Washington, B.C., (Draft Report), 1981.
RC83 Radian Corporation, Boiler Radionuclide Emissions Control:
The Feasibility and Costs of Controlling Coal-fired Boiler
Particulate Emissions, Prepared for the Environmental
Protection Agency, January 1983.
Re83 Memorandum from T. Reavey, EPA, Office of Radiation Programs,
to T. McLaughlin, EPA, Office of Radiation Programs,
Washington, D.C., February 1983.
Sm80 Smith, R.D., The Trace Element Chemistry of Coal during
Combustion and the Emissions from Coal-Fired Plants, Progress
in Energy and Combustion Science 6_, 53-119, 1980.
TRI81 Teknekron Research, Inc., Draft Background Information
Document, Coal-fired Electric Generating Stations, EPA
Contract 68-01-5152, April 1981.
Wa82 Wagner P. and Greiner N. R., Third Annual Report, Radioactive
Emissions from Coal Production and Utilization, October 1,
1980-September 30, 1981, LA-9359-PR, Los Alamos National
Laboratory, Los Alamos, N. M., 1982.
4.1-20
-------
4.2 Industrial Boilers
4.2.L General Description
Coal-fired industrial boilers (CFIBs) are used mainly to produce
process steam, generate electricity (for the producer's own use), and
provide space heat. The boilers are used in virtually every industry
from small manufacturing plants to large production concerns. The
major users are the steel, aluminum, chemical, and paper industries.
Of the coal consumed by industrial boilers in 1974, more than 87
percent was used by these four industries alone. A breakdown of the
percent of total coal consumed by each industry is given in Table 4.2-1.
Table 4.2-1. Industrial coal consumption, 1974
Industry ,Coal Consumption
3 (Percent of total)
Chemicals 33
Paper 26
Steel and aluminum 28
Food 10
Other manufacturing 3
Source: (TRI81).
4.2.2 Process Description
Types of Boilers
Three basic types of boilers are used in the industrial sector:
(1) water tube, (2) fire tube and, (3) cast iron.
Water tube boilers are designed so that water passes through the
inside of tubes that are heated externally by direct contact with hot
combustion gases. The process produces high pressure, high temperature
steam with a thermal efficiency of about 80 percent. Water tube
boilers range in capacity from less than 3 MW to over 200 MW thermal
input.
Fire tube boilers are designed to allow the hot combustion gas to
flow through the tubes. Water to be heated is circulated outside the
tubes. The boilers are usually smaller than 9 MW thermal input.
Cast iron boilers are designed like fire tube boilers with heat
transfer from hot gas inside the tubes to circulating water outside the
tubes, but cast iron is used rather than steel. Cast iron boilers are
generally designed for capacities less than 3 MW.
4.2-1
-------
Number and Capacity of Boilers
Table 4.2-2 lists the number of boilers and their total installed
capacity (EPA81). Water tube units represent 89 percent of the total
installed capacity of all boilers in terms of the thermal input. Since
the capacity (amount of coal burned) influences the level of emissions
to the environment, the radiological impact of coal-fired industrial
boilers will be that associated with emissions from water tube type
units. Cast iron and fire tube units will not be considered further in
this report.
Table 4.2-2. Number and capacity of coal-fired industrial boilers
Unit capacity (MW thermal input)
Boiler type _ __ __ __ _____
Water Tube Units 683 2309 1290 1181 423
Total MW 835 22225 27895 50825 59930
Fire Tube Units 8112 1224
Total MW 5650 7780
Cast Iron Units 35965
Total MW 6330
Coal-Firing Mechanisms
There are two main types of coal-fired water tube boilers:
pulverized coal and stoker-fired. Pulverized coal units burn coal
while it is suspended in air. Units range in sizes from 30 MW to over
200 MW heat input. A stoker unit has a conveying system that serves to
feed the coal into the furnace and to provide a grate upon which the
coal is burned. Stokers are generally rated at less than 120 MW heat
input. The three main types of stoker furnaces are spreader, overfeed
(or chain grate), and underfeed. Each of the boiler types is discussed
below.
Pulverized coal-fired boilers
Coal is pulverized to a light powder and pneumatically injected
through burners into the furnace. If the furnace is designed to
operate at a high temperature (typically 1600° C), the ash remains in
a molten state until it collects in a hopper at the bottom of the
furnace. The high temperature units are known as "wet bottom" units.
4.2-2
-------
Dry bottom units operate at lower combustion temperatures with the
bottom ash remaining in the solid state. Combustion temperatures
initially reach about 1200-1600° C.
Spreader stoker
Coal is suspended and burned as a thin, fast-burning layer on a
grate, which may be stationary or moving. Feeder units are used to
spread the coal over the grate area, and air is supplied over and under
the grate to promote good combustion.
Overfeed stokers
Coal is fed down from a hopper onto a moving grate that enters the
furnace. Combustion is finished by the time the coal reaches the far
end of the furnace, and ash is discharged to a pit.
Underfeed stokers
Coal may be fed horizontally or by gravity, and the ash may be
discharged from the ends or sides. Usually the coal is fed
intermittently to the fuel bed with a ram, the coal moving in what is
in effect a retort, and air is supplied through openings in the side
grates.
Particulate Emissions by Boiler Type
The fractional distribution of ash between the bottom ash and fly
ash directly affects the particulate emissions rate and is a function
of the following parameters:
Boiler firing method. The type of firing is the most important
factor in determining ash distribution. Stoker-fired units emit
less fly ash then pulverized coal-fired boilers.
Wet or dry bottom furnaces. Dry bottom units produce more fly ash.
Boiler load. Particulate emissions are directly proportional to
the amount (load) of coal burned.
4.2.3 Control Technology
Radionuclides are removed from flue gas with the particulates.
The following paragraphs discuss technologies commonly used to remove
particulates.
Electrostatic Precipitators
Particle collection in an electrostatic precipitator (ESP) occurs
in three steps: (1) suspended particles are given an electric charge,
4.2-3
-------
(2) the charged particles migrate to a collecting electrode of opposite
polarity where they are collected, and (3) the collected particulates
are dislodged from collecting electrodes. Energy is needed to operate
the precipitator in amounts equivalent to 0.02 to 0.1 percent of the
fuel energy input to the boiler. ESP efficiency varies with a number
of factors, of which particle size is most significant. Table 4.2-3
shows typical efficiences.
Table 4.2-3. ESP collection efficiency as a function of particle size
Particle diameter Average collection efficiency
(micrometer) (Percent)
0-5 72
5-10 94.5
10-20 97
20-44 99.5
Greater than 44 100
Fabric Filter
In fabric filtration, particle-laden flue gas is passed through
the fabric to trap particles; the cleaned gas passes through the fabric
into the atmosphere.
Energy is required to operate equipment, such as fans, cleaning
equipment, and the ash conveying system. The energy requirement
depends on the type of boiler and its capacity; it ranges from 3 to 8
times as great as the energy required for an ESP.
The overall mass collection efficiency of a fabric filter ranges
from 99 to 99.9 percent with an average of roughly 99.7 percent.
Fabric filter control efficiency is not affected by changes in coal
sulfur and alkali content, variables which can signicantly affect ESP
performance. The efficiency of the fabric filter is also not sensitive
to the inlet particle size distribution.
Wet Scrubber
Scrubbers operate on the principle of capturing particulates by
bringing them into contact with liquid droplets or wet scrubber walls.
They require significant amounts of energy to operate fans and liquid
pumps. The energy requirements, which range from 0.2 to 0.7 percent of
4.2-4
-------
the fuel energy input to the boiler, depend on the type of boiler and
its capacity, characteristics of coal consumed, and level of
particulate matter control.
The control efficiency of wet scrubbers is a function of system
pressure drop and inlet particle size distribution. Typical collection
efficiencies, as a function of pressure drop are shown in the Table
4.2-4.
Table 4.2-4. Typical wet scrubber efficiency
Pressure drop Overall collection efficiency
(KPa)' (percent)
1.24
2.5
5.0
7.5
88-95
92-97
95-98
96-99
Mechanical Collectors
The typical mechanical collector is the cyclone collector. The
cyclone collector transforms the velocity of an inlet gas stream into a
confined vortex from which centrifugal forces tend to drive the
suspended particles to the wall of the cyclone body.
The energy requirements are roughly 1 to 2 1/2 times greater than
that of ESPs or about 0.12 percent of the fuel energy input to the
boiler.
The level of efficiency of the mechanical collector (cyclone) is
much lower than ESPs, fabric filters, or wet scrubbers. Additionally,
the mechanical collector becomes less efficient as particle size
decreases. Accordingly, they are not used to remove small particules.
4.2.4 Radionuclide Emissions
Radionuclide emission rates from coal-fired industrial boilers
have not been measured. However, by knowing the radionuclide con-
centrations in either fly ash or coal, radionuclide emissions from
boilers can be estimated.
Table 4.2-5 lists the estimated emission rates of uranium-238 for
several industrial boilers in the paper, steel/aluminum, and chemical
4.2-5
-------
industries. These boilers were selected, on the basis of particulate
emissions data obtained from EPA's National Emissions Data System, to
represent the highest controlled and uncontrolled sources. Uranium-238
emission rates were calculated by using an average value of 9 pCi of
uranium-238 per gram of fly ash (EPA81).
Table 4.2-5. Estimated radionuclide emissions for boilers with the
highest particulate emission levels
„ . . . , Uranium-238 emissions (Ci/y)
Emission control —*
Paper Steel/Aluminum Chemical
None^a)
Mechanical collection
Electrostatic precipitator
3E-2
1E-2
1E-2
6E-2
3E-2
—
5E-2
2E-2
1E-2
Assumes that 50 percent of the particulate matter contains
particles too large to enter the lungs.
Table 4.2-6 lists other estimates of uranium-238 emission rates
from representative coal-fired industrial boilers. The estimates are
based on uranium-238 concentrations in the coal used to fire the
boilers (TRI81).
Table 4.2-6. Estimated uranium-238 emission rates
for representative coal-fired industrial boilers
Boiler capacity
(MWJ
9
22
44
59
118
Emission
control
yes
no
yes
no
yes
no
yes
no
yes
yes
Particulate
matter
(ng/J)
194
782
172
712
138
1850
129
2420
86
43
Uranium-238 emissions
(Ci/y)
1E-4
4E-4
3E-4
1E-3
4E-4
6E-3
7E-4
7E-3
9E-4
4E-4
4.2-6
-------
We estimate the uranium-238 emission rate for the entire
population of large (15 MW and greater) coal-fired industrial boilers
subject to SIP particulate matter limits to be 3 Ci/y.
4.2.5 Reference Coal-Fired Boiler
We chose the source term of the reference case (see table 4.2-7)
industrial boiler to resemble the amount of radionuclides that could be
released from a large industrial boiler to air under normal
operations. Our source term assumptions were conservative so that our
projected radiological impacts should be greater than most, but
possibly not all, new and existing industrial boilers. There could be
different combinations of plant size, coal radionuclide content, levels
of control technology, etc., that would yield a source term
approximately equal to the one we selected for the reference case.
The source term was calculated using the same methodology used for
utility boilers (see Section 4.1) and reflects the relatively smaller
thermal capacity and coal consumption of industrial boilers. Table
4.2-8 lists other characteristics of the reference boiler used in the
health impact assessment.
Table 4.2-7. Radionuclide emissions from the reference boiler
Radionuc1ide
Emissions
(Ci/y)
Uranium series:
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-222
Lead-210
Po-212
Thorium series:
Thorium-232
Radium-228
Actinium-228
Thorium-228
Radium-224
Radon-220
Lead-212
Bismuth-212
Thallium-208
l.OE-2
l.OE-2
5.0E-3
7.5E-3
2.5E-1
2.5E-2
2.5E-2
4.3E-3
6.5E-3
4.3E-3
4.3E-3
6.5E-3
8.3E-2
2.2E-2
4.3E-3
4.3E-3
4.2-7
-------
4.2.6 Health Impact Assessment of Reference Industrial Boiler
The estimated annual radiation doses from the reference industrial
boiler are listed in Table 4.2-9. Table 4.2-10 presents estimates of
the maximum lifetime risk and the number of fatal cancers to the
regional population from these doses.
4.2.7 Total Health Impact of Coal-Fired Industrial Boilers
The estimated total number of fatal cancers caused by all coal-
fired industrial boilers is about one per year* This estimate was
derived by multiplying the health effects for the reference boiler by
the ratio of the total (estimated) uranium-238 emissions of the entire
CFIB industry and the reference boiler.
Table 4.2-8. Reference coal-fired industrial boiler
Parameter Value
Site Midwest location (St. Louis),
Population 2.5 million people within
80 km of the site
Stack
Effective height 150 meters
Diameter 1.5 meters
Table 4.2-9. Radiation dose rates from radionuclide emissions
from the reference industrial boiler
Q Maximum individual Regional population
(mrem/y) (person-rem/y)
Lung
Red marrow
Kidney
Bone
Liver
3.4E-1
4.0E-2
4.0E-2
4.3E-1
2.0E-2
7.6E+1
6.6
9.0
9.0E+1
3.2
4.2-8
-------
Table 4.2-10. Fatal cancer risks due to radionuclide emissions
from the reference industrial boiler
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Industrial boiler IE-6 4E-3
4.2.8 Existing Emission Standards and Air Pollution Control
No Federal or state regulations currently exist that limit
emissions of radionuclides from coal-fired industrial boilers.
However, the states, through State Implementation Plans (SIPs), and the
Federal government, through New Source Performance Standards (NSPS),
regulate particulate matter emissions and thus effectively limit
radionuclide emissions.
All existing coal-fired industrial boilers are subject to SIPs.
Since the individual SIPs reflect local conditions and needs,
particulate matter emissions vary from state to state.
All new coal-fired industrial boilers with capacities greater than
73.3 MW (thermal input) are subject to a particulate emission limit of
43.3 ng/J (40 CFR 60, subpart D.) New boilers with capacities less
than 73 MW are subject to limits prescribed by the SIPs.
4.2.9 Supplemental Control Technology
Currently, large coal-fired industrial boilers (15 MW and
greater), which are subject to SIP particulate matter limits, emit
about 0.37 million tons of particulate matter per year. Table 4.2-11
lists the costs, particulate matter emission levels, and
cost-effectiveness to retrofit large boilers to meet specific uniform
emission levels (RC82).
Table 4.2-12 lists estimated uranium-238 emissions for existing
and retrofitted large boilers (15 MW and larger) subject to SIP
particulate matter control.
Table 4.2-13 lists estimated current risks and risk reductions for
particulate matter limits for large (15 MW and greater) coal-fired
industrial boilers.
4.2-9
-------
Table 4.2-11. Estimated costs and particulate matter reductions
from retrofit controls for coal-fired boilers'3'
Emission level
Costs
(0.1 lbs/106BTU) (0.05 lbs/106BTU)
Capital Cost $2.5 billion $3.4 billion
Annual Cost $550 million $730 million
Particulate matter
reduction 0.15 million 0.19 million
tons/y tons/y
Cost effectiveness:
($/ton) $3,600 $3,800
(a)15 MW and greater boilers.
Table 4.2-12. Estimated uranium-238 emission rates for existing
and retrofitted large coal-fired industrial boilers'3'
Particulate matter
control level rate Uranium-238 emission rate
(lbs/106 BTU) (Ci/y)
Various under SIPs 2.9
0.1 1.7
0.05 1.4
^a^15 MW and greater.
4.2-10
-------
Table 4.2-13. Risks associated with large coal-fired
industrial boilers(a)
Particulate matter Risks
control level (Fatal cancers/y)
Various under SIP's 1
0.1 7E-1
0.05 6E-1
(a)l5 MW and greater.
4.2-11
-------
REFERENCES
EPA81 Environmental Protection Agency, The Radiological Impact of
Coal-fired Industrial Boilers (Draft), EPA Office of Radiation
Programs, Washington, D.C., October 1981.
TRI81 Teknekron Research, Inc., Draft Background Information
Document for Coal-Fired Industrial Boilers, May 1981.
RC82 Radian Corporation, Development of Coal Combustion
Radionuclide Particulate Emission and Control Cost
Information, Draft Report, November 1982.
4.2-12
-------
Chapter 5: URANIUM MINES
5.1 General Description
In uranium mining operations, ore is removed from the ground in
concentrations of 0.1 to 0.2 percent U^ or 280 to 560 microcuries
of uranium-238 per metric ton of ore. Since the uranium-238 in the ore
is normally present in secular equilibrium with its daughter products,
these ores also contain equal amounts of each member of the uranium-238
decay series.
After mining, the ores are shipped to a uranium mill to separate
the uranium. Radioactive emissions to air from uranium mines and mills
consist of uranium bearing dust and radon-222 gas.
Uranium is mined in both open pit and underground mines. In 1981
there were 167 underground and 50 open pit uranium mines in operation
in the United States (Table 5-1). These mines accounted for about 79
percent of the uranium produced (DOE82).
Table 5-1. Distribution of 1981 ^Og production in ore
by mining method (DOE82)
Source
Underground mines
Open pit mines
Solution Mining
(In-Situ)
Others:
heap-leach,
mine water,
byproduct, and
low-grade stockpiles
Total
Number
167
50
14
23
254
Tons U30g(a)
8,500
7,000
2,100
2,000
19,600
Percent
of total
43
36
11
10
100
(a)Short tons
5-1
-------
In recent years in-situ solution mining has been more widely used;
this method is expected to increase in future years. During 1981 this
method accounted for 11 percent of the uranium mined in the United
States. The radioactive emissions from this source are small compared
to the other sources.
Table 5-2 indicates that at present all uranium is mined in the
western United States, mostly in the states of New Mexico, Wyoming, and
Texas. Exploration for uranium is being conducted, however, in the
eastern and midwestern parts of the United States.
Table 5-2 Distribution of 1981 ^Og production in
ore by State (DOES2)
State
New Mexico
Wyoming
Texas
U3°8
(Short tons)
6,600
4,400
3,200
Percent of
total
34
22
16
Arizona, Colorado,
Texas, Utah, & Washington 5,400 28
Total 19,600 100
Major publicly-held corporations account for a large share of
ownership in the uranium industry. The industry grew rapidly in the
early and mid-1970's, stimulated by expectations of rapid increases in
demand. However, the expectations were too optimistic, with supply
outstripping demand. The result was an economic slump for the
industry. The industry is now faced with excess capacity, large
inventories, lower-than-expected demand, and the potential for
increased competition from imports.
5.2 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
5-2
-------
established underneath the ore body. Vertical raises are 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 drifts to the vertical raises
and gravity fed to the haulage way for transport to the main shaft for
hoisting to the surface.
Figure 5-1 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 nH/min. The principal radioactive effluent in the mine
ventilation air is radon-222 which is released during mining
operations. Additional radon-222 and particulate (uranium and thorium)
emissions result from surface operations at the underground mine.
Surface Mining
Open pit mining usually is carried out by excavating a series of
pits in sequence. The topsoil and overburden are removed from above
the ore zone and stockpiled 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 5-2 is an example of an open pit mining operation. As the
mining progresses, mining and reclamation operations take place
simultaneously—pits 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 radon-222 gas and fugitive dust
containing uranium and thorium.
In-Situ Mining
In this method, a leaching solution is injected through wells into
the uranium-bearing ore body to dissolve in the uranium. Production
wells bring the uranium-bearing solution to the surface where the
uranium is extracted. The solution (lixiviant) can be recovered and
reused.
Radon-222 gas is emitted from the processing operations and waste
piles. With solution mining, less than 5 percent of the radium from an
ore body is bought to the surface (EPA82). Consequently, the amount of
radon released is considerably less than that from conventional
mining. The major sources of radon are the surge ponds, enclosed surge
tanks, inplant surge tanks and absorption columns (Br81). It is
5-3
-------
GENERALIZED UNDERGROUND URANIUM MINE
MODIFIED DOOM AND PILIAR METHOD OF MINING
Figure 5.1-1. An underground uranium mining operation.
-------
Ul
Figure 5.1-2. An open pit uranium mining operation.
-------
estimated that the radon released is about 19 percent of the amount
released from a conventional uranium mill (EPA82).
A small amount of radon is released from the waste piles formed as
a result of the operation. Some examples of solid wastes that might be
generated by the alkaline leach in-situ process are:
(1) Materials filtered from the lixiviant line,
(2) Sediments from the surge tanks,
(3) Calcium carbonate from the calcium control unit,
(4) Barium sulfate from the contaminated control in the
elution/precipitation circuit of the recovery process,
(5) Materials deposited in the evaporation ponds,
(6) Drill hole residues,
(7) Solids from aqiufer restoration.
EPA has previously evaluated radionuclide emissions from uranium
mining activities (EPA79, EPA83). These evaluations indicate that
underground uranium mining releases the largest quantities of radon-222
to air and results in the most significant health impacts when compared
to other mining methods. Because of the lower amounts of radon
released from surface mines, in-situ solution mining, and other mining
methods, the potential health impact of underground uranium mining is
of the most concern and therefore, Sections 5.3 through 5.8 of this
chapter deal only with underground uranium mines.
5.3 Control Technology
Several methods to control radon emissions from underground
uranium mines have been evaluated. These are: 1) use of sealant
coating on exposed ore surfaces; 2) bulkheading of worked-out areas; 3)
activated carbon adsorption of radon from contaminated mine air; 4)
mine pressurization; and 5) miscellaneous technologies.
Sealant Coating
One of the the best methods for controlling radon in an
underground uranium mine is to prevent radon from entering the mine air
by sealing exposed surfaces. A summary of field tests and a review of
the literature on this subject performed for EPA (Ko80) is summarized
as follows:
1) Under laboratory conditions sealants are very effective in
attenuating radon emissions from ore surfaces, but in an actual
mine application, the presence of "pinholes" and the difficulty of
applying a perfect coating on an ore surface reduces the
effectiveness of these sealants considerably.
2) In field tests a three-coat system of HydroEpoxy 156 and
HydroEpoxy 300, preceded by Shotcrete base coating, was found to
5-6
-------
be effective (50 to 75 percent radon stoppage). For the
theoretical mine, the sealant probably would be 60 percent
effective with an eight-month lifetime.
3) The amount of sealants used varied considerably for different
mines. Kown and his associates (Ko80) chose the following amounts
for their study which were greater than other studies on this
subject.
Shotcrete - 909 gal per 1000 ft2
HydroEpoxy 156 - 18 gal per 1000 ft2
HydroEpoxy 300 - 32 gal per 1000 ft2
4) The sealant coating applied to drifts of an underground mine
have a limited life of about eight months because the drift area
is mined after pillars are extracted in a room-and-pillar stope
mine.
5) An asphalt emulsion sealant has been tested in the laboratory
and on tailing piles and is found to be an effective, inexpensive
sealant. However, it has not yet been tested in an underground
mine atmosphere.
The cost of coating 530,000 ft2 of drift surfaces in the mine
was $344,300 ($1.45 per ton of ore removed). The floors were not
considered to be coated because ore loaders will destroy the coating on
the semiconsolidated muck. The three sealants were applied every two
months. Cost estimates of other sealants range from $0.30 to $1.10 per
square foot which is comparable to the sealants used in this study
(Fr81).
This EPA study has shown that sealants could reduce the radon
emanation from the active stopes of the mines by 23 percent. If the
total mine is included (25 extracted stopes), only 11 percent of the
radon was reduced. This second figure should be used when determining
the amount of radon released from the mine. Other studies by the
Bureau of Mines (Fr81) have shown that 50 to 75 percent of the radon
can be retained in the rock by sealants.
Bulkheading
Bulkheading of mined-out areas, such as extracted stopes, is the
most common radon control method currently practiced in underground
mines (Ko80). In general, it is used to isolate worked-out areas or
stopes from workers so that the radon concentrations in the working
areas of the mine will be lower. If the bulkhead is air tight, the
radon behind the barrier will decay to innocuous levels. However, all
bulkheads leak to some extent, and usually a small 3- to 6-inch pipe is
used as a bleeder pipe to provide negative pressure in the extracted
stope (Fr81) and to allow the contaminated air to be diverted to the
5-7
-------
ventilation system. A small fan may be required to maintain the
negative pressure. Ideally, only 10 percent of the air behind the
bulkhead would be diverted to the outside atmosphere. This air stream
can also be connected to an activated carbon filter or trap to reduce
concentrations further.
In an EPA study (Ko80) it was assumed that 12.5 stopes per year
would be sealed using 100 bulkheads. The cost for material, labor, and
maintenance was estimated to be $80,400 or $0.34 per ton of ore
removed. It was also assumed that a six-inch pipe provided a 100 cfm
bleeding rate from each bulkheaded area.
An estimate of the effectiveness of reducing radon by this system
was made using many crude assumptions. For the total mine, bulkheading
was estimated to achieve about a 14 percent reduction in radon
emissions. A small preliminary study conducted by Battelle on an
actual mine indicated that a radon reduction of 35 percent could be
obtained by using bulkheads (DraSO).
Radon Adsorption on Activated Carbon
Leakage of high radon concentrations through bulkheads used to
control radon concentrations in mines is another problem. One method
to relieve this, problem is to insert a small bleeder pipe in the
bulkhead to provide negative pressure within the enclosed area behind
the bulkhead. This bleeder pipe is usually connected to the exhaust
ventilation system. Although this may prevent exposure to the workers,
the radon emissions to the environment may still be high. An activated
carbon adsorption system may be attached to the radon effluent pipe
before releasing this air to the exhaust ventilation system (Ko80).
An effective radon cleanup system for the bleeder pipes is still
under study. The system chosen by investigators in an EPA study (Ko80)
is shown in Figure 5-3. It consists of two carbon adsorption systems
in series. The flow from the bleeder pipe is filtered to remove dust
particles and radon daughter products. The radon is then adsorbed in
the carbon column. The carbon column is regenerated once a day, using
hot air. The contaminated air from the regeneration is sent through a
second carbon column to again adsorb the radon gas. Occasional drying
may be required in the second column due to buildup of moisture.
In evaluating control technology in a model mine, EPA (Ko80) found
that an average of 12.5 activated carbon systems must be installed each
year to treat the contaminated air from the stopes sealed by the
bulkheads. The capital and operating costs for each unit is as follows:
Capital Cost of Each Unit
Major equipment $22,000
Auxiliaries & Installation $11,000
$33,000
5-8
-------
tn
FILTER BLOWER
PRIMARY CARBON BED |
I HSOOIbs) |
100 - 300JCFA
BLOWER
0.5HP)
HEATER
(50KW)
SECONDARY |
CARBON BED I
(150lbsl
1 BLOWER
(2HP>
CONDENSED
WATER TO
DRUMS
I
SPARE
PRIMARY
ADSORPTION
CARBON
REGENERATION &
SECONDARY
ADSORPTION
Figure 5.1-3. Radon removal from mine air by carbon adsorption.
-------
Annual Cost of Each Unit
Material (carbon, filters, piping) $ 1,000
Utilities ($25,000 kwh @ 4c/kwh) $ 1,000
Labor (0.25 man-year) $ 8,000
Amortizing (an avg. 5-year life at 10% interest) $ 8.700
Total $18,700
Assuming the lifetime of each unit is 5 years and 12.5 units per
year are needed, the annual cost over five years would be $233,750 or
$4.32 per ton of ore mined. The carbon system was assumed to be 95
percent efficient in removing radon.
The effectiveness of the entire system, including bulkheading and
carbon traps, was estimated to be 49 percent. A study by Battelle
(DraSO) estimates a 45-68 percent effectiveness, using absolute
sorption traps in combination with bulkheading. The total cost for
bulkheading and carbon traps would be $4.66 per ton of ore mined.
There are some definite disadvantages to the carbon adsorption
system. Skilled operators, usually not available in mining
communities, are necessary to operate and maintain the system. Safety
problems to the miners are possible due to interrupted electrical
service or system malfunction. Excess radon concentrations would then
be present. The carbon columns would have to be shielded to prevent
gamma exposure to the miners.
The system does appear to be technically feasible utilizing
commercial carbons and standard equipment. However, additional
developmental work is necessary before such a system can be used in a
mine environemt.
Mine Pressurization
Positive mine pressurization has been tried several times to force
the radon in the mine atmosphere back into the walls of the mine (Ko80
and Fr8l). iIn general, these efforts have been successful in reducing
the radon concentrations in the mine itself. An "air" sink is
necessary to accept the radon. If the radon is forced through the ore
body or surrounding area to the surface, the radon can decay before
coming to the surface. If the area is impermeable however, radon
levels will return to previous levels. In the latest tests (Fr81), the
radon levels in the mine were reduced by 20 percent; releases to the
atmosphere were not determined. The surrounding area needs to be
permeable enough to reduce gas flo.w to the surface and also to allow
decay of the radon (Ko80). The costs of mine pressurization are not
available because the process is in a development stage.
Miscellaneous Radon Control Technology
Argonne National Laboratory is experimenting with strong oxidizing
agents, such as bromine triflouride and dioxygenyl hexaflouro-antimonate,
5-10
-------
to convert the radon to another form that can be absorbed on a scrubber
or absorption bed (Fr81). However, the corrosive and toxic nature of
the reactants makes their use in mines impracticable and questionable.
Backfilling of worked-out areas is practiced by mine operators to
get rid of excess tailings from the mine and/or mill (Fr8l). This
procedure is used mainly for ground support and to reduce ventilation
requirements. A study, by the Bureau of Mines and Kerr-McGee Nuclear,
to determine the effectiveness of reducing radon emissions by
backfilling mill tailings into the mine stopes indicated a radon
reduction of 57 percent in exhaust air (Fr81). However, radioactive
and nonradioactive contamination of mine water was increased due to
seepage from the tailing slurry.
Increasing the height of vents is a possible method to reduce
ground level radon concentrations in ambient air (DraSO). One of the
conclusions based on a theoretical model was that "a 20-meter release
height reduces the annual average concentration (when compared to a
ground-level release) by about 60 percent at one mile from a source and
by about 30 percent at ten miles from the source." A preliminary
estimate of cost is about $35,000 for a 20-meter stack (B183). The
average number of vents for a mine is about 6 (Ja80). Thus, the cost
per mine would be about $210,000 or $.40 per ton of ore produced.
Summary of Costs and Efficiencies
A summary of the costs and efficiencies of the various radon
control technologies is shown in Table 5-3.
Table 5-3. Cost and efficiencies of radon control technologies
for underground uranium mines
Radon reduction Cost
Method (Percent) ($/ton of ore)
Sealant coating 11 1.45
Bulkheading 14 0.34
Activated carbon 35 4.32
Bulkheading with activated
carbon 49 4.66
Mine Pressurization 20-25
Stacks 60(a) 0.40
Backfilling 57
(a)Reduction in exposure to maximum individual.
5-11
-------
Table 5-4. Summary of measurements of radon-222 emissions from
underground uranium mine vents (Ja80)
Mine
A
B
C
D
E
F
G
H
I
J
K
L
R
T
U
V
Y
Z
AA
BB
CC
DD
EE
FF
GG
HH
II
Number
of
vents
4
6
6
2
14
15
5
10
11
12
4
9
11
5
3
2
7
3
3
8
3
2
5
3
3
2
2
Average 6
Ore
produc-
tion ., V
(tons/dra;
2190
712
946
1070
1000
715
794
480
300
368
352
250
114
420
500
550
2630
500
-
—
_
-
-
-
150
-
"—
740
Ore
grade
%u3o8
0.19
0.24
0.21
0.20
0.16
0.19
0.18
0.10
0.12
0.19
0.47
0.06
0.18
0.20
0.15
0.11
0.15
0.14
-
—
-
-
-
-
0.16
-
—
0.18
Years in
produc-
tion
3
9
9
7
21
20
4
21
-
20
19
29
20
-
4
2
6
17
-
—
—
-
-
-
—
-
—
13
Radon-222
emissions,. N
(Ci/y) (b)
7,400
4,500
4,600
3,630
29,800
9,400
1,800
15,200
1,690
7,900
6,400
1,400
14,800
1,890
890
1,010
17,500
2,640
1,800
2,000
2,120
960
6,500
2,510
170
1,040
470
5,600
(a'Based on 1978 survey of mines.
(b)Average of data from measurements made in 1978 and 1979.
5.4 Radionuclide Emission Measurements
Radon-222 is the radionuclide emitted from underground uranium
mines which causes the greatest risk to people. The major source of
5-12
-------
radon-222 emissions to air are the mine vents through which the
ventilation air is exhausted. A large underground mine will usually
have several vents; some mines have as many as 15 vents. Radon-222
emissions from these vents are highly variable and depend upon many
interrelated factors including: ventilation rate, ore grade,
production rate, age of mine, size of active working areas, mining
practices, and several other variables.
Pacific Northwest Laboratories (PNL) has measured the radon-222
emissions from 27 underground uranium mines (Ja80) (Table 5-4). The
average radon-222 emission rate for these 27 mines was 5,600 curies/
year.
In addition to the mine vents, radon-222 is emitted to air from
several above-ground sources at an underground uranium mining
operation. These sources are the ore, subore, and waste rock storage
piles. PNL has estimated the radon-222 emissions from these sources to
be about 2 to 3 percent of the emissions from the vents (Ja80). EPA
has estimated the emissions from the above-ground sources to be about
10 percent of mine vent emissions (Table 5-5).
The above-ground sources also emit radionuclides to air as
particulates. The particulate emissions result from ore dumping and
loading operations and wind erosion of storage piles. EPA has
estimated that about 2E-2 Ci/y of uranium-238 and 3E-4 Ci/y of
thorium-232 and each of their decay products would be emitted into the
air at a large underground mine (EPA83). An assessment of the health
Table 5-5. Estimated annual radon-222 emissions from underground
uranium mining sources (EPA83)
o Average large mine(a)
Underground
Mine vent air 3,400
Aboveground
Ore loading and dumping 15
Sub-ore loading and dumping 5
Waste rock loading and dumping 0
Reloading ore from stockpile 15
Ore stockpile exhalation 53
Sub-ore pile exhalation 338
Waste rock pile exhalation 3
Total 3829
(a)0re grade = 0.1% 11303. Annual production of ore and sub-ore
2 x 105 MT, and waste rock = 2.2 x 104 MT.
5-13
-------
risks from these emissions showed that the risks from the particulate
emissions were much smaller (a factor of 100 less) than the risks from
radon-222 emissions (EPA83). Therefore, the health risk assessment
presented in the subsequent sections of this chapter will be limited to
radon-222 emissions.
5.5 Reference Underground Uranium Mine
Table 5-6 describes the parameters of the reference mine which are
used to estimate the radon-222 emissions to the atmosphere and the
resulting health impacts. These parameters were chosen primarily from
information in Table 5-4. The reference mine has 5 vents in the
configuration as shown in Figure 5-4.
The radon-222 emissions to air from the reference underground mine are
listed in Table 5-7. These emissions are based on information in Table
5-4 and on the aboveground percent emissions from Table 5-5. In the
reference mine, each of the mine vents is assumed to emit the same
quantity of radon-222.
Table 5-6. Reference underground uranium mine
Parameter Value
Ore grade 0.18% U308
Ore production 1.7E+5 MT/y
Days of operation 250 days/y
Number of vents 5
Vent height 3 meters
Table 5-7. Radionuclide emissions from the reference
underground uranium mine
Source Emissions (Ci/y)
Radon-222
Mine vents(a) 5,000
Ore, subore
and waste rock piles 500
Total 5,500
(a)l,000 Ci/y from each vent.
5-14
-------
Figure 5-4. Reference underground mine.
5-15
-------
5.6 Health Impact Assessment of the Reference Underground Uranium Mine
The lifetime risk to nearby individuals and the number of fatal
cancers per year of operation due to radon-222 emissions from the
reference underground uranium mine are presented in this section.
The risks to individuals are treated in more detail than the
population risks because the individual risks can be relatively high
for people living near the mine. These exposures generally occur in
structures built around the mines. Radon-222 enters the building and
decays into other radionuclides which become attached to dust particles
in the air. The concentration of these radionuclides build up in the
air within the structures.
EPA estimated the health risks from radon-222 emissions from
uranium mines using the general assumption discussed in Appendix B. It
is important to recognize that the actual risk to individuals may
differ from these estimates because the circumstances involving the
exposure may differ significantly from the assumptions used to make the
estimates. For example, because mines have a limited lifetime (10 to
20 years), the period of exposure is likely to be less in real cases
than assumed in the models. Furthermore, people need to be occupying a
structure and not just standing outdoors for these estimates to be
applicable.
Individual Risks
In assessing the health risks to individuals living near
underground uranium mines, an evaluation was made of (a) the radon-222
emissions from a single vent, (b) a series of vents from a single mine,
(i.e., the reference mine), and (c) multiple vents from multiple mines
which are close together.
Estimates of the radon-222 concentration at various distances from
an underground uranium mine vent emitting 1000 Ci/y of radon-222 are
shown in Table 5-8. Also shown in this table are the estimated
individual lifetime risks of fatal cancer from the inhalation of
radon-222 decay products produced (inside a house) by these radon-222
concentrations. Similarly, Table 5-9 shows the estimated radon-222
concentrations in air and the individual lifetime risks of fatal cancer
at various distances from the reference underground uranium mine (i.e.,
5 mine vents distributed as shown in Figure 5-4 and each emitting 1000
Ci/y of radon-222).
The radon-222 concentration in air at any specific location near a
uranium mine with multiple vents is highly dependent upon the spatial
distribution of the vents with respect to the location of interest and
the wind frequency distribution. The data in Table 5-9 illustrate the
levels which could occur in a given situation. For other situations
(i.e., different spatial distributions of the vents, wind frequencies,
etc.)> tne radon-222 concentrations in air could be higher or lower
than the values shown in Table 5-9.
5-16
-------
Table 5-8. Estimates of radon-222 concentrations in air and
individual lifetime risks at selected distances from an
underground uranium mine vent(a)
Distance
(meters)
100
200
500
Maximunr '
Radon-222
(pCi/L)
8.9
5.4
1.3
Individual
lifetime risk
1E-1
6E-2
2E-2
( c)
Average
Radon-222
(pCi/L)
2.7
1.5
0.37
Individual
lifetime risk
3E-2
2E-2
4E-3
1,000
2,000
3,000
5,000
10,000
0.38
0.12
0.06
0.03
0.01
5E-3
IE-3
7E-4
4E-4
IE-4
0.11
0.03
0.017
0.009
0.003
IE-3
4E-4
2E-4
IE-4
4E-5
(a)The lifetime risks were estimated using the relationship that 1
pCi/L of radon-222 in air results in a radon-222 decay product
concentration of 0.007 working levels inside a house.
(^/Predominant wind direction.
(c)Average of all wind directions.
Table 5-9. Estimates of radon-222 concentrations in air and
individual lifetime risks at selected distances from
the reference underground uranium mine(a)
Maximum
Distance
(meters)
(c)
Radon-222Vb; individual
(pCi/L) lifetime risk
Average
Radon-222^13^ Individual
(pCi/L) lifetime risk
500
1,000
2,000
3,000
5,000
10,000
1.5
0.54
0.24
0.18
0.12
0.06
2E-2
6E-3
3E-3
2E-3
1E-3
7E-4
0.60
0.27
0.11
0.07
0.04
0.015
7E-3
3E-3
1E-3
8E-4
5E-4
2E-4
(a)The lifetime risks were estimated using the relationship that 1
pCi/L of radon-222 in air results in a radon-222 decay product
concentration of 0.007 working levels inside a house.
(k)The data used to estimate these concentrations and examples of the
method of calculation are shown in Appendix C. The concentrations
listed under the heading "maximum" are believed to be the highest
concentrations which would occur at these distances for the
configuration of mine vents postulated and the meteorology used.
(c)The distance from mine vent 5 in southeasterly direction which is
the predominant wind direction (See Figure 5-4).
5-17
-------
To evaluate the extent to which emissions from multiple mines
located close together will influence the radon-222 concentrations in
air, PNL carried out a modeling study using the Ambrosia Lake District
of New Mexico as a "case study" (DrbSl). Using a Gaussian diffusion
model, estimates were made of the radon-222 concentrations in air
resulting from emissions from 117 mine vents. Figure 5-5 shows the
distribution of mine vents used in the study and Figure 5-6 the
computed radon-222 concentration (above background) in air for this
region. Although these computed concentrations are only approximate
values due to the many complexities of this type of modeling study, the
results of this study indicate that the radon-222 concentrations in an
intensive underground uranium mining area will be significantly
elevated above background. The vents are also the greatest sources of
the radon concentrations in the immediate area of mining and milling
activities. The study shows that, in areas surrounded by mine vents,
the radon-222 concentrations may range up to 12pCi/L above background.
Population Risks
The radon decay product exposures and the number of fatal cancers
per year of operation for the reference underground uranium mine are
shown in Table 5-10. These estimates are for a site near Grants, New
Mexico, with a regional population of 36,000. The number of fatal
cancers per year of operation of the reference mine is estimated to be
0.03 to the regional population and 0.07 to the national population.
Table 5-10. Annual radon-222 decay product exposures and number of
fatal cancers to the population from radon-222
emissions from the reference underground uranium mine
Regional population National population
Source (Person- (Fatal cancers/y (Person- (Fatal cancers/y
WL-y) of operation) WL-y) of operation)
Underground
uranium mine 1.1 3E-2 3.1 7E-2
5.7 Health Impact from Underground Uranium Mining
An estimate of the total health impact from radon-222 emissions
from all underground uranium mining (using production values for 1981)
may be made by multiplying the number of fatal cancers caused by
emissions from the reference mine by the ratio of the amount of uranium
produced by all underground mines to the amount produced by the
reference mine. This estimate for the regional population was 0.7
fatal cancers/year and for the national population was about 2 fatal
cancers/year.
5-18
-------
15km
10km
Skin
5km
Figure 5-5. Detailed map of mining area showing source.
5-19
-------
. j
Figure 5-6. Computed radon concentration map for region isopleths
are in pCi/L.
5-20
-------
Rather than control radon emissions at the source, it may be more
practical to limit the exposure to individuals near underground mines
by controlling land near the vents to prevent people from living in
houses in these areas. At the request of EPA, Battelle Northwest
Laboratory conducted a field study in January and February 1983 to
determine the population, type of ownership, and cost of land around 30
large uranium mines. These mines represented about 90 percent of the
uranium production from underground mines at that time.
Table 5-11 shows the population data gathered from the Battelle
study. An estimate was made of all residents within 5 km of the mine
shaft by locating all the residences on a map. The average 1980 census
figure of residents per home in each county was used to estimate the
population. If mines were close together, populations were evenly
distributed among the mines according to the distances from the mines.
Table 5-12 represents; the percent distribution of land ownership
around the 30 surveyed mines. County tax assessor's records were
reviewed for all properties within a 5-km radius of each mine. The
ownership of the land was determined and percentages, according to
three types of ownership (private, mine, or government), are shown for
each mine.
Table 5-13 summarizes the cost of the land around each mine.
Since the land owned by the mine operator or a government agency can
already be controlled, only costs to purchase private land were
determined.
The Schwartzwalder mine near Denver, Colorado, is not included in
the total cost of all surveyed mines shown in Table 5-13 because it is
not a typical mine site. It is located near a large metropolitan area
and the cost of the land is quite high since the land can be purchased
or subdivided for mountain resort homes. The mine is also isolated in
a mountainous region so that radon emissions would be confined in the
immediate area of the mine and any land control which may be necessary
would be relatively small.
The information in Tables 5-11 through 5-13 can be used to obtain
a rough estimate of the cost to control land around underground uranium
mines. The cost to control land within a 2-km radius of the mines
surveyed is as follows:
5-21
-------
Total cost Yearly cost
Type of cost (millions) (millions)
Land cost (100%
contingency with 10% yearly cost) $12.5 $1.3
Structures (100% contingency with
30% amortization) 3.0 .9
Relocation of residents ($5,000/
person with 30% amortization) 3.3 1.0
Purchase of Indian dwellings ($20,000/
person—220 Indians, with 30%
amortization) 4.4 1.3
Total yearly cost 4.5
The 10 percent yearly cost assumes that the land value does not
change and thus is a nondepreciated asset. The 30 percent amortization
figure assumes that the mine will operate for five years with an
interest rate of 10 percent. A small amortization percentage is added
to the interest rate for taxes.
Assuming that the 29 mines produced 90 percent of the underground
mine yearly production of 8,600 tons of V^Og for the industry
(DOE82), the cost of land control per pound of l^Og can be
estimated as follows:
cost/lbU_0, - _J__ = $0.29/lb U308
J ° (.9)(8,600)(2,000)
If production costs for U308 are $30/lb, the increased cost to
the industry would be 1 percent of the cost of production.
5-22
-------
Table 5-11. Population around selected underground
uranium mines (B183)
Mine
Sunday
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kabab North
Derma -Snyder
Wilson-
Silverbell
Lisbon
LaSal
Hecla
Big Eagle
Golden Eagle
Sheep Mtn.
Mt. Taylor
Old Church
Rock
Church
Rock-NE
Church
Rock-1
Church
Rock-East
Kerr-McGee
Sec 30 East
Kerr-McGee
Sec 30 West
Kerr-McGee
Sec 19
Kerr-McGee
Sec 35
Kerr-McGee
Sec 36
State
Colo.
Colo.
Utah
Utah
Arizona
Arizona
Arizona
Colo /Utah
Utah/ Colo.
Utah
Utah
Utah
Wyoming
Wyoming
Wyoming
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
Distance from
0-1/2
0
0
0
0
1
0
0
0
0
0
0
16
0
0
0
0
0
12
12
0
3
0
0
0
0
0-1
0
0
0
0
1
0
0
8
0
0
0
16
0
0
0
100
0
31
30
0
3
0
0
0
0
0-2
0
0
0
0
1
0
0
33
0
0
53
20
0
0
0
317
65
51
51
19
8
0
0
0
0
mine
0-3
0
0
0
0
1
0
0
53
12
4
101
40
0
9
0
336
93
53
54
19
8
0
0
0
0
(km)
0-4
0
0
0
0
1
0
0
73
20
44
194
73
0
12
0
336
192
53
54
30
12
0
0
0
0
0-5
0
0
0
0
1
0
0
83
23
44
194
73
3
12
12
336
469
93
93
78
13
0
0
0
0
5-23
-------
Table 5-11. Population around selected underground
uranium mines (B183) (Continued)
Mi no •? f-a *-o .--IT -
0-1/2
Home stake
Sec 23 New Mexico 0
Home stake
Sec 25 New Mexico 0
Nose
Rock(a) New Mexico 2
Mariano
Lake(a) New Mexico 3
Schwartz-
walder'") Colorado 3
Totals 49
Distance from mine (km)
0-1 0-2 0-3 0-4 0-5
0033 4
0000 0
3 14 31 55 86
5 21 47 83 130
3 63 102 136 147
197 653 855 1235 1963
^'Estimates for intermediate distances apportioned on area. These
were on Indian land. Roads were impassable. Estimates out to 5 km
were obtained from the Office of Environmental Health, USPHS Indian
Hospital, Crown Point, New Mexico.
population around this mine is not included in the total
because the location is not typical of the industry.
5-24
-------
Table 5-12. Percent distribution of land ownership around
selected underground "uranium mines (B183)
Mine
Sunday
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kanab North
Dermo-Snyder
Wilson-
Sil verb ell
Lisbon
LaSal
Hecla
Big Eagle
Golden Eagle
Sheep Mtn.
Mt. Taylor
Old Church
Rock
Church Rock
ME
Church Rock
#1
Church Rock
East
Kerr^lcGee
Sec 30 East
Kerr-McGee
Sec 30 West
Kerr-McGee
Sec 19
Kerr-McGee
Sec 35
Kerr-McGee
Sec 36
Distance from mine (km)
0-1/2U)
0/0/100
0/0/100
14/0/86
0/0/100
0/0/100
0/0/100
0/0/100
84/0/16
80/0/20
0/0/100
8/0/92
25/0/75
0/100/0
60/20/20
30/45/25
75/19/6
0/0/100
0/0/100
0/0/100
0/0/100
11/89/0
11/89/0
0/100/0
0/100/0
5/42/53
0-1
0/0/100
0/2/98
10/0/90
0/0/100
0/0/100
0/0/100
0/0/100
87/0/13
95/0/5
0/0/100
25/0/75
25/0/75
0/88/12
89/7/4
18/42/40
58/26/16
0/0/100
0/7/93
0/7/93
0/7/93
4/91/5
24/76/0
23/77/0
0/85/15
14/22/64
0-2
0/0/100
0/5/95
6/0/94
0/0/100
0/0/100
0/0/100
0/0/100
84/0/16
95/0/5
6/0/94
34/0/66
48/0/52
0/80/20
85/3/2
5/28/69
58/25/17
0/0/100
0/23/77
0/23/77
0/6/94
2/70/28
17/72/11
46/39/15
8/59/33
27/14/59
0-3
3/1/97
0/3/97
12/0/88
0/0/100
0/0/100
0/0/100
0/0/100
89/0/11
94/0/6
17/2/81
41/0/59
37/0/63
0/8/92
94/1/5
2/18/80
47/17/36
0/0/100
0/13/87
0/13/87
3/4/93
4/78/18
16/69/15
45/39/16
14/55/31
36/8/56
0-4
8/1/91
0/3/97
24/0/76
0/0/100
0/0/100
0/0/100
0/0/100
85/0/15
91/0/9
21/1/78
34/0/66
28/0/72
0/5/95
91/1/8
4/11/85
39/13/48
2/0/98
0/8/92
0/8/92
5/2/93
10/79/11
22/66/12
32/37/31
10/57/33
36/5/59
0-5
10/1/89
0/3/97
27/0/73
0/0/100
0/0/100
0/0/100
1/0/99
81/0/19
81/0/19
16/1/83
26/0/74
21/0/79
1/3/96
90/1/9
12/8/80
40/8/52
3/1/96
0/5/95
0/5/95
3/1/96
13/77/10
27/57/16
29/38/33
14/52/34
39/3/58
See footnotes at end of table.
5-25
-------
Mine
Table 5-12. Percent distribution of land ownership around
selected underground uranium mines (B183) (Continued)
Distance from mine (km)
0-1/2
(a)
0-1
0-2
0-3
0-4
0-5
Home stake
Sec 23
Home stake
Sec 25
Nose Rock
74/0/26
100/0/0
0/50/50
68/0/32
85/0/15
0/50/50
61/6/33
59/0/41
0/45/55
50/18/32
58/1/41
0/41/59
47/17/36
50/2/48
0/38/62
53/12/35
43/10/47
0/35/65
Mariano Lake
Schwartz
Average
0/0/100 0/0/100 0/0/100 0/0/100 0/0/100 0/0/100
100/0/0 100/0/0 100/0/0 100/0/0 100/0/0 100/0/0
20/22/58 22/20/58 22/17/61 23/13/64 22/12/66 22/11/67
first figure in the column represents the percent of private
land, the second is land owned by the mine owner, and the third shows
the percentage of land owned by a government agency. For example, in
the case of the Sunday mine (at 0-1/2 km), 100 percent is owned by the
government.
land ownership percentage for the Schwartzwalder mine was not
included in the average for all the mines since the location is not
typical of the industry.
5-26
-------
Table 5-13. Estimated value of private land around
selected underground uranium mines(a) (B183)
(In thousands)
Mi TIP
Distance from mine (km)
0-1/2
Sunday
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kanab North
Dermo-Snyder
Wilson-
Silverbell
Lisbon
LaSal
Hecla
Big Eagle
Golden Eagle
Sheep Mtn.
Mt. Taylor
Old Church Rock
Church Rock NE
Church Rock-1
Church Rock-East
Kerr-McGee
Sec 30 East
Kerr-McGee
Sec 30 West
Kerr-McGee
Sec 19
Kerr-McGee
Sec 35
Kerr-McGee
Sec 36
NA
NA
5.5
NA
NA
NA
NA
79.7
39.1
NA
4.0
36.8
NA
35.4
18.0
39.6
NA
NA
NA
NA
35.0
31.1
NA
NA
3.4
0-1
NA
NA
16.0
NA
NA
NA
NA
260.4
186.4
NA
228.4
147.3
NA
209.0
42.3
225.1
NA
NA
NA
NA
35.0
132.2
194.4
NA
23.5
0-2
NA
NA
36.0
NA
NA
NA
NA
922.6
535.8
50.0
920.9
380.0
NA
796.2
42.3
890.7
NA
NA
NA
NA
35.0
147.8
844.8
37.0
124.3
0-3
48.0
NA
172.8
NA
NA
NA
NA
1,852.1
1,667.2
306.0
1,427.8
691.0
NA
2,121.0
42.3
1,211.8
NA
NA
NA
122.2
53.5
157.9
1,229.4
137.8
336.0
0-4
208.0
NA
603.2
NA
NA
NA
NA
3,028.9
2,861.6
810.5
2,484.5
965.9
NA
3,584.0
150.0
1,604.9
543.3
NA
NA
355.6
147.6
194.8
1,405.1
168.0
588.0
0-5
384.0
NA
1,048.0
NA
NA
NA
(b)
4,432.8
3,968.7
810.5
2,534.5
1,000.5
NA
5,435.0
898.0
2,296.3
1,443.1
NA
NA
355.6
240.0
235.1
1,532.8
336.0
977.8
See footnotes at end of table.
5-27
-------
Table 5-13. Estimated value of private land around
selected underground uranium minesCa) (B183) (Continued)
(In thousands)
Mine
Home stake
Sec 23
Home stake
Sec 25
Norse Rock
Mariano Lake
Schwartz
walder^c)
Totals
0-1/2
217.8
295.6
NA
NA
880.0
841.0
0-1
528.0
622.2
NA
NA
3,400.0
2,850.2
Distance
0-2
994.1
987.8
NA
NA
15,200.0
7,745.2
from mine
0-3
1,158.7
1,478.0
NA
NA
33,600.0
14,213.4
(km)
0-4
1,485.2
1,632.2
NA
NA
58,400.0
22,821.3
0-5
2,361.8
1,645.6
NA
NA
89,200.0
31,937.1
^'Includes cost of land and structures.
100 acres of patented mining claims.
costs for this mine were not included in the total costs
because the location and cost of land is not typical of the industry.
NA Not assessed; all land owned by either the mine owner or the
government.
5-28
-------
REFERENCES
B183 Written communication between C. Bloomster of
Battelle-Pacific Northwest Laboratory and J. Silhanek, EPA,
January and February 1983.
Br81 Brown S. H and Smith R. C., A Model for Determining the
Overall Radon Release Rate and Annual Source Term for a
Commerical In-Situ Leach Uranium Facility, Proceedings of
International Conference on Radiation Hazards in Mining:
Control Measurement, and Medical Aspects, Colorado School of
Mines, Golden, Colorado, October 1981.
DOE82 Department of Energy, Statistical Data of the Uranium
Industry, GJO-100(82), Grand Junction, Colorado, January 1982.
DraSO Droppo, J. G., et al., An Environmental Study of Active and
Inactive Uranium Mines and Their Effluents, Part I, Task 3,
Pacific Northwest Laboratory, PNL-3069, Part I, August 1980.
DrbSl Droppo, J. G. and Glissmeyer, J. A., An Assessment of the
Radon Concentrations in Air Caused by Emissions from Multiple
Sources in a Uranium Mining and Milling Region. A Case Study
of the Ambrosia Lake Region of New Mexico, Pacific Northwest
Laboratory, PNL-4033, December 1981.
EPA79 Environmental Protection Agency, Radionuclide Impact Caused by
Emissions of Radionuclides into Air in the United States, EPA
520/7-79-006, EPA, Office of Radiation Programs, Washington,
D.C., August 1979.
EPA82 Environmental Protection Agency, Draft Environmental Impact
Statement for Remedial Action Standards for Active Uranium
Processing Sites, EPA, Office of Radiation Programs,
Washington, D.C., 1982.
EPA83 Environmental Protection Agency, Potential Health and
Environmental Hazards of Uranium Mines Wastes (Draft), Office
of Radiation Programs, Washington, D.C., March 1983.
Fr81 Franklin J. C., Control of Radiation Hazards in Underground
Mines, Bureau of Mines, Proceedings of International
Conference on Radiation Hazards in Mining: Control
Measurement, and Medical Aspects, Colorado School of Mines,
Golden, Colorado, October 1981.
5-29
-------
REFERENCES—continued
Ja80 Jackson P. 0., et al., An Investigation of Radon-222 Emissions
from Underground Uranium Mines—Progress Report 2, Pacific
Northwest Laboratory, Richland, Washington, February 1980.
Ko80 Kown B. T., et al., Technical Assessment of Radon-222 Control
Technology for Underground Uranium Mines, EPA Contact No.
68-02-2616, ORP/TAD-80-7, EPA, Office of Radiation Programs,
Washington, D.C., April 1980.
5-30
-------
Chapter 6: PHOSPHATE INDUSTRY FACILITIES
6.1 Phosphate Rock Processing Plants
6.1.1 General Description
Phosphate rock is the starting material for the production of all
phosphate products. Mining of phosphate rock is the fifth largest
mining industry in the United States in terms of quantity of material
mined (Da68). Phosphate rock mines of significant commercial
importance are located in Florida, North Carolina, Tennessee, Idaho,
Wyoming, Utah, and Montana (Figure 6.1-1).
The U.S. production of phosphate rock was estimated to be 57.9
million metric tons in 1978 with production increasing an average of
about 5 percent per year (EPA79). The industry consists of 20 firms
which are currently mining phosphate rock at 31 locations. Another
five mines are expected to be operational by 1983, and four others have
been planned with indefinite start-up dates. Most firms have mining
operations and rock processing plants at the same location, while a few
companies mine in several areas and ship the rock to a central
processing plant. Table 6.1-1 shows the phosphate rock producing
companies, plant locations, 1977 production, and percent of U.S. market.
The southeastern U.S. is the center of the domestic phosphate rock
industry, with Florida, North Carolina, and Tennessee having over 90
percent of the domestic rock capacity. Florida, with approximately 78
percent of 1978 domestic capacity, dominates the U.S. industry and is
the world's largest phosphate rock producing area. Most of these
plants are located around Polk and Hillsborough counties in Central
Florida, with expansion taking place in Hardee and Manatee counties.
Hamilton County, located in North Florida, is another phosphate rock
producing area.
Tennessee's phosphate rock industry, located in the middle of the
State, has declined in importance over the last several years and is
now the least important rock producing area in the country. The
Tennessee Valley Authority and two private corporations have
discontinued mining in Tennessee, and no new plant expansion is planned.
6.1-1
-------
Figure 6.1-1. Geographical location of phosphate rock operations.
6.1-2
-------
Table 6.1-1. Phosphate rock producers and capabilities (EPA79)
1977 production _, . f
- • Percent of
^ \
Company and location total
International Minerals and Chemicals 11,340 20.5
Bonnie, Florida
Kingsford, Florida
Noralyn, Florida
Agrico Chemical Co. (Williams) 8,618 15.6
Pierce, Florida
Ft. Green, Florida
Occidental Agricultural Chemicals 2,722 4.9
White Springs, Florida
Mobile Chemical 4,264 7.7
Nichols, Florida
Fort Meade, Florida
Brews ter Phosphate 3,175 5.7
Brewster, Florida
Bradley, Florida
U.S. Steel-Agri-Chem, Inc. 1,814 3.3
Ft. Meade, Florida
Gardinier 1,966 3.6
Ft. Meade, Florida
Swift Chemical 2,903 5.3
Bartow, Florida
W.R. Grace & Company 4,808 8.7
Hookers Pr. , Florida
Bonnie Lake, Florida
Manatte Co., Florida
Borden Chemical Company 907 1.6
Teneroc, Florida
Big Four, Florida
T-A Minerals 454 °'8
Polk City, Florida
6.1-3
-------
Table 6.1-1. Phosphate rock producers and capabilities (EPA79)
—continued
Company and location
Beker Industries
Dry Valley, Idaho
J.R. Simplot
Ft. Hall, Idaho
Cominco-American
Garrison, Montana
George Re 1 yea
G- rrison, Montana
Texasgulf
Aurora, North Carolina
Stauffer Chemical Company
1977 production
(Metric tons)
(103)
1,089
1,814
249
91
4,536
1,950
Percent of
total
2.0
3.3
0.5
0.2
8.2
3.5
Mt. Pleasant, Tennessee
Vernal, Utah
Wooley Valley, Utah
Hooker Chemical Company
Columbia, Tennessee
Presnell Phosphate
Columbia, Tennessee
Monsanto Industrial Chemical Co.
Columbia, Tennessee
Henry, Idaho
454
454
1,814
0.8
0.8
3.3
Summary by Region
Location
Florida
North Carolina
Tennessee
Western States
Percent of total U.S..
78.3
7.8
4.1
9.8
6.1-4
-------
North Carolina possesses a rich phosphate rock deposit in Beaufort
County along the Pamlico River. Texasgulf, the only company currently
exploiting this resource, recently expanded plant capacity by 43
percent and has plans for further expansion. Another company has
announced plans for a large operation in Washington, North Carolina.
The western U.S. phosphate rock industry is located in eastern
Idaho, northern Utah, western Wyoming, and southern Montana. This area
accounts for almost six million metric tons per year of the U.S.
capacity, or about 10 percent. Six companies currently operate seven
mines and six processing plants.
The U.S. industry is relatively concentrated as the 10 largest
producers control about 84 percent of the capacity. The two largest
companies control over 34 percent. In the Florida region, two firms
have nearly 44 percent of the State's capacity, while the five largest
companies control over 70 percent (EPA79).
The principal ingredient of the phosphate rock that is of economic
interest is tricalcium phosphate, Ca3(P04>2. However, phosphate
rock also contains appreciable quantities of uranium and its decay
products. The uranium concentration of phosphate rock ranges from 20
to 200 ppm which is 10 to 100 times higher than the uranium
concentration in natural material rocks and soils (2 ppm). The
radionuclides of significance which are present in phosphate rock are:
uranium-238, uranium-234, thorium-230, radium-226, radon-222, lead-210,
and polonium-210. Because phosphate rock contains elevated
concentrations of these radionuclides, handling and processing the rock
can release radionuclides into the air either as dust particles, or in
the case of radon-222, as a gas.
6.1.2 Process Description
After phosphate rock has been mined and beneficiated, it is
usually dried and ground to a uniform particle size to facilitate
processing. The drying and grinding operations produce significant
quantities of particulate material (phosphate rock dust).
Phosphate rock is dried in direct-fired rotary or fluidized-bed
dryers. The rock contains 10-15 percent moisture as it is fed to the
dryer and is discharged when the moisture content reaches 1-3 percent.
Dryer capacities range from 5 to 350 tons per hour (tph), with 200 tph
a representative average.
Crushing and grinding are widely employed in the processing of
phosphate rock. Operations range in scope from jaw crushers which
reduce 12-inch hard rock to fine pulverizing mills which produce a
product the consistency of talcum powder. Crushing is employed in some
locations in the western field; however, these operations are used for
less than 12 percent of the rock mined in the U.S. Fine pulverizing
mills or grinders are used by all manufacturers to produce fertilizer.
Roller or ball mills are normally used to process from 15 to 260 tph.
6.1-5
-------
Some phosphate rock must be calcined before it can be processed.
The need for calcining is determined primarily by the quantity of
organic materials in the beneficiated rock. Since Florida rock is
relatively free of organics, it usually is not calcined. Most
calcining is done in fluidized-bed units, but rotary calciners are also
used. The rock is heated to 1400°-1600° F in the calciner to remove
unwanted hydrocarbons. Calciners range in capacity from 20 to 70 tph;
a representative average is about 50 tph (EPA79).
6.1.3 Control Technology (TRW82a)
At phosphate rock plants, the normal sequence of operation is:
mining, beneficiation, conveying of wet rock to and from storage,
drying or calcining, conveying and storage of dry rock, grinding, and
conveying and storage of ground rock.
Over 98 percent of the phosphate rock produced in the United
States is mined from ground where the moisture content is high enough
to preclude particulate emissions during extraction of the ore. In the
relatively small -amount of mining performed in areas where ground
moisture content is not sufficient to prevent emissions, such as the
hard rock areas of Utah and Wyoming, some particulates are generated
during blasting and handling of the overburden and ore body. These
emissions are minimized by wetting the active mining area with water
from tank trucks.
Beneficiation is performed in a water slurry. Since the rock is
wet, it does not become airborne and presents no particulate problem.
Mined rock is normally moved by conveyor belts. Some are open, others
closed for weather protection. In all except the relatively small
plants in the hard rock areas of Utah and Wyoming, the high moisture
content of the rock prevents emission of particulates. Weather-
protected conveyors also offer some emission control in arid or windy
locations.
Particulates from conveying and storage of ground rock are due
primarily to fugitive emissions. Conveying and storage of ground rock
usually takes place in totally enclosed systems, where proper
maintenance will minimize fugitive losses.
Particulate emissions from dryers, calciners, and grinders could
be reduced by applying particulate control equipment to "non-fugitive"
emission sources.
Controlled emission levels from dryers and calciners can vary
considerably from unit to unit, even with the same control device, due
primarily to the effects of feed rock characteristics. Industrial
representatives have indicated that feed rock characteristics greatly
outweigh the effects of dryer or calciner unit types. Several feed
rock characteristics can affect the emission levels and particle size
6.1-6
-------
distribution of the exhaust gas streams. Surface properties affect
emission levels; rough or pitted surfaces can have greater clay
adhesion, resulting in higher emission levels and smaller average
particle size.
During beneficiation, the least-washed rock will have more fines,
higher emission levels, and smaller average particle size. The
residence time during which the rock is dried or calcined may also
affect emission levels. Although increasing the residence time may
lower particulate concentration per volume of exhaust gas, the total
weight of particulate emission per weight of feed rock will increase.
Other feed rock characteristics can also cause fluctuations in the
particulate emission levels.
Coarse pebble rock from Florida is beneficiated the least and has
the longest residence time in the dryer of all Eastern rock. Along
with other properties, including hardness and clay adhesion, these
properties cause coarse pebble rock to produce the most adverse, or
worst-case, control levels for Eastern operations. However,
unbeneficiated Western rock has a slightly smaller average particle
size than Eastern rock and represents the most adverse of all feed rock
control situations.
Dryer and Calciner Controls
Phosphate rock calciners and dryers have similar emission
characteristics. Scrubbers are the most common control device used in
the operation of phosphate rock dryers and calciners. Probably the
most important design parameters for scrubbers are the amount of
scrubber water used per unit volume of gas 'treated (liquid-to-gas
ratio) and the intimacy of contact between the liquid and gas phases.
The latter parameter is generally related to the pressure drop across
the scrubber*NxBecause of the similarities in emissions from dryers and
calciners, scrubbers can attain similar reduction efficiencies; up to
greater than 99.0 percent for high-energy venturi scrubbers.
Electrostatic precipitators (ESP) can be an economical control
technique. Plate (electrode) voltage and the ratio of plate area to
the volume of gas to be treated are the most important design
parameters of an ESP. Particle resistivity and the ease of cleaning
collected dust from the plates also affect ESP performance.
Electrostatic precipitation is sometimes an economically attractive
control technique in cases where fine dust particles predominate.
Removing fine particles with a venturi scrubber requires relatively
large power inputs (high pressure drops) to achieve the necessary
efficiency. If power cost savings effected by the ESP exceed the
increased capital charges, this system can be more economical than the
venturi scrubber.
Two phosphate rock dryers now use electrostatic precipitators.
One has a conventional dry ESP to control emissions from two rotary
6.1-7
-------
dryers. The precipitator was designed for 95 percent efficiency, but
typically operates at 93 percent. The other uses a wet ESP to control
emissions from two dryers operated in parallel, one a rotary design and
the other a fluid bed. The ESP was designed for an efficiency of 90
percent, but is probably operating at a higher efficiency because the
gas flow rate is about 60 percent of design capacity. With variation
in plate voltage and plate area, ESP's can be designed to achieve
reduction efficiencies up to greater than 99 percent. A calciner at
one existing operation has a two-stage, dry ESP which operates with an
indicated overall efficiency of 99.8 percent.
No fabric filters are known to be in use for phosphate rock dryer
and calciner emission control. Many industry members believe that
moisture condensation would be a major problem because water droplets
could mix with the clay-like dust mat formed on the fabric media and
cause a mud cake. Were this condition to occur, it would "blind" the
bags. Furthermore, since the dust usually has no economical value, dry
recovery for reprocessing is not an attractive incentive to operators.
High exhaust gas temperatures associated with calciners are also
commonly cited as a major difficulty expected with this type control
device. However, manufacturers of these devices believe fabric filters
can be effective for this application. They state that successful
operation of fabric filters are common in more difficult operations,
such as asphalt plants, cement plants, fertilizer dryers, and the clay
industry. Under proper operating conditions, fabric filters generally
exceed 99 percent efficiency.
Grinder Controls
Dried and calcined rock is ground before it is used for the
manufacture of fertilizers. The grinding or milling circuit operates
under slightly negative pressure to prevent the escape of gases
containing ground rock dust. The system is not airtight; hence, the
air that is drawn into the system must be vented. This vent stream
usually discharges through a fabric filter or, sometimes, a wet
scrubber. Electrostatic precipitators are not used for this operation
at existing facilities.
Fabric filters are normally used to control emissions from
grinders, probably because the dust collected by a fabric filter can be
added directly to the product and thereby increase yields. Also, the
low moisture content of 5 percent or less and low temperatures make
fabric filtration technically and economically feasible. A well
maintained and operated baghouse routinely controls particulate
emissions to levels greater than 99 percent.
In some plants higher moisture content of the ground rock dust
causes difficulty. At these plants, wet collectors are usually chosen
for control. These devices can typically control emissions from 90 to
6.1-8
-------
98 percent depending on the pressure drop. There has been a recent
move toward wet grinding of rock for the manufacture of wet-process
phosphoric acid (WPPA). The rock is ground in a water slurry, then
added to the WPPA reaction tanks without drying. This offers the
advantages of lower fuel costs and ability to meet more stringent
particulate emission regulations. Two companies are now using the wet
grinding process.
6.1.4 Radionuclide Emission Measurements
Phosphate rock dust is a source of particulate radioactivity in
the atmosphere because the dust particles have approximately the same
specific activity (pCi/g) as in the phosphate rock. Very limited data
are available for actual field measurements of radioactivity in
dryer/grinder air emissions. Measurements made by EPA (EPA78) are
summarized in Table 6.1-2.
Table 6.1-2. Radionuclide stack emissions measured
at phosphate rock dryers (EPA78)
Parameter
Dryer 1
Dryers 3 and 4
Total particulates (g/y)
Operating time (hr/y)
Stack emissions (Ci/y)
Uranium-234
Uranium-235
Uranium-238
Thorium- 227
Thorium-228
Thorium- 2 30
Thorium-232
Radium-226
2.2E+7
4114
7.0E-4
3. OE-5
6.6E-4
5. OE-5
1.4E-4
9. 7E-5
3. OE-5
9. 3E-4
5.0E+7
4338
2.6E-3
2.4E-4
2.7E-3
2.0E-4
2.3E-4
2.5E-3
8. OE-5
2.9E-3
In estimating the radionuclide emissions from phosphate rock
processing plants in the following sections, the emissions for the
calciner plants are assumed to be similar to those of dryer plants
(i.e., the radionuclide concentration of the particulates emitted to
air is similar to the concentration in the phosphate rock processed).
However, unlike dryer plants, no measurements of radionuclide emissions
from calciner plants have been made. Therefore, some uncertainty
6.1-9
-------
exists as to the validity of the above assumption with respect to
polonium-210 emissions. Because calciners operate at higher
temperatures than dryers, polonium-210 may be volatized from the
phosphate rock and emitted in larger quantities than the other
radionuclides in the uranium-238 decay series. In this case, our
assumption that calciner plant emissions are similar to dryer plant
emissions would underestimate the polonium-210 emissions. EPA is
planning to conduct radionuclide emission studies at calciner plants to
resolve this uncertainty.
6.1.5 Reference Plant
Table 6.1-3 describes the parameters of a reference phosphate rock
drying and grinding plant which are used to estimate the radioactive
emissions to the atmosphere and the resulting health impacts. The
radioactive emissions from the reference plant are listed in Table
6.1-4. These emissions are representative of dryers with low energy
scrubbers which releases 130 grams of particulates per MT of rock
processed and of grinders with medium energy scrubbers which release 25
grams of particulates per MT of rock processed.
Table 6.1-3. Reference phosphate rock drying and grinding plant
Parameter
Dryers
Grinders
Number of units(a)
Phosphate rock processing
rate (MT/y)
Operating factor (hr/y)
3
2.7E+6
6570
4
1.2E+6
6460
Uranium-238 content of
phosphate rock (pCi/g)(b'
Stack parameters
Height (meters)
Diameter (meters)
Exit gas velocity (m/s)
Exit gas temperature (°C)
Type of control system
Particulate emission rate (g/MT)
40
20
2
10
60°
Low energy
scrubber
40
20
2
10
60°
Medium energy
scrubber
130 (0.26)(c) 25
(a)pryer units process 145 MT/hr; grinder units process 45 MT/hr.
^^Uranium-238 is assumed to be in equilibrium with its daughter
products.
(c)values in Ib/ton.
6.1-10
-------
Table 6.1-4. Radionuclide emissions from the reference
phosphate rock drying and grinding plant
Radionuclide Emissions (Ci/y)
Dryers Grinders
Uranium-238
Uranium- 2 34
Thorium-230
Radium-226
Lead-210
Polonium-210
1.4E-2
1.4E-2
1.4E-2
1.4E-2
1.4E-2
1.4E-2
l.OE-3
l.OE-3
l.OE-3
l.OE-3
l.OE-3
l.OE-3
6.1.6 Health Impact Assessment of Reference Plant
The estimated annual radiation doses from radionuclide emissions
from the reference phosphate rock drying and grinding plant are listed
in Table 6.1-5. These estimates are for a model site in central
Florida with a regional population of 1.4E+6. The maximum individual
is located 750 meters from the plant.
Table 6.1-6 presents estimates of the maximum individual lifetime
risk and the number of fatal cancers per year of operation from these
doses.
The lifetime risk to the maximum individual is estimated to be
about 3E-5 and the number of fatal cancers per year of operation is
estimated to be 3E-3. These risks result primarily from doses to the
lung from inhalation of radioactive particulates released from drying
operations.
6.1.7 Alternative Control Technology
The annualized costs and risk reductions achieved by adding
alternative controls to the reference phosphate rock drying and
grinding plant are shown in Table 6.1-7. Two alternative levels of
control are evaluated for dryers:
1. Reduction of the particulate emissions to 50 g/MT through the
use of medium energy venturi scrubbers or ESP's.
2. Reduction of the particulate emissions to 30 g/MT (level of
New Source Performance Standards—NSPS) through the use of
high energy venturi scrubbers or high energy ESP's.
For grinders, only one alternative level of control is evaluated; the
reduction of the particulate emissions to 6 g/MT (level of NSPS)
through the use of fabric filters or high energy venturi scrubbers.
6.1-11
-------
Table 6.1-5. Annual radiation dose from radioactive particulate
emissions from the reference phosphate rock drying and grinding plant
_ Maximum individual Regional population
Organ /• / \ / i \
& (mrem/y) (person-rem/y)
Lung 7.2 6.0E+1
Endosteal 1.5E+1 1.1E+2
Red marrow 1.3 9.2
Kidney 1.0 6.8
Weighted Sum 2.7 2.2E+1
Table 6.1-6. Fatal cancer risks due to radioactive emissions
from the reference phosphate rock drying and grinding plant
s Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Dryers 3E-5 3E-3
Grinders 2E-6 2E-4
Total 3E-5 3E-3
6.1-12
-------
Table 6-1-7. Annualized cost and risk reductions of alternative/ ,
controls for the reference phosphate rock drying and grinding plant^3'
Process
Dryers^b)
Grinders
Emission
Control rate
f -. •. „ „ -
option*- "'
Existing
B-l
B-2
A-l
A-2
Existing
A-l
A-2
(g/MT)
130
50
50
30
30
25
g(d)
6
Total
Fatal cancer risks
annual Individual
cost (^lifetime
($1,000)
861
1770
1000
2320
—
124
4
risk
3E-5
1E-5
1E-5
7E-6
7E-6
2E-6
5E-7
5E-7
Population
(cancers/y of
operation)
3E-3
1E-3
1E-3
7E-4
7E-4
2E-4
5E-5
5E-5
Cost/fatal
cancer
avoided
(in millions)
430
885
435
1000
-
825
27
dryers: B-l = venturi scrubber (15" W.G.)
B-2 = ESP
A-l = venturi scrubber (25" W.G.)
A-2 = high energy ESP
For grinders: A-l
A-2
venturi scrubber (16" W.G.)
fabric filter
^'Incremental cost for installing and operating alternative control
system (i.e., cost above the existing costs).
of control for New Source Performance Standards.
6.1.8 Total HealthImpact of Phosphate Rock Processing Plants
Phosphate rock processing plants (dryers, calciners, and grinders)
release about 5500 MT of particulate matter per year with the existing
level of control (TRW82). This particulate matter contains about
220 mCi of uranium-238 and each of its daughter products. These
emissions are estimated to cause about 5E-2 fatal cancers per year of
operation. This estimate was derived from a ratio of the amount
particulate matter released from all plants to the amount released from
the reference facility:
Number of fatal cancers = 5500 MT PM/yr x _Q()32 HE/yr (reference
per year from all plants 38Q MT m/yr facility)
0.046
6.1-13
-------
6.1.9 Costs and Risk Reductions for Retrofitting Existing Plants
The industry incremental annualized costs to retrofit existing
phosphate dryer, calciner and grinding units are shown in Table 6.1-8.
To retrofit existing dryers with medium energy venturi scrubbers
would cost an additional $6 million per year and would avoid 0.007
fatal cancers/year, or a cost of $785 million per fatal cancer
avoided. Retrofitting to the NSPS level (Control Option A) would cost
an additional $12 million per year and avoid 0.02 fatal cancers per
year, or a cost of $610 million per fatal cancer avoided.
To retrofit existing calciners with medium energy venturi
scrubbers would cost about an additional $3 million per year and would
avoid 0.002 fatal cancers per year, or a cost of $1.4 billion per fatal
cancer avoided. Retrofitting to the NSPS level (Control Option A)
would cost an additional $12 million per year and would avoid 0.006
fatal cancers per year, or a cost of $2 billion per fatal cancer
avoided.
Retrofitting the existing grinders to the NSPS levels (Control
Option A) would cost an additional $340,000 per year and avoid 0.002
fatal cancers per year, or a cost of $170 million per fatal cancer
avoided.
Table 6.1-8. Industry annualized costs and risk reductions for
retrofitting existing phosphate rock dryers, calciners and grinders^3)
„ ._ Control,,.. Total cost
Process unit . (b) , . t, . *
option (millions)
(c)
Fatal cancers
avoided/y
Cost/fatal
cancer avoided
(in millions)
Dryers
Calciners
Grinders
B
A
B
A
A
5.5
12.2
2.7
12.3
0.34
7E-3
2E-2
2E-3
6E-3
2E-3
785
610
1350
2000
170
(a>TRW82.
(b'For dryers and calciners Option B is a venturi scrubber (15" W.G.)
and Option A is a venturi scrubber (25" W.G.). For grinders, Option
A is a fabric filter.
^'Incremental cost for installing and operating alternative control
system (i.e., costs above existing costs).
6.1-14
-------
REFERENCES
Da68 Dames and Moore, Airborne Radioactive Emission Control
Technology, Report on EPA Contract 68-01-4992, White Plains,
New York, Unpublished.
EPA78 Environmental Protection Agency, Radiation Dose Estimates due
to Air Particulate Emissions from Selected Phosphate Industry
Operations, ORP/EERF-78-1, Office of Radiation Programs,
Montgomery, Alabama, 1978.
EPA79 Environmental Protection Agency, Phosphate Rock Plants,
Background Information for Proposed Standards,
EPA-450/3-79-017, Office of Air Quality Planning and
Standards, Research Triangle Park, North Carolina, 1979.
TRW82 Particulate Emissions and Control Costs of Radionuclide
Sources in Phosphate Rock Processing Plants. A report
prepared by Butch Smith (TRW) for Office of Radiation
Programs, December 1982.
6.1-15
-------
6.2 Wet Process Fertilizer Plants
6.2.1 General Description
Most phosphate rock produced in the United States is used for the
production of high-analysis agricultural fertilizers. In 1976, 50
million metric tons of phosphate rock were used to produce 9 million
metric tons of phosphoric acid, the starting material for ammonium
phosphate and triple superphosphate fertilizers (EPA79).
6.2.2 Process Description
Wet process phosphoric acid is produced by mixing ground phosphate
rock with 93 percent sulfuric acid and water. In the process gypsum
(calcium sulfate) is produced as a byproduct. The simplified overall
reaction is represented by:
+ 9H2S(>4 + 18H20 = 6H3P04 + 9CaS04 • 2H20 (1)
Phosphate rock is not the pure compound indicated above, but a
f luoroappitite material containing minor quantities of flourine, iron,
aluminum, silica and uranium. Following the reaction in the digester,
the mixture of phosphoric acid and gypsum is pumped to a filter which
mechanically separates the particulate gypsum from the phosphoric acid
(approximately 30 percent phosphorus pentoxide concentration). An
enormous amount of the byproduct gypsum is produced— each 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 (EPA79).
The phosphoric acid is then used to produce several different
grades of agricultural fertilizers. Triple superphosphate (TSP) •
fertilizer is made using ground phosphate rock and phosphoric acid as
in the following equation:
Ca3(P04)2 + 4H3P04 + 3H20 = 3Ca(H2P04)2 • 2H20 (2)
Ammonium phosphate fertilizer is made using ammonia and wet process
phosphoric acid. Monoammonium phosphate (MAP) and diammonium phosphate
(DAP) are produced as in the following equations:
H3P04 + NH3 - HHP0 MAP (3)
H3P04 + 2NH3 = (NH4)2HP04 DAP
The steps involved in the wet process production of agricultural
fertilizers are summarized in Table 6.2-1. The major sources of
radionuclide emission in particulate dust results in the product drying
and handling areas.
6.2-1
-------
Figure 6.2-1. Flow diagram of the wet process (EPA79).
6.2-2
-------
6.2.3 Control Technology (TRW82)
Production processes for diammonium phosphate (DAP) and granular
triple superphosphate (GTSP) are similar. The same process equipment
in certain plants is used to produce both DAP and GTSP on an
alternating basis; therefore, the control equipment for DAP and GTSP
processes is similar. The particulate matter emission points within
the DAP and GTSP production processes are as follows:
- reactor/granulator exhaust(s);
- dryer exhaust;
- cooler exhaust where appropriate; and
- screens, mills, and materials handling ventilation system(s)
and exhaust(s).
Additional particulate matter (PM) emission sources exist in the ground
rock raw materials handling (GTSP only) and final product handling
systems (DAP and GTSP). These sources, however, are mostly "fugitive"
sources and not process sources.
The DAP and GTSP processes currently in operation employ a variety
of wet scrubbing systems on each of the major process exhaust streams.
In most instances, scrubbers are installed in series. Generally,
individual scrubbing systems are designated as "primary," "secondary,"
etc., referring to their order in the series of control devices.
Scrubbing systems have not been installed to control particulate
matter; rather, process economic considerations and flouride emissions
control have prompted installation of the scrubbing systems. In the
DAP process, the primary scrubber uses phosphoric acid as a scrubbing
solution to recover ammonia raw materials that otherwise would be
lost. Without ammonia recovery, the cost of manufacturing DAP is not
competitive. Secondary scrubbing systems have been installed by and
large to control flouride emissions, to ensure worker safety, and to
meet environmental regulations. Secondary scrubbing systems generally
use recirculated process water (pond water) to enhance flouride
removal. Some plants operate tertiary scrubbers for the same reasons.
The primary, secondary, and sometimes tertiary scrubbing systems,
however, also control particulate matter emissions.
The control technologies that can be applied to these PM emission
sources include:
- cyclone systems;
- wet scrubbing systems;
- bag filters; and
- electrostatic precipitators.
In practice, however, electrostatic precipitators have not been
the technology of choice. Moreover, the use of bag filters has been
limited to the cooler exhausts from certain processes and product
6.2-3
-------
screening, milling and handling ventilation system exhausts. This is
primarily because the major PM emission points (the reactor granulator
exhausts, dryer exhausts, and cooler exhausts on certain processes) are
also emission points for other pollutants. In particular, gaseous
flouride emissions (GTSP and DAP) and gaseous ammonia emissions (DAP
only) are largely unaffected by electrostatic precipitators or
baghouses. In addition, the moisture in the reactor and dryer exhaust
streams and the sticky nature of the particulate matter in these
streams complicates the use of bag filter devices. Consequently, PM
control technologies applicable to DAP and GTSP production processes
are realistically limited to dry cyclone systems, wet scrubbing
systems, and bag filters (for dry materials handling sources only).
Dry cyclone systems are routinely employed on dryer, cooler,
screens, and milling operation exhausts to recover entrained product
that otherwise may be lost. As such, the cyclone systems are as much a
part of the process as they are control equipment.
Controls in place were estimated in a survey of 14 plants (25 DAP
and 14 GTSP processes) based on state air permit files and
conversations with plant personnel. Although 100 percent of the DAP
and GTSP production in the United States is not represented in the
survey, based on published production capacity data, greater than 90
percent of domestic production is represented. It was found that
primary scrubbing systems are employed on 100 percent of the existing
processes. Venturi scrubbers make up about 60 to 95 percent of the
primary scrubbers. In addition, secondary scrubbing systems are
employed on about 60 to 80 percent of the existing processes. About
half of the secondary scrubbers in the industry are packed bed
scrubbers. Tertiary scrubbers also are employed on about 8 to 15
percent of the DAP process units (i.e. reactors, dryers, etc.) and 28
percent of the GTSP process units.
6.2.4 Radionuclide Emission Measurements
EPA has measured radionuclide emission in particulate stack
releases at two wet process phosphate fertilizer plants (EPA78). The
samples were collected on product dryer stacks in accordance with EPA
guidelines established in the Code of Federal Regulations, Title 40,
Part 60. The annual emission rates based on these measurements are
listed in Table 6.2-1.
6.2.5 Reference Facility
Table 6.2-2 describes the parameters of a reference wet process
phosphate fertilizer plant which are used to estimate the radionuclide
emissions to the atmosphere and the resulting health impacts. The
reference plant produces both diammonium phosphate (DAP) and granular
triple superphosphate (GTSP) from phosphoric acid derived from
phosphate rock. The radionuclide emissions to air from the DAP and
GTSP process stacks of the reference facility are listed in Table
6.2-3. The emissions are representative of plants using only primary
scrubbers to control DAP and GTSP process off gases.
6.2-4
-------
Table 6.2-1. Radionuclide stack emissions at wet process
phosphate fertilizer plants (EPA78)
Parameter
Total particulates (g/y)
Operating time (hr/y)
Stack emissions (Ci/y)
Uranium-234
Uranium-235
Uranium-238
Thorium-227
Thorium-228
Thorium- 2 30
Thorium-232
Radium-226
Polonium-210
TSP dryer
Plant A
2.0E+7
4. 6E+3
1.1E-4
m
9.0E-5
ND
4.0E-5
9. OE-5
ND
3. OE-5
6.3E-4
TSP dryer
Plant B
1.2E+7
7.4E+3
3.0E-4
2. OE-5
2.7E-4
ND
3. OE-5
2. 5E-4
7. OE-5
2.2E-4
NA
DAP dryer
Plant B
1.5E+7
7.5E+3
2.6E-3
1 . 9E-4
3. 3E-3
ND
8. OE-5
3.0E-3
5. OE-5
2.6E-4
NA
ND Not detectable.
NA Not available.
6.2.6 Health Impact Assessment of Reference Plant
The estimated annual radiation doses from radionuclide emissions
from the reference wet process phosphate fertilizer plant are listed in
Table 6.2-4. These estimates are for a model site in central Florida
with a regional population of 1.4E+6. The maximum exposed individual
is located 1500 meters south of the reference plant.
Table 6.2-5 presents estimates of the individual lifetime risk and
the number of fatal cancers per year of operation from these doses.
The lifetime risks to the maximum individual is estimated to be
about 5.E-6 and the number of fatal cancers per year of operation is
estimated to be l.E-3. These risks result primarily from doses to the
lung from inhalation of radioactive particulates released from fertil-
izer production.
6.2.7 Alternative Control Technology
All wet process phosphate fertilizer plants use primary scrubbers
on the DAP and GTSP exhausts. The annualized costs and risk reduction
of adding alternative controls to the reference wet process phosphate
fertilizer plant are shown in Table 6.2-6.
6.2-5
-------
Table 6.2-2. Reference wet process phosphate fertilizer plant
Parameter
Process
DAP
GTSP
Production rate (MT/y) 5.2E+5 2.7E+5
Operating factor (hr/y) 8160 8160
Radionuclide content of
product (pCi/g)(a)
Uranium-238, uranium-234, thorium-230 60 60
Radium-226 5 20
Lead-210, polonium-210 30 30
Stack parameters
Height (meters) 40 40
Diameter (meters) 2 2
Exit gas velocity (m/s) 10 10
Exit gas temperature (°C) 60 60
Type of control system Venturi Venturi
scrubber scrubber
Particulate emission rate (g/MT) 164 100
-------
Table 6.2-4. Radiation dose rates from radionuclide emissions
from the reference wet process phosphate fertilizer plant
Organ Maximum individual Regional population
(mrem/y) (Person-rem/y)
1.2 2.4E+1
Endosteal 2.2 4.1E+1
Red marrow 1.5E-1 2.7
Kidney 6. 3E-2 1.3
Weighted sum 4.1E-1 8.0
Table 6.2-5. Fatal cancer risks due to radioactive emissions
from reference wet process phosphate fertilizer plant
Lifetime risk Regional population
Source to maximum individual fatal cancers/y of operation
DAL and GTSP
process emissions 5E-6 1E-3
6.2.8 Total Health Impact of Wet Process Phosphate Fertilizer Plants
Wet process phosphate fertilizer plants release about 1500 MT per
year of particulates from the DAP and GTSP process stacks with the
existing control systems. This amount of particulate matter contains
about 90 mCi each of uranium-238, uranium-234, and thorium-230 and
lesser quantities of radium-226, polonium-210, and lead-210. This
estimate is based on the conservative assumption that the specific
activity (pCi/g) of the particulate material released is the same as
DAP and GTSP fertilizers. These emissions are estimated to cause about
0.02 fatal cancers per year. This estimate is based on a ratio of the
amount of particulate material released from all plants to the amount
released from the reference plant in a manner similar to that shown in
Section 6.1.8.
6.2.9 Costs and Risk Reductions for Retrofitting Existing Plants
The annualized costs to the industry to retrofit existing
phosphate fertilizer plants with secondary scrubbers are shown in Table
6.1-7. To retrofit existing DAP process exhausts with packed bed
6.2-7
-------
scrubbers (28 percent of the existing production capacity) would cost
an additional $3 million per year and would avoid 0.002 fatal cancers
per year, or a cost of $1.5 billion per fatal cancer avoided.
Retrofitting GSTP process exhausts with packed bed scrubbers (19
percent of existing production capacity) would cost an additional
$500,000 per year and would avoid 0.0008 fatal cancers per year, or a
cost of $6.5 billion per fatal cancer avoided.
Table 6.2-6. Annualized costs and risk reductions of
alternative controls for the reference wet process
phosphate fertilizer
Total
Fatal cancer risks
p Emission annual Individual
rocess control rate cost (^lifetime
DAP
GTSP
option^
Existing
Alternative
Existing
Alternative
(g/MT) ($1,000)
164
100 500
100
79 300
risk
4E-6
2E-6
1E-6
8E-7
Population
(cancers/y of
operation)
8E-4
5E-4
2E-4
1.6E-4
Cost/fatal
cancer
avo ided
(in millions)
1.7E+3
7.5E+3
(^Source: TRW82.
(^'Existing controls are venturi scrubbers. Alternative controls are
packed bed scrubbers in series with venturi scrubbers.
'c'Particulate material emission rate.
(^Incremental cost for installing and operating alternative control
systems, i.e., additional costs for installing and operating
packed bed scrubbers.
Table 6.2-7. Industry annualized costs and risk reductions for
adding secondary scrubbers to existing wet process phosphate
fertilizer plants^3'
Process
Total cost
(millions)
(b)
Fatal cancers
avoided/y
Cost/fatal
cancer avoided
(in millions)
DAP
GTSP
3
0.5
2E-3
8E-5
1.5E+3
6.3E+3
Incremental cost of installing and operating packed bed scrubbers
in series with existing venturi scrubbers. Twenty-eight percent of
DAP production capacity and 19 percent of GTSP production capacity
require retrofit.
6.2-8
-------
REFERENCES
EPA78 Environmental Protection Agency, Radiation Dose Estimates due
to Air Particulate Emissions from Selected Phosphate Industry
Operations, ORP/EERF-78-1, Office of Radiation Programs,
Montgomery, Alabama, 1978.
EPA79 Environmental Protection Agency, Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States,
EPA-520/7-79-006, Office of Radiation Programs, Washington,
B.C., 1979.
TRW82 Industry and Particulate Matter Control Technology Information
for Diammonium Phosphate and Granular triple Superphosphate
manufacture. A report prepared by TRW Environmental Division
for the Environmental Protection Agency, Dec 15, 1982.
6.2-9
-------
6.3 Elemental Phosphorus Plants
6.3.1 General Description
About ten percent of the marketable phosphate rock mined in the
United States is used for the production of elemental phosphorus.
Elemental phosphorus is used primarily for the production of high grade
phosphoric acid, phosphate-based detergents, and organic chemicals. In
197,7 approximately 285 thousand metric tons of elemental phosphorus
were produced from 4 million metric tons of phosphate rock.
Phosphate rock contains appreciable quantities of uranium and its
decay products. The uranium concentration of phosphate rock ranges
from about 20 to 200 ppm, which is 10 to 100 times higher than the
uranium concentration in natural rocks and soil (2 ppm). The
radionuclides of significance which are present in phosphate rock are:
uranium-238, uranium-234, thorium-230, radium-226, radon-222, lead-210,
and polonium-210. Because phosphate rock contains elevated
concentrations of these radionuclides, handling and processing this
material can release radionuclides into the air in the form of dust
particles. More importantly for elemental phosphorus plants, heating
the phosphate rock to high temperatures in calciners and electric
furnances can volatilize lead-210 and polonium-210, resulting in the
release of large quantities of these radionuclides into the air.
There are 8 elemental phosphorus plants in the United States—
located in Florida, Idaho, Montana, and Tennessee. Table 6.3-1 shows
the owners, locations, and the estimated elemental phosphorus
production rates for these plants.
6.3.2 Process Description
Phosphate rock which has been crushed and screened is fed into
calciners where it is heated to the melting point, usually 1300° C.
The calcining serves two purposes: (1) it burns any organic matter
present in the rock, and (2) it transforms the finely divided rock into
large stable agglomerates or nodules which are needed for proper
operation of the reduction furnaces. The hot nodules are passed
through coolers and then to storage bins prior to being fed to electric
furnaces. The furnace feed consists of the nodules, silica and coke.
The proper amount of silica is needed to form slag with the flow
properties necessary to facilitate removal from the furnace. Coke is
added as a carbon source to reduce the calcium phosphate to elemental
phosphorus. A simplified chemical equation for the electric furnace
reactor is as follows:
2Ca3(P04)2 + 6Si02 + IOC = ?4 + 10CO + 6CaS103 (1)
6.3-1
-------
In addition, the iron naturally present in the rock reacts with
some of the phosphorus to produce FeP. The blended furnace feed enters
the furnaces continually from the top and progresses downward until
reaching the molten layer on the bottom. Phosphorus and carbon
monoxide (CO) are driven off as gases and are vented near the top of
the furnace. The slag and FeP which are continually collecting in the
furnace are periodically "tapped off."
Furnace off-gasses pass through dust collectors and then through
water spray condensers. Phosphorus is cooled to the molten state in
the condensers. The mix of phosphorus and water—phossy water—and mud
go to a processing system where phosphorus is separated and piped to
storage. The clean off-gases leaving the condensers contain a high
concentration of CO and are used as fuel in the calciners. A flow
diagram of the process is shown in Figure 6.3-1.
Table 6.3-1. Location and size of elemental phosphorus plants'3/
Locatlon
Florida
Pierce Mobil (Electro Phos.) 1.8E+4
Tarpon Springs(b) Stauffer Chemical Co. 2.3E+4
Idaho
Pocatello FMC Corporation 1.3E+5
Soda Springs Monsanto Chemical Co. l.OE+5
Montana
Silver Bow Stauffer Chemical Co. 3.8E+4
Tennessee
Columbia
Columbia
Mt. Pleasant
Hooker Chemical Co.
Monsanto Chemical Co.
Stauffer Chemical Co.
4.1E+4
1.2E+5
4.5E+4
^ a<> Data from TRI81.
(b'Plant is presently shut down.
6.3.3 Control Technology
Emissions from calciners are typically controlled by low energy
scrubbers. Emissions from nodule coolers and transfer points and
furnace tap holes are controlled by either fabric filters or wet
scrubbers. Screening plant emissions are usually controlled by fabric
filters. Fugitive dust emissions and radon gas emissions are not
controlled.
6.3-2
-------
INPUT
PROCESS
PRODUCTS
& BY-PRODUCTS
'PHOSPHATEX
ROCK \
CALCINER
^CSTACK VENT EXHAUST
CALCINED
BRIQUETTE
FERROPHOSPHORUS
SALES
ELECTRIC
SLAG SALES
PRECIPITATOR
ELEMENTAL
PHOSPHORUS SALES
CONDENSERS
CARBON MONOXIDE
FLARE STACK
CARBON
. MONOXIDE
Recycled
Figure 6.3-1. Flow diagram of the thermal process for
production of elemental phosphorus.
6.3-3
-------
6.3.4 Radionuclide Emission Measurements
EPA has measured the radionuclide emissions from three elemental
phosphorus plants (EPA77, AnSla, AnSlb). The stack emission rates
measured during these studies are summarized in Table 6.3-2.
All of the radionuclides are released as particulates except for
radon-222, which is released as a gas. Lead-210 and polonium-210 are
the particulate form radionuclides released in the largest quantities.
Essentially all of the radon-222 and greater than 95 percent of the
lead-210 and polonium-210 emitted from these facilities are released
from the calciner stacks (plants B and C). The high temperature of the
calciners and reduction furnaces volatilize the lead-210 and
polonium-210 from the phosphate rock, resulting in the release of much
greater quantities of these radionuclides than the uranium, thorium and
radium radionuclides.
6.3.5 Reference Facility
Table 6.3-3 describes the parameters of a reference elemental
phosphorus plant which are used to estimate the radioactive emissions
to the atmosphere and the resulting health impacts.
The radioactive emissions to air from the reference facility are
listed in Table 6.3-4. These emissions are representative of a plant
with no radon control which releases 10 percent of the polonium-210 and
5 percent of the lead-210 in the phosphate rock processed. These are
similar to the releases estimated for Plant C in Table 6.3-2 and are
believed to be typical of plants operating with low energy scrubbers on
the calciner exhausts.
6.3.6 Health Impact Assessment of Reference Elemental Phosphorus Plant
The estimated annual radiation doses and working level exposures
from radionuclide emissions from the reference elemental phosphorus
plant are listed in Tables 6.3-5 and 6.3-6. These estimates are for a
southeastern Idaho site with a regional population of 1.4E+5. The
maximum individual is assumed to be located 1500 meters from the plant.
Table 6.3-7 presents estimates of the maximum individual lifetime
risk and the number of fatal cancers to the regional population from
these doses and working level exposures.
The lifetime risk to the maximum individual is estimated to be
1E-4 and the total number of fatal cancer per year of operation is
estimated to be 8E-3. These risks result primarily from polonium-210
emissions from the calciner stack. The risks from radon-222 stack
emissions are small (less than 5 percent) compared to the risks from
the particulate emissions.
6.3-4
-------
Table 6.3-2.
Radionuclide stack emissions measured at elemental
phosphorus plants(a)
Parameter
Plant
Rock processing rat^ (MT/y) 1.6E+6
Uranium-238 concentration
of rock (pCi/g)(c) 22.0
Calciner stacks emission rate (Ci/y):
Uranium-238 1.2E-3
Uranium-234 1.3E-3
Thorium-230 2.2E-3
Radium-226 1.3E-3
Radon-222
Lead-210 3.0E-3
Polonium-210 6.9
Other stacks emission rate (Ci/y):
Uranium-238 4.OE-2
Uranium-234 4.6E-2
Thorium-230 5.3E-3
Radium-226 5.9E-3
Radon-222
Lead-210 1.5E-2
Polonium-210 4.0E-1
Fraction of input radionuclides emitted:
Uranium-238 1.2E-3
Uranium-234 1.4E-3
Thorium-230 2.1E-4
Radium-226 2.0E-4
Radon-222
Lead-210 5,
5.3E+5) 1.7E+6
-------
Table 6.3-3. Reference elemental phosphorus plant
Parameter Value
Phosphate rock processing rate 1.6E+6 MT/y
Elemental phosphorus production rate 1.2E+5 MT/y
Operating factor 7000 h/y
Number of calciners 2
Uranium-238 concentration of
phosphate rock 25 pCi/g(a)
Stack parameters :
Height 30 meters
Diameter 2 meters
Exit gas velocity 15 meters/sec
Exit gas temperature 60° C
Emission control system low energy scrubber
Particulate emissions:
Calciner stack 0.25 kg/MT
Other stacks 0.25 kg/MT
(a)Uranium-238 is assumed to be in equilibrium with its daughter
products.
(b)Parameters for the calciner stack.
Table 6.3-4. Radionuclide emissons from the reference elemental
phosphorus plant
„ ,. ,., Emissions (Ci/y)
Radionuclide _ . . : n -. ;
Calciner stack Other stacks
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-222
Lead-210
Polonium-210
1E-2
1E-2
1E-2
1E-2
4E+1
2.0
4.0
1E-2
IE -2
1E-2
1E-2
_
1E-2
IE -2
6.3-6
-------
6.3.7 Existing Emission Standards and Air Pollution Controls
There are no radionuclide emission standards for elemental
phosphorus plants. However, the states regulate particulate emissions
from elemental phosphorus plants. Although these plants are not
subject to industry-specific standards, they must comply with the
general process source standards set forth in each State Implementation
Plan (SIP). Identical standards have been adopted for plants in Idaho,
Montana, and Tennessee (existing plants only); slightly more stringent
standards have been set in Florida. Table 6.3-8 shows the particulate
emission limits for general process sources for these States.
6.3.8 Alternative Control Technology
The costs and risk reductions achieved by adding alternative
controls to the calciner stacks of the reference plant to reduce
lead-210 and polonium-210 emissions are shown in Table 6.3-9. The
existing baseline level of control for all plants is a low energy
scrubber. However, primary and secondary scrubbers in series are
already in use in at least one plant.
Table 6.3-5. Radiation dose rates from radionuclide emissions
from the reference elemental phosphorus plant
o Maximum individual Regional population
r8an (mrem/y) (person-rem/y)
Lung 36 136
Red marrow 4 13
Kidney 45 133
Endosteal 23 70
Weighted Sum 15 52
The costs presented in Table 6.3-9 are the annualized costs for
installing and operating the alternative control systems. Because
information is not available on the operating costs of the existing
systems, we could not estimate the incremental costs above the existing
costs for these alternative controls. The capital and operating unit
costs used in estimating the costs for the reference plant are
presented in Table 6.3-10. The values of annualized costs in Table
6.3-9 are twice the values in Table 6.3-10 because the reference plant
operates with two calciners.
6.3-7
-------
Table 6.3-6. Annual radon decay product exposures from radon-222
emissions for the reference elemental phosphorus plant
e Maximum individual Regional population
bource (WL-y) (person-WL-y)
Plant Stacks 3.2E-6 1.3E-2
Table 6.3-7. Fatal cancer risks due to radioactive emissions from
the reference elemental phosphorus plant
s Lifetime risk Regional population
°urce to maximum individual (Fatal cancers/y of operation)
Particulates 1E-4 8E-3
Radon-222 5E-6 3E-4
Total 1E-4 8E-3
6.3.9 Total Health Impact of Elemental Phosphorus Industry
The estimated total number of fatal cancers caused by all
elemental phosphorus plants is about 0.05 per year. This estimate was
derived from the ratio of the capacity of the reference plant to the
capacity of each individual plant taking into consideration the
population density of the individual plant site and the radionuclide
concentration of the processed phosphate rock (see Table 6.3-11).
Because this estimate is based on plant capacity and not production
rates, it represents an upper bound estimate of the health impact.
Table 6.3-8. Particulate emission limits for
general process sources(a)
State Particulate emission limits (Ibs/h)
30 tons/h(b) 100 to.ns/h 500 tons/h
Florida
Tennessee, Idaho,
and Wyoming
30
40
36
51
47
79
(a)Data from EPA79.
(b)Material processing rate.
6.3-8
-------
Table 6.3-9. Costs and risk reductions of adding alternative controls
to the calciner stack of the reference elemental phosphorus plant
Type
of
control
Low energy
scrubber
Medium energy
scrubber
High energy
scrubbers
Emission
rate(a>
(g/MT)
250-500
150
50
Level of
polonium-210
control
(%)
90(c)
94(e)
98(c)
Fatal cancer risks
Annual,, ,
cost (b)
($1000)
(d)
1,700
2,000
Individual
lifetime
risk
1E-4
6E-5
2E-5
Fatal
cancers/y
8E-3
5E-3
2E-3
Ca)Particulate matter emission rate.
(b)Reference plant has two (2) calciner units.
(c)Based on emission measurement data from Table 6.3-2.
(d)Existing control.
(e)Estimated value.
Table 6.3-10. Capital and operating costs for alternative control
systems for calciner units of the reference
elemental phosphorus plant
Type of
control
Medium energy
scrubber (a)
High energy
scrubber (b)
Installed
capital cost
($1000)
1,300
1,600
Annual
capital
cost
($1000)
260
320
Annual
operating
cost
($1000)
600
700
Total
annual ized
cost
($1000)
860
1,020
Ca)l5" water gauge.
(b)30" water gauge.
6.3-9
-------
Table 6.3-11.
Total health impact of radionuclides emissions from
elemental phosphorus industry
Location
of plants
Number
of
plants
Plant
capacity
(MT/y)
(a)
Radionuclide
concentration
(pCi/g)
Regional
population
(Persons/km^)
Fatal
cancers/y
Idaho
Tennessee
Florida
Montana
Total
2
3
1
1
3.5E+6
2.9E+6
2.5E+5
5.3E+5
25
5
45
25
7
15
70
5
2E-2
6E-3
2E-2
2E-3
5E-2
(a'Phosphate rock processing rate.
6.3-10
-------
REFERENCES
AnSla Andrews V. E., Emissions of Naturally Occurring Radioactivity
from Stauffer Elemental Phosphorus Plant, ORP/LV-81-4, EPA,
Office of Radiation Programs, Las Vegas, Nevada, August 1981.
AnSlb Andrews V. E., Emissions of Naturally Occurring Radioactivity
from Monsanto Elemental Phosphorus Plant, ORP/LV-81-5, EPA,
Office of Radiation Programs, Las Vegas, Nevada, August 1981.
EPA77 Environmental Protection Agency, Radiological Surveys of Idaho
Phosphate Ore Processing—The Thermal Plant, ORP/LV-77-3, EPA,
Office of Radiation Programs, Las Vegas, Nevada, 1977.
EPA79 Environmental Protection Agency, Phosphate Rock Plants,
Background Information on Proposed Standards, EPA-450/3-79-
0017, USEPA Research Triangle Park, N.C., September 1979.
TRI81 Teknekron Research, Inc., Draft, Partial and Supplemental
Background Information Document—Primary Pyrometallurgical
Extraction Process, Report to Environmental Agency under
Contract No. 68-01-5142, USEPA Docket Number A-79-11, May 1981.
6.3-11
-------
Chapter 7: MINERAL EXTRACTION INDUSTRY FACILITIES
Metal Mines, Mills, and Smelters
Almost all industrial operations involving the removal and
processing of ores to recover metals release some radionuclides into
air. This chapter presents an assessment of the radionuclide emissions
from the aluminum, copper, zinc, and lead industries. These industries
were studied because they involve the processing of large quantities of
ore and because they all involve pyrometallurgical processes which have
the greatest potential for radionuclide emissions. Two types of
assessments are presented in this chapter. One is an assessment of
emissions from a reference facility which, in each case, involves a
high temperature operation. The information for these assessments is
taken directly from a report prepared for EPA by Teknekron Research,
Inc. (TRI81), except for the dose and risk calculations for the zinc
and copper smelters which were made using methodology described in
Appendices A and B..
The second approach is an assessment of emissions measured by EPA
at surveyed facilities. Reports on these emission measurements became
available subsequent to the assessments of the reference facilities.
Assessments are presented for an alumina plant, an aluminum reduction
plant, an underground copper mine and mill, an open pit copper mine and
mill, and a zinc mine and mill.
7.1 Aluminum Industry
7.1.1 General Description
Bauxite is the principal aluminum ore found in nature. The ore is
processed at the mine to produce alumina (A1203), the basic feed in
the aluminum reduction process. Aluminum metal is produced by the
reduction of alumina in a molten bath of cryolite. The production of
aluminum differs from other primary metals in that no purification of
the metal produced in the electric cells is needed; contaminants in the
ore are removed in the milling rather than the smelting phase of the
process.
7.1-1
-------
Of the 12 domestic companies producing primary aluminum, only
Alcoa and Reynolds perform all stages of production, from domestic
mining through the primary metal stage. Almost all of the bauxite used
in aluminum production is imported. Five other domestic firms own
bauxite and/or alumina facilities in other countries and import raw
materials. Only five of the twelve firms that own primary aluminum
plants also own domestic plants producing the input product, alumina.
These five companies (Aluminum Company of America, Kaiser Aluminum and
Chemical Corporation, Reynolds Metals Co., Martin Marietta Aluminum
Co., and Ormet Corp.) own 73 percent of the current U.S. primary
aluminum capacity (St78).
There are currently 32 operating primary aluminum smelters in the
United States (Table 7.1-1). With one exception, all of the plants are
located in rural areas. Population densities in the vicinities of the
plants range from 12 to 62 persons per square kilometer (EPA79).
Table 7.1-1. Location and size of primary aluminum production plants
(TRI81)
Location
Alabama
Arkadelphia
Jones Mills
Listerhill
Scottsboro
Ind iana
Evansville
Kentucky
Hawesville
Sebree
Company
Reynolds Metals Company
Reynolds Metals Company
Reynolds Metals Company
Revere Copper & Brass Co.
Aluminum Company of America
National Southwire
Anaconda Aluminum Company
Capacity
(1000 MT/y)
56
103
166
95
239
148
148
Louisiana
Chalmette
Lake Charles
Maryland
Frederick
Missouri
New Madrid
Kaiser Aluminum & Chemical Corp. 215
Consolidated Aluminum Corporation 30
Eastalco Aluminum Company 145
Noranda 115
7.1-2
-------
Table 7.1-1.
Location and size of primary aluminum production plants
(Continued)
Location
Company
Capacity
(1000 MT/y)
Montana
Columbia Falls
North Carolina
Bad in
New York
Massena
Massena
Ohio
Hannibal
Oregon
The Dalles
Troutdale
Tennessee
Alcoa
New Johnsville
Texas
Point Comfort
Palestine
Rockdale
San Patricio
Washington
Ferndale
Goldendale
Longview
Mead
Ravenswood
Tacoma
Vancouver
Wenatchee
Total
Anaconda Aluminum Company 148
Aluminum Company of America 103
Aluminum Company of America 177
Reynolds Metals Company 104
Ormet Corporation 215
Martin-Marietta Aluminum Co. 75
Reynolds Metals Company 104
Aluminum Company of America 182
Consolidated Aluminum Corporation 119
Aluminum Company of America 153
Aluminum Company of America 13
Aluminum Company of America 268
Reynolds Metals Company 94
Intalco Aluminum Corp. 215
Martin-Marietta Aluminum Company 99
Reynolds Metals Company 174
Kaiser Aluminum & Chemical Corp. 182
Kaiser Aluminum & Chemical Corp. 135
Kaiser Aluminum & Chemical Corp. 66
Aluminum Company of America 95
Aluminum Company of America 173
4354
7.1.2 Process Description
Of the 32 aluminum reduction plants in the United States, all but
one produce aluminum in electric furnaces (cells) by the Hall-Hiroult
7.1-3
-------
process. In the Hall-Hiroult process, alumina (A1203) is reduced
electrolytically in a molten bath of cryolite (NaAlFg). The Aluminum
Company of America's pilot plant in Palestine, Texas, employs aluminum
chloride as the electrolyte.
Two basic types of cells are used by the industry: prebake and
Soderberg. The chief difference between the two types is the means by
which carbon is supplied to the reduction cells. At prebake plants,
both center- and side-worked cells use preformed carbon anodes baked
into a solid mass. Soderberg cells use carbon anode paste which is fed
to the cell continuously.
Both types of reduction cells are operated at termperatures in
excess of 950° C, the melting point of the cryolite. Approximately
2.6 metric tons of raw materials, along with large quantities of
electricity, are required to produce 1 NT of aluminum. The breakdown
of raw materials is shown in Table 7.1-2.
Table 7.1-2. Raw materials used in producing aluminum (EPA77)
Raw material MT Feed/MT Al produced
Alumina (A1203) 1.9
Cryolite (NaAlF6) 0.03-0.05
Aluminum Fluoride (A1F3) 0.03-0.05
Fluorspar (CaF2) 0.003
Petroleum Coke 0.455-0.490
Pitch Binder 0.123-0.167
Carbon (cathode) 0.02
The particulate emissions from the process reflect the composition
of the feed materials, and include alumina, carbon, cryolite, aluminum
fluoride, and trace elements. Generation of particulate emissions
varies with the type of cells. At prebake plants, particulate
emissions from the anode furnace range from 0.5 to 2.5 kg/MT of
aluminum produced, with 1.5 kg/MT being a typical value (EPA76).
Particulate emissions generated by the cells vary from 5.95 to 88.5
kg/Ml, with 40.65 kg/MT being typical (EPA76).
Quality of Feed Materials
No evidence could be found that the quality of feed materials
varies to any significant degree. Radionuclide concentrations for
input materials are given in Table 7.1-3.
7.1-4
-------
Table 7.1-3. Radionuclide concentrations of
feed materials to aluminum plants (EPA82)
Feed material Radionuclide concentration (pCi/g)
Uranium-238 Thorium-232
Alumina 0.10 <0.2
Aluminum Fluoride 0.11 <0.2
Cryolite 0.11 <0.2
7.1.3 Control Technology For Primary Aluminum Reduction Plants
Controls for emissions from aluminum plants are either primary or
secondary controls. Primary controls handle the emissions captured by
the cell hoods, while secondary controls are used to treat the entire
building effluent, including cell emissions that escape the primary
hoods. Primary controls are used at all plants, but secondary controls
are generally used only by the plants that employ Soderberg cells
(EPA79).
Control devices used for primary control vary widely from plant to
plant, and include multicyclones, dry and fluid bed alumina adsorbers
followed by fabric filters or electrostatic precipitators, and spray
towers with spray screens. Not only do the efficiencies of these
devices vary over a considerable range (70-99+ percent), but the
collecting hoods for the various types of cells range from less than 80
percent to greater than 95 percent capture efficiency (EPA79). A more
detailed discussion of control systems for primary aluminum plants is
presented in TRI81.
7.1.4 Radionuclide emissions
Particulate material emitted from an aluminum reduction plant
contains radionuclide concentrations (pCi/gO similar to or greater than
the concentrations in the alumina processed. Because of the high
temperatures of the reduction cells, some radionuclides (particularly
lead-210 and polonium-210) may be volatilized and released in greater
quantities than the other radionuclides in the alumina. EPA has
recently measured the radionuclide emissions from an aluminum reduction
plant. The emission estimates for the reference aluminum reduction
plant are based on preliminary data from these measurements.
7.1.5 Reference Facility
Table 7.1-4 describes the parameters of a reference aluminum
reduction plant which are used to estimate the radionuclide emissions
to air and the resulting health impacts.
7.1-5
-------
Since the currently operating facilities have similar particulate
emission rates and use roughly the same process and feed stocks, one
reference plant characterizes the primary aluminum source category. It
uses center-worked prebake cells, the most commonly used equipment now
in operation. The capacity chosen (136,000 metric tons/y of aluminum)
is approximately the average size of all existing plants. A capacity
factor of 0.94 is applied to the plant, the 1979 industry-wide average
(DOI80).
Since no stack parameters were available for the main stack,
default values of 30 meters for the height and 1.8 meters for the
diameter are used. Exhaust gas flow rates were determined by scaling
available flow rates linearly with the difference in capacity between
an actual facility and the reference plant.
As of 1975, 95 percent of all plants had at least primary control
of particulate emissions, and 73 percent were reported to have "best"
primary control; only 11 percent had "best" primary plus secondary
control (EPA79). It is presumed that "best" primary control consists
of the best available hooding, plus a fluidized-bed scrubber since this
unit can achieve the highest reported control efficiencies (97-99
percent removal). Based on this information, the model plant is
equipped with a fluidized-bed scrubber for primary control. The plant
has no secondary control equipment. As for the anode bake plant, a
spray scrubber constitutes the particulate control system.
Radionuclide emissions for the reference plant were based on
actual measurements of radionuclide concentrations in the particulate
emissions from an existing plant. The measurement data were scaled to
take into account the differences in capacity and particulate emission
rate between the actual and the reference plant. The resulting
releases are listed in Table 7.1-5.
7.1.6 Health Impact Assessment of Reference Plant
The estimated annual radiation doses from radionuclide emissions
from the reference aluminum reduction plant are listed in Table 7.1-6.
These estimates are for a rural site with a regional population of
2.7E+5 (rural southeastern site from TRI81).
Table 7.1-7 presents estimates of the maximum individual lifetime
risk and number of fatal cancers per year of operation from these doses.
7.1.7 Health Impact Assessment of Surveyed Plants
EPA has recently carried out radionuclide measurements at both an
alumina plant and an aluminum reduction plant. Reports on these
studies (EPA82) were completed subsequent to the analysis of the
reference plant assessed in the previous sections. These reports
7.1-6
-------
Table 7.1-4. Reference aluminum reduction plant (TRI81)
Parameter
Value
Capacity
Capacity factor
Type of equipment
136,000 MT/y aluminum
0.94
Center-worked prebake cells
Stack Parameters:
Main stack
Height
Diameter
Exit gas velocity
Exit gas temperature
Roof monitor
Height
Diameter
Exit gas velocity
Exit gas temperature
Anode bake plant
Height
Diameter
Exit gas velocity
Exit gas temperature
30 m
1.8 m
104.7 m/s
93° C
10 m
1.2 m
0.01 m/s
37° c
30 m
1.8 m
4.5 m/s
96<> C
Table 7.1-5. Radionuclide emissions from the
reference aluminum reduction plant (TRI81)
Emission (Ci/y)
Radionuclide
Main stack
Roof monitor Anode bake plant
Uranium-238
Uranium-234
Thorium-230
Radium-226
Lead-210
Polonium-210
Thorium-232
Radium-228
8.6E-3
8.6E-3
1.8S-2
2.9E-3
6.9E-2
6.9E-2
1.4E-2
1.4E-2
8.1E-9
8.1E-9
3.8E-8
7.4E-9
2.0E-7
2.0E-7
2.9E-8
2.9E-8
8.0E-5
8.0E-5
4.0E-5
6.0E-5
2.0E-4
2.0E-4
3.2E-5
3.2E-5
7.1-7
-------
provide information on the radionuclide concentrations of process
samples and annual radionuclide emission rates to air. The results of
these measurement studies and an assessment of the health risks from
these emissions for a generic site are presented in this section.
Alumina Plant
The alumina plant studied uses a modified "American Bayer" process.
The uranium-238 and thorium-232 concentration measured in the process
samples are listed in Table 7.1-8. The estimated annual radionuclide
emissions from this plant are listed in Table 7.1-9.
The bauxite ore was elevated in both uranium-238 and thorium-232
with concentrations of 6.8 and 5.5 pCi/g. The low radioactivity of
alumina is reflected in the low radionuclide emissions from the alumina
kilns. Particulate emissions of radionuclides from the red mud sinter
kiln were below measurable concentrations except for lead-210 and
polonium-210. The high temperatures of the kilns caused a large
fraction of these radionuclides to be volatilized. Emissions of the
two nuclides were essentially equal with 7.8 mCi/y for lead-210 and 9.3
mCi/y for polonium-210. The estimated radiation doses and health risks
from these emissions are listed in Tables 7.1-10 through 7.1-12. These
estimates are for a rural site with a regional population of 6E+5.
Table 7.1-6. Radiation dose rates from radionuclide
emissions from the reference aluminum reduction plant (TRI81)
_. Maximum individual Regional population
Organ , , . , / \
(mrem/y) (person-rem/y)
Lung
Red marrow
Endosteal
Muscle
Liver
2.5
4.5E-1
7.1
2.2E-2
5.3E-2
4.8
1.3
1.8E+1
2.9E-1
1.2
Table 7.1-7. Fatal cancer risks due to radionuclide emissions
from the reference aluminum reduction plant (TRI81)
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Aluminum
reduction plant 9E-6 3E-4
7.1-8
-------
Table 7.1-8.
Radionuclide concentrations in surveyed alumina
plant process samples (EPA82)
Sample
Concentration (pCi/g)
Uranium-238
Thorium-232
Bauxite ore
Alumina kiln feed
Alumina product
Red mud
Brown mud
6.8
0.05
0.28
7.5
5.5
5.5
0.05
0.2
5.0
12.5
Table 7.1-9.
Radionuclide emissions from the surveyed alumina plant
(EPA82)
Radionuclide
Emissions (Ci/y)
Alumina kilns
Red mud kilns
Uranium-238
Uranium-234
Radium-226
Radon-222
Lead-210
Polonium-210
6.8E-5
6.8E-5
5.5E-5
2.0
7.8E-3
9.3E-3
Table 7.1-10. Radiation dose rates from radioactive particulate
emissions from the surveyed alumina plant(a/
Organ
Maximum individual
(mrem/y)
Regional population
(person-rem/y)
Lung
Red marrow
Endosteal
Breast
Liver
Weighted sum
2.5
2.7E-1
1.3
1.6E-1
8.9E-1
9.8E-1
1.9
1.2E-1
4 . 3E-1
8.7E-2
3.8E-1
6.7E-1
(a)fiased on a 10-meter stack height.
7.1-9
-------
Table 7.1-11. Annual radon decay product exposures from radon-222
emissions from the surveyed alumina plant
0 Maximum individual Regional population
Source (WL-y) (person-WL-y)
Stack 4.6E-6 7.3E-3
Table 7.1-12. Fatal cancer risks from radionuclide
emissions from the surveyed alumina plant
s Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
Particulates 1E-5 1E-A
Radon-222 8E-6 2E-4
Total 2E-5 3E-4
Aluminum Reduction Plant
The aluminum reduction plant studied uses the "Hall" reduction
process. The uranium-238 and thorium-232 concentrations measured in
the process samples are listed in Table 7.1-3. The estimated annual
radionuclide emissions from the plant are listed in Table 7.1-13. Only
lead-210 and polonium-210 were present in measurable quantities in
plant emissions. The high temperatures involved in the process cause
these radionuclides to be volatilized. The estimated annual radiation
doses from these emissions are presented in Tables 7.1-14. These
estimates are for a rural site with a regional population of 6E+5. The
maximum individual lifetime risk and number of fatal cancers per year
of operation from these doses are shown in Table 7.1-15.
Table 7.1-13. Radionuclide emissions from the surveyed
aluminum reduction plant (EPA82)
Radionuclide
Lead-210 3.2E-2
Polonium-210 2.7E-2
7.1-10
-------
Table 7.1-14. Radiation dose rates from radionuclide
emissions from the surveyed aluminum reduction plant(a)
Organ Maximum individual Regional population
(mrem/y) (person-rem/y)
Lung
Red marrow
Endosteal
Breast
Liver
Kidney
Spleen
Weighted sum
6.0E-1
1.7E-1
1.0
9.2E-2
6 . 1E-1
1.6
2.4
3.3E-1
4.6
3.6E-1
1.6
2.4E-1
1.2
4.2
6.8
1.7
(a)Based on a 36-meter stack height.
Table 7.1-15. Fatal cancer risks from radionuclide
emissions from the surveyed aluminum reduction plant
Lifetime risk Regional population
ource to max-[mum individual (Fatal cancers/y of operation)
Particulates 3E-6 3E-4
7.1-11
-------
REFERENCES
DOI80 U.S. Department of the Interior, 1980, Mineral Commodity
Summaries, Bureau of Mines, January 1980.
EPA76 Environmental Protection Agency, Compilation of Air Pollution
Emission Factors, Second Ed., Part B, AP-42, Feburary 1976.
EPA77 Environmental Protection Agency, Technical Guidance for
Control of Industrial Process Fugitive Particulate Emissions,
EPA-450/3-77-010, March 1977.
EPA79 Environmental Protection Agency, Primary Aluminum: Draft
Guidelines for Control of Flouride Emissions from Existing
Primary Aluminum Plants, EPA-450/2-78-049, February 1979.
EPA82 Environmental Protection Agency, Emissions of Naturally
Occurring Radioactivity from Aluminum and Copper Facilities,
EPA 520/6-82-018, Las Vegas, Nevada, November 1982.
St78 Stamper J.W. and Kurtz H.F., Mineral Commodity
Profile-Aluminum, U.S. Department of the Interior, Bureau of
Mines, Washington, D.C.
TRI81 Teknekron Research, Inc., Draft, Partial and Supplemental
Background Information Document—Primary Pyrometallurgical
Extraction Process, Report to Environmental Agency under
Contract No. 68-01-5142, USEPA Docket Number A-79-11, May 1981.
7.1-12
-------
7.2 Copper Industry
7.2.1 General Description
Copper ores are milled to produce a concentrate containing copper,
sulfur, iron, and some insoluble material (primarily silica and
aluminum). This concentrate is the basic feed to the copper smelter
that eventually produces the refined copper product. Copper mills and
smelters are located near copper mines. Copper concentrates and
precipitates are generally smelted by melting the charge and suitable
fluxes in a reverberatory furnace. Prior to smelting, part or all of
the concentrates may receive a partial roast to eliminate some of the
sulfur and other impurities.
The 15 operating primary copper smelters in the United States and
their capacities are listed in Table 7.2-1. Total production of
primary copper in 1978 was 1.5 million metric tons (Sc79).
All primary copper smelters are located in rural areas with low
population densities. Ninety percent of U.S. copper smelter capacity
is located in the arid and semi-arid climates of Arizona, Montana,
Nevada, New Mexico, Texas, and Utah. The other 10 percent are in
Washington, Michigan, and Tennessee, areas of moderate-to-high
precipitation. The sites tend to be quite large and generally contain
associated mining and milling operations.
Most companies perform all production processes from mining
through refining. Seven of the eight companies that own smelters also
operate mines and own refineries; Cities Services, which owns the
smallest of the smelters, is the only exception (Sc79).
7.2.2 Process Description
The three major steps in the smelting of copper are roasting,
smelting, and converting. All of these processes result in releases of
sulfur dioxide and particulate matter in process off-gas. Each step in
the smelting process is described below.
Roasting
Roasting is the first step in the process of copper smelting. In
the roaster, copper ore concentrates are heated to a high temperature
(550° C) in an oxidizing atmosphere which partially drives off some
of the sulfur as sulfur dioxide (in addition to producing particulate
emissions). Seven of the fifteen domestic copper smelters have
roasters; four plants feed ore concentrates to a rotary dryer to reduce
moisture before smelting; and three feed concentrates directly to the
furnace with no pretreatment.
7.2-1
-------
Table 7.2-1. Primary Copper Smelters in the United States, 1978
(TRI81)
Plant
location
Arizona
Hayden
Miami
Hayden
San Manuel
Morenci
Douglas
Ajo
Michigan
White Pine
New Mexico
Hurley
New Mexico
Hidalgo
New York
McGill
Tennessee
Copper Hill
Texas
El Paso
Utah
Garfield
Washing ton
Tacoma
Total
(a>Rebuilt as
Company
ASARCO, Inc.
Inspiration Consolidated
Kennecott Copper Corp.
Magma Copper Company
Phelps Dodge Corporation
Phelps Dodge Corporation
Phelps Dodge Corporation
Copper Range Company
Kennecott Copper Corp.
Phelps Dodge Corporation
Kennecott Copper Corp.
Cities Services Company
ASARCO, Inc.
Kennecott Copper Corp.
Corporation
ASARCO, Inc.
of 1979.
Capacity
(1000 MT)
163
136
73
181
161
115
63
82
73
127
45
20
104
254
91
1688
First year
of operation
1890
195l(a)
1958
1950
1942
1910
1950
1905
1939
1976
1907
1845
1905
1907
1890
7.2-2
-------
Smelting
All domestic copper smelters use smelting furnaces to melt and react
copper concentrate and/or calcine in the presence of silica and limestone
flux to form two immiscible liquid layers, one being the slag or waste
layer containing most of the iron and silica compounds and the other
containing copper and iron sulfide and other metals, referred to a matte
copper. Smelting is conducted in either reverberatory or electric
furnaces. Reverberatory furnaces are refractory-lined, box-shaped
structures heated by either natural gas, oil, or coal. Reverberatory
smelting furnaces are more common than electric furnaces. Currently, 2
out of 15 smelters use electric furnaces to smelt copper. Electric
furnaces have basically the same construction as reverberatory furnaces.
Converting
The converter processes matte copper from the reverberatory furnace
by removing iron compounds and converting copper at high temperatures
(550 to 800° C). The resulting blister copper is further purified by
processing in a refining furnace and by electrolytic refining .
7.2.3 Control Technology
Of the 15 primary copper smelters currently operating, 11 use
reverberatory furnaces and 7 have roasters. Of these seven, four use
multi-hearth roasters while the other three use fluid-bed roasters. The
actual smelting process used by those plants with reverberatory furnaces
does not differ from facility to facility. Acid gas cleanup plants have
been installed on all but three currently operating smelters to treat
converter off-gases. A cyclone, a water spray chamber, and an
electrostatic precipitator (ESP) are used to clean these gases prior to
their entering the S02 plant. Off-gases from the reverberatory furnace
are controlled via an ESP in virtually all of the operating plants.
Three of the four multi-hearth roasters currently operating treat their
roaster off-gases by using ESPs.
7.2.4 Radionuclide Emission Measurements
Particulate material emitted from a copper smelter contains
radionuclides in concentrations (pCi/g) similar to or greater than the
ore concentrates. Because of the high temperatures of the roasting and
smelting, some radionuclides (particularly lead-210 and polonium-210) may
be volatilized and released in greater quantities than the other
radionuclides in the ore concentrates.
Very little information is available on radionuclide emissions from
copper smelters. EPA has recently surveyed two copper smelters. The
preliminary data from these studies were used in estimating radionuclide
emissions from the reference copper smelter.
7.2-3
-------
7.2.5 Reference Facility
Table 7.2-2 describes the parameters of a reference copper smelter
which were used to estimate the radioactive emissions to the atmosphere
and the resulting health impacts. The capacity of the plant is 97,000
MT/y of copper, the average size of all existing plants without
roasters. The capacity factor chosen for this plant is 0.75. Main stack
heights for facilities without roasters range from 61 to 228 meters. The
control equipment applied to the reference facility was chosen to
represent typical equipment on actual copper smelters.
Total annual emissions of radionuclides from the reference copper
smelter are given in Table 7.2-3. These values were derived from
preliminary data on radionuclide releases from an existing plant.
Reported release rates were adjusted to account for differences between
the actual and reference facility in annual particulate emissions and
total capacity.
Table 7.2-2. Reference copper smelter (TRI81)
Parameter
Value
Capacity
Capacity factor
Type of equipment used
Stack Parameters
Main stack:
Height
Diameter
Exhaust gas velocity
Exhaust gas temperature
Acid plant
Height
Diameter
Exhaust gas velocity
Exhaust gas temperature
Particulate Emission Rate
Main stack:
Acid plant
97,000 MT/y
0.75
Reverberatory furnace
150 m
3.3 m
18 m/s
93° C
30.4 m
1.8 m
16.5 m/s
79° C
247 kg/h
11 kg/h
7.2-4
-------
Table 7.2-3. Radionuclide emissions from the reference copper smelter
(TRI81)
Radionuclide
Uranium-238
Uranium-234
Thorium-230
Radium-226
Lead-210
Polonium-210
Thorium-232
Radium-228
Thorium-228
Emissions
(Ci/y)
l.OE-2
l.OE-2
1.9E-2
1.7E-3
2.0E-1
2.0E-1
1 . 3E-2
1.3E-2
1.3E-2
7.2.6 Health Impact Assessment of the Reference Copper Smelter
The estimated radiation doses from radionuclide emissions from the
reference copper smelter are listed in Table 7.2-4. 'These estimates
are for a low population density southwestern site with a regional
population of 3.6E+4.
Table 7.2-5 presents estimates of the maximum individual lifetime
risk and number of fatal cancers per year of operation resulting from
these doses.
Table 7.2-4. Radiation dose rates from radionuclide particulate
emissions from the reference copper smelter
Maximum individual Regional population
r8an (mrem/y) (person-rem/y)
Lung 7.7E-2 9.5E-1
Red marrow 1.2E-1 2.6E-1
Endosteal 3.4E-1 2.8
Breast 9.7E-2 7.5E-2
Liver 1.6E-1 3.2E-1
Weighted sum 1.1E-1 4.3E-1
7.2-5
-------
Table 7.2-5. Fatal cancer risks from radionuclide
emissions from the reference copper smelter
Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
Particulates 2E-6 6E-5
7.2.7 Health Impact Assessment of Surveyed Plants
EPA has recently carried out radionuclide measurement studies at
both an underground copper mine and mill and an open pit copper mine
and mill. A report on these studies was completed in November 1982
(EPA82). This report provides information on the radionuclide
concentrations of process samples and annual radionuclide emission
rates to air. Results from these measurement studies and an assessment
of the health risks from the measured emissions for a generic site are
presented in this section.
Table 7.2-6 lists the uranium-238 and thorium-232 concentrations
in the process samples from both the underground mine and mill and the
open pit mine and mill.
Table 7.2-6. Radionuclide concentrations in surveyed copper mine
and mill process samples (EPA82)
Type
of
Sample
Ore
Concentrate
Underground mine and mill
Uranium-238 Thorium-232
(pCi/g) (pCi/g)
0.79 0.62
0.65 0.07
Open pit mine and mill
Uranium-238 Thorium-232
(PCi/g) (pCi/g)
2.2 3.1
1.4 1.1
Underground Mine and Mill
The underground mine selected for this survey was an underground
copper-iron-zinc-sulfide mine. The Mine Safety and Health
Administration (MSHA) had reported average radon decay product
measurements of 0.087 WL with a maximum of 0.21 WL. The ore runs less
7.2-6
-------
than 1 percent copper, less than 1 percent zinc, 20 percent iron, and
25 percent sulfur. A mill and flotation plant produces concentrates of
the three sulfides. The uranium decay chain nuclides were at or
slightly above the average concentrations found in natural rock. The
ore concentration averaged about 1 pCi/g (see Table 7.2-6).
The radionuclide emissions from the underground mine and mill
consisted primarily of radon-222 released from the mine exhausts (Table
7.2-7). The radon decay product exposures from these emissions are
listed in Table 7.2-8. These estimates are for a low population
density southwestern site with a regional population of 3.6E+4. The
maximum individual lifetime risk and number of fatal cancers per year
of operation from these exposures are listed in Table 7.2-9.
Table 1.1-1. Radionuclide emissions from the surveyed
underground copper mine (EPA82)
Radionuclide Emissions
(Ci/y)
Radon-222 6.5
Table 7.2-8. Annual radon decay product exposures from radon-222
emissions from the surveyed copper mine'3)
„ Maximum individual Regional population
rce (WL-y) (person-WL-y)
Mine vent 6.5E-5 1.3E-3
(a)Based on a ground level release.
Table 7.2-9. Fatal cancer risks from radionuclide
emissions from the surveyed copper mine
Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
Radon-222 1E-4 3E-5
7.2-7
-------
Open Pit Mine and Mill
Radionuclide emissions from the copper mill associated with the
surveyed open pit mine are listed in Table 7.2-10. Particulate
emissions containing radionuclides are released primarily from the
crushing operations and truck hoppers. Radon-222 flux measurements of
the open pit mine surfaces did not identify any quantifiable radon-222
emissions (above natural levels) from the pit. The estimated radiation
dose rates and radon-222 decay product exposures from the radionuclide
emissions from the surveyed copper mill are listed in Table 7.2-11 and
7.2-12. These estimates are for a low population density southwestern
site with a regional population of 3.6E+4. The maximum individual
lifetime risk and number of fatal cancers per year of operation from
these exposures are shown in Table 7.2-13.
Table 7.2-10. Radionuclide emissions from surveyed copper mill
(EPA82)
„ , . , . , Emissions
Radionuclide
Uranium-238 3.1E-4
Uraniium-234 3.8E-4
Radium-226 1.8E-4
Radon-222 1.9
Lead-210 8.9E-4
Table 7.2-11. Radiation dose rates from radioactive particulate
emissions from the surveyed copper mill(a)
0 Maximum individual Regional population
(mrem/y) (person-rem/y)
Lung 3.3 6.1E-2
Red Marrow 3.2E-2 1.9E-3
Endosteal 1.3E-1 1.6E-2
Breast 2.2E-2 7.3E-4
Liver 3.8E-2 3.2E-3
Weighted sum 9.7E-1 1.9E-2
(a)Based on a 10-meter stack height.
7.2-8
-------
Table 7.2-12. Annual radon decay product exposures from radon-222
emissions from the surveyed copper mill
Source Maximum individual Regional population
(WL-y) (person-WL-y)
Stack 8.3E-6 3.7E-4
Table 7.2-13. Fatal cancer risks from radionuclide
emissions from the surveyed copper mill
„ Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
Particulates 1E-5 3E-6
Radon-222 IE-5 9E-6
Total 2E-5 1E-5
7.2-9
-------
REFERENCES
EPA82 Environmental Protection Agency, Emissions of Naturally
Occurring Radioactivity from Aluminum and Copper Facilities,
EPA 520/6-82-018, Las Vegas, Nevada, November 1982.
Sc79 Schroeder H. J., Mineral Commodity Profiles—Copper, U.S.
Department of the Interior, Bureau of Mines, Washington, D.C.,
1979.
TRI81 Teknekron Research, Inc., Draft, Partial and Supplemental
Background Information Document—Primary Pyrometallurgical
Extraction Process, Report to the Environmental Protection
Agency under Contract No. 68-01-5142, USEPA Docket Number
A-79-11, May 1981.
7.2-10
-------
7.3 Zinc Industry
7.3.1 General Description
Zinc is usually found in nature as a sulfide ore called
sphalerite. The ores, which usually contain impurities of lead,
cadmium, and traces of other elements, are processed at the mine to
form concentrates typically containing 62 percent zinc and 32 percent
sulfur. These concentrates are processed at the smelter to recover
zinc metal.
The five operating primary zinc production facilities in the
United States and their capacities are listed in Table 7.3-1. Total
production capacity for primary zinc in 1980 was 401,000 metric tons.
The domestic demand for zinc is expected to grow at a rate of about
2 percent per year through 1985 (Ca78).
In the past 10 years, li.S. demand for zinc metal has grown slowly,
out U.S. smelting capacity has declined by over 50 percent. Plants
closed because they were obsolete, could not meet environmental
standards, or could not obtain sufficient concentrate feed.
Consequently, the metal has replaced concentrate as the major form of
import. This situation is expected to continue.
7.3.2 Process Description
A zinc smelter produces 99.99+ percent zinc from the approximately
62 percent zinc concentrate feed produced by the mill. The zinc
concentrates are roasted at approximately 600° C to convert sulfur to
sulfur dioxide and to produce an impure zinc oxide or calcine. The
calcine is transferred to tanks, leached with dilute sulfuric acid, and
treated with a small amount of zinc oxide dust to remove impurities,
such as lead, gold, and silver.
The leaching step varies somewhat from plant to plant, but the
basic process of selective precipitation of the impurities from the
leach solution remains the same. This solution is purified and piped
to electrolytic cells, where the zinc is electro-deposited on aluminum
cathodes. Domestic zinc smelters use electrolytic reduction to reduce
the quantity of sulfur and particulate emissions.
The cathodes are lifted from the tanks at intervals and stripped
of the zinc, which is melted in a furnace and cast into slabs. Elec-
trolysis of the solution regenerates sulfuric acid which is used in
succeeding cycles of leaching.
7.3.4 Control Technology
Ore concentrates are heated in roasters to temperatures ranging
from 5000 c to 700° C to remove most of the sulfur in the sulfide
7.3-1
-------
Table 7.3-1. Location and size of primary zinc production plants(a)
(TRI81)
T . - First year Capacity
Location Company , ' . ,„,, , f .,.,1^
* of operation (Thousands of MT)
Idaho Bunker Hill 1928 95
Kellogg(b)
Illinois
Sauget
Oklahoma
Bartlesville
Tennessee
Clarksville
Texas
Corpus Christi
Total
AMAX
National
Jersey Miniere
Asarco
Rebuilt in 1970 's
1976
1978
1942
76
51
81
98
401
(a)All plants use the electrolytic process.
is now shut down.
ore and form calcine. Roaster off-gases containing sulfur dioxide are
treated in single or double contact acid plants. The off-gas also
contains significant amounts of calcine, which is recovered in waste
heat boilers, cyclones, and ESP's and then recycled. In addition, most
acid plants have wet scrubbers, wet ESP's, and demisting towers before
the plant catalyst to remove residual particulate matter which could
foul the catalyst bed.
The electrolytic (or hydrometallurgical) zinc smelting process is
a minor source of particulate emissions, and is not serviced by a
particulate control device (TRI81).
7.3.5 Radionuclide Emissions
Particulate material emitted from a zinc smelter contains
radionuclides in concentrations similar to or greater than the
concentrations in the materials processed. Because of the high
temperatures to which the concentrates are heated, some of the
radionuclides (particularly lead-210 and polonium-210) may be
volatilized and released in greater quantities than the other
radionuclides in the ore concentrates. Although EPA has recently
7.3-2
-------
measured the radionuclide emissions from a zinc smelter, the results of
these measurements are not yet available. Therefore, the radionuclide
emissions for the reference plant are based on estimates of particulate
emissions and enrichment factors for radionuclides which may be
volatilized during processing of the ore concentrates (TRI81).
Control systems which control particulates will also control
radionuclide emissions. Particulate material in the roaster off-gases
is generally removed via a combination of cyclones, scrubbers, and
electrostatic precipitators. The off-gases are then treated in an acid
plant and released. As far as can be determined, off-gases from the
electrolytic process itself are vented directly to the atmosphere.
7.3.6 Reference Facility
Table 7.3-2 describes the parameters of a reference zinc smelter
which were used to estimate the radioactive emissions to the atmosphere
and the resulting health impacts.
The reference zinc smelter has a total production capacity of
about 88,000 MT/y, typical of the industry. The plant produces zinc by
electrolytic reduction and operates at an annual capacity factor of
0.80, the 1976 industry-wide average (DOI76). The flow rate was
derived by adjusting available data for differences in capacity and
capacity factor. The stack height was estimated from available data.
This value, in turn, was used to estimate the stack diameter.
Table 7.3-2. Reference zinc plant (TRI81)
Parameter Value
Process Electrolytic reduction
Capacity 88 E+3 MT/yr zinc
Capacity factor 0.8
Radionuclide concentration
of input ore
Uraniuim-238 0.43 pCi/g
Thorium-232 0.35 pCi/g
Stack Parameters
Number 1
Height 61 meters
Diameter 2.4 meters
Exhaust gas velocity 2.7 m/s
Exhaust gas temperature 93° C
7.3-3
-------
Roaster off-gases are treated for dust removal by a cyclone in
series with an electrostatic precipitator. The cleaned gases are then
passed through a sulfur dioxide (SC>2) plant. Off-gases from the
electrolytic reduction step are vented directly to the atmosphere.
The total annual radionuclide emissions for the reference zinc
smelter are listed in Table 7.3-3.
Table 7.3-3. Radionuclide emissions from the reference zinc smelter
(TRI81)
Radionuclide
Uranium-238
Uranium-234
Thorium-230
Radium-226
Lead-210
Polonium-210
Thorium-232
Radium-228
Emissions
(Ci/y)
2.6E-4
2.6E-4
1 . 3E-4
1.9E-4
6.5E-4
6.5E-5
7.9E-5
7.9E-5
7.3.6. Health Impact Assessment of Reference Zinc Smelter
The estimated annual radiation doses from radionuclide emissions
from the reference zinc smelter are listed in Table 7.3-4. These
estimates are for a rural site with a regional population of 6E+5. The
maximum individual lifetime risk and number of fatal cancers per year
of operation of the reference plant are shown in Table 7.3-5-
Table 7.3-4. Radiation dose rates from radionuclide
emissions from the reference zinc smelter
Q Maximum individual Regional population
(mrem/y) (person-rem/y)
Lung
Red marrow
Endosteal
Breast
Liver
7.4E-3
1.1E-2
6.6E-2
6.1E-3
1.4E-2
7.4E-1
7.6E-2
8.0E-1
2.6E-2
4 . OE-2
Weighted sum 9.1E-3 2.6E-1
7.3-4
-------
Table 7.3-5. Fatal cancer risks from radionuclide
emissions from the reference zinc smelter
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Particulates 1E-7 4E-5
7.3.7 Health Impact Assessment of Surveyed Plants
EPA has recently carried out measurements at a zinc mine and
mill. A report on these measurements was issued in November 1982
(EPA82). This report provided information on the radionuclide
concentrations in process samples and on annual radionuclide emissions
to air. The zinc mine and mill surveyed were chosen because of their
high production rates and the high WL measurements in the mine reported
by the Mine Safety and Health Administration. Results of these
measurements and an assessment of the health risks from these emissions
for a generic site are presented in this section.
Table 7.3-6 lists the uranium-238 and thorium-232 concentrations
in process samples from the mine and mill.
Table 7.2-6. Radionuclide concentrations in surveyed zinc mine
and mill process samples (EPA82)
Type of Concentration (pCi/g)
Sample Uranium-238 Thorium-232
Ore 0.18 0.08
Concentrate 0.16 0.04
Zinc Mine
Measurements of the radionuclide emissions from the zinc mine
showed radon-222 to be the principal radionuclide emitted to air.
Other radionuclides were emitted in much smaller quantities. Table
7.3-7 lists the radionuclide emissions from the surveyed zinc mine.
The estimated annual radiation doses and radon decay product exposures
from these emissions are presented in Tables 7.3-8 and 7.3-9. These
7.3-5
-------
estimates are for a rural site with a regional population of 6E+5. The
maximum individual lifetime risks and number of fatal cancers per year
of operation of the zinc mine are shown in Table 7.3-10.
Table 7.3-7. Radionuclide emissions from the surveyed
zinc mine (EPA82)
Radionuclide
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-222
Lead-210
Polonium-210
Thorium-232
Emissions
(Ci/y)
1.2E-7
1.2E-7
l.OE-7
7.0E-8
2 . 3E+2
3.2E-7
1.5E-7
5.0E-8
Table 7.3-8. Radiation dose rates from radioactive particulate
emissions from the surveyed zinc mine(a)
Q Maximum individual Regional population
(mrem/y) (person-rem/y)
Lung
Red marrow
Endosteal
Breast
Liver
2.4E-3
7.2E-4
4.6E-3
4.1E-4
4.0E-4
9. IE -4
3.9E-4
1 . 6E-3
2.8E-4
2.5E-4
Weighted sum 1.1E-3 4.8E-4
(a)Based on ground level release.
Table 7.3-9. Annual radon decay product exposures from radon-222
emissions from the surveyed zinc mine
Source Maximum individual Regional population
(WL-y) (person-WL-y)
Mine vent 1.2E-3 8.6E-1
7.3-6
-------
Table 7.3-10. Fatal cancer risks from radlonuclide
emissions from the surveyed zinc mine
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Particulates 1E-8 1E-7
Radon-222 2E-3 2E-2
Total 2E-3 2E-2
Zinc Mill
Table 7.3-11 lists the radionuclide emissions from the surveyed
zinc mill. The small emissions reflect the low radionuclide
concentration of the zinc ore and the low particulate emissions from
the plant. The estimated annual radiation doses and radon decay
product exposures from these emissions are presented in Table 7.3-12
and 7.3-13. These estimates are for a rural site with a regional
population of 6E+5. The maximum individual lifetime risk and number of
fatal cancers per year of operation from these exposures are shown in
Table 7.3-14.
Table 7.3-11. Radionuclide emissions from the surveyed zinc mill
(EPA82)
Radionuclide Emissions
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-220
Lead-210
Polonium-210
Thnrliim-232
1.7E-6
1.7E-6
1.4E-6
7.5E-7
1.0
1.3E-6
2.0E-6
5.1E-7
7.3-7
-------
Table 7.3-12. Radiation dose rates from radioactive particulate
emissions from the surveyed zinc mill(a)
Organ
Lung
Red marrow
Endosteal
Breast
Liver
Weighted sum
Maximum individual
(mrem/y)
5.8E-3
7.5E-4
1.1E-2
8.2E-5
1.6E-4
2.0E-3
Regional population
(person-rem/y)
l.OE-2
1.3E-3
1.8E-2
1.5E-4
2.0E-4
3.6E-3
(a)Based on a 15-meter stack height.
Table 7.3-13. Annual radon decay product exposures from radon-222
emissions from the surveyed zinc mill
c Maximum individual Regional population
(WL-y) (person-WL-y)
Zinc mill 9.9E-7 3.5E-3
Table 7.3-14. Fatal cancer risks from radionuclide
emissions from the surveyed zinc mill
Source Lifetime risk Regional population
to maximum individual (Fatal cancers/y of operation)
Particulates 2E-8 5E-7
Radon-222 2E-6 8E-5
Total 2E-6 8E-5
7.3-8
-------
REFERENCES
Ca78 Cammarota V. A., Jr., Mineral Commodity Profiles-Zinc,
MCP-12, U.S. Department of the Interior, Bureau of Mines, May
1978.
DOI76 Department of Interior, Preprint from the 1976 Bureau of
Mines Minerals Yearbook: Zinc, Washington, D.C., 1976.
EPA82 Environmental Protection Agency, Emissions of Naturally
Occurring Radioactivity: Underground Zinc Mine and Mill,
EPA 520/6-82-020, Las Vegas, Nevada, November 1982.
TRI81 Teknekron Research, Inc., Draft, Partial and Supplemental
Background Information Document—Primary Pyrometallurgical
Extraction Process, Report to Environmental Agency under
Contract No. 68-01-5142, USEPA Dosket Number A-79-11, May
1981.
7.3-9
-------
7.4 Lead Industry
7.4.1 General Description
Galena (PbS), frequently containing cerussite (PbC03) and
anglesite (PbSC>4), is the principal lead-bearing ore found in
nature. A sulfide ore, galena contains small amounts of copper, iron,
zinc, and other trace elements (EPA75). In the smelting process, lead
bullion (95-99 percent lead metal) is separated from ore concentrates
(45-80 percent lead).
Table 7.4-1 lists the location and size of the primary lead
smelters. Three facilities have integrated smelter/refinery complexes
and two facilities (ASARCO's El Paso and East Helena smelters) ship
their drossed lead bullion to the company's Omaha refinery for final
processing. Refinery operations, including those co-located with
smelters, are not considered part of the primary lead source category.
Three of the smelters are located in southeastern Missouri and
process only ores from the Missouri lead belt. The smelters located in
Texas and Montana are custom smelters, designed to handle larger
variations in ore composition than the Missouri smelters. Both
domestic and foreign ores are smelted at the western plants.
The design capacities of the primary lead smelters, expressed as annual
lead metal output, range from 82,000 to 204,000 tons. Total production
from primary smelters in 1979 was 594,000 tons (DOC80).
7.4.2 Process Description
Lead smelting involves three distinct processes: sintering, to
convert the ore from a sulfide to an oxide or sulfate form and prepare
the feed materials for furnacing; furnacing, to reduce the oxide feed
to lead metal; and dressing, to reduce the copper content of the lead
bullion from the furnace. After dressing, additional refining steps,
as dictated by the specific impurities present and the intended end-use
of the product, are performed to yield the purified lead metal.
7.4.3 Control Technology
Off-gases from the sintering machine and the blast furnace are the
most significant sources of particulate emissions from the lead
smelting process; together these two sources account for more than 95
percent of particulate emissions.
7.4-1
-------
Table 7.4-1. Location and size of primary lead production plants (TRI81)
Capacity
T . _ First year (Thousands of
Location Company , * .. . , p, ^
* J of operation tons of Pb)
Idaho
Kellogg(a> Bunker Hill 1917 117
Missouri
Boss Amax-Homestake 1968 127
Glover ASARCO 1968 100
Herculanium St. Joe Minerals 1892 204
rebuilt 1970's
Montana
East Helena ASARCO 1888 82
Texas
El Paso ASARCO 1887 82
(a) Now shut down.
Sintering Machines
Particle size distribution of particulate matter entrained in
off-gas from sintering machines indicated that the majority of
particles are less than 10 microns in diameter. This relatively small
particle size precludes the use of mechanical collectors or wet
scrubbing systems, which decrease in efficiency substantially with
decreasing size of the particle collected. Consequently, five of the
six existing lead sintering machines use fabric filters for particulate
emission control; the sixth employs an ESP (IERL79). The final control
devices, in many cases, are preceded by ballon flues or settling
chambers for gravitational collection of more massive particles before
off-gases enter the ESP or fabric filter.
Sinter off-gas is typically fed to an acid plant for recovering of
sulfur dioxide after particulate cleaning, as described above.
Efficient operation of the acid plant requires gases containing 5
percent or more S02. The circuit of gases through the sinter machine
may be quite complex with weak (in S02> gases being recirculated
through an upstream section of the machine to enrich the S02 content
before going to the acid plant.
7.4-2
-------
Blast Furnaces
The majority of particles in the lead blast furnace off-gas are
smaller than 10 microns in diameter. Consequently, all blast furnace
systems currently in operation are serviced by baghouses. The
particulate collection efficiencies of baghouses treating lead blast
furnace off-gas is roughly 99 percent.
7.4.4 Radionuclide emissions
Particulate material emitted from a lead smelter contains
radionuclides in concentrations similar to or greater than the
concentrations in the materials processed. Since enrichment takes
place when nuclides volatilize during the high-temperature phase of
production, the concentration of some radionuclides will be higher in
the particulates than in the original ore. Although EPA has recently
measured the radionuclide emissions at a lead smelter, results of these
measurements are not yet available. Therefore, the releases for the
reference lead smelter have been calculated by assuming that the
radionuclide content in the particulate released is the same as that in
the input ore and applying the appropriate enrichment factors for
volatile radionuclides. Multiplying the concentrations of
radionuclides in the ore by the total annual particulate release then
yields the total annual radionuclide release.
The particulate emission rates and enrichment factors used in
estimating the emissions from the reference plant were taken from TRI81.
7.4.5 Reference Facility
Table 7.4-2 describes the parameters of the reference facility
which were used to estimate the radioactive emissions to the atmosphere
and the resulting health impacts.
The reference lead smelter has a capacity of 107,000 MT lead per
year, typical of existing plants. The plant operates at a load factor
of 0.92 which was the industry-wide average for 1979 (DOC80). There
are two stacks at the plant—a main stack and an acid plant tail gas
stack. The height of the main stack was chosen by averaging the
heights of all stacks in the industry.
The particulate emissions and flow rate through the stack were
determined by taking weighted averages of values of the actual
facilities.
7.4-3
-------
Table 7.4-2. Reference lead smelter (TRI81)
Parameter Value
Capacity 1.07E+3 MT/yr lead
Capacity factor 0.92
Radionuclide concentration
of input ore:
Uranium-238 0.43 pCi/g
Thorium-232 0.35 pCi/g
Stack Parameters
Number 2
Main stack
Height 38 meters
Diameter 1.8 meters
Exit gas velocity 14.7 m/s
Exhaust gas temperature 149 °C
Acid plant stack
Height 30 meters
Diameter 1.8 meters
Exhaust gas velocity 1.7 m/s
Exhaust gas temperature 93 °C
Table 7.4-3. Radionuclide emissions from the
reference lead plant (TRI81)
Radionuclide Emissions(a)
(Ci/y)
Uranium-238 1.1E-4
Uranium-234 1.1E-4
Thorium-230 5.5E-5
Radium-226 8.2E-5
Lead-210 2.7E-4
Polonium-210 2.7E-4
Thorium-232 4.4E-5
Radium-228 4.4E-5
(a)Main stack only.
7.4-4
-------
7.4.6 Health Impact Assessment of Reference Smelter
The estimated radiation doses from radionuclide emissions from the
reference lead smelter are listed in Table 7.4-4. These estimates are
for a rural site with a regional population of 2.9E+5 (rural central
site from TRI81).
Table 7.4-5 presents estimates of the maximum individual lifetime
risk and number of fatal cancers per year of operation of the reference
smelter.
Table 7.4-4. Radiation dose rates from radionuclide
emissions from the reference lead smelter (TRI81)
Orean Maximum individual Regional population
(mrem/y) (person-rem/y)
Lung
Red marrow
Endosteal
Muscle
Liver
5.6E-2
4.7E-3
7. IE -2
3.8E-4
7.3E-4
9.8E-2
1.3E-2
1 . 7E-1
3.2E-3
1 . 1E-2
Table 7.4-5. Fatal cancer risks due to radionuclide emissions
from the reference lead smelter (TRI81)
Lifetime risk Regional population
ource to maximum individual (Fatal cancers/y of operation)
Lead smelter 2E-7 5E-6
7.4-5
-------
REFERENCES
DOC80 U.S. Department of Commerce, U.S. Industrial Outlook for 200
Industries with Projections for 1984, Washington, D.C., 1980.
EPA75 Environmental Protection Agency, Development for Interim Final
Effluent Limitations Guidelines and Proposed New Source
Performance Standards for the Lead Segment of the Nonferrous
Metals Manufacturing Point Source Category, EPA 440/ 1-75/032-9,
Washington, D.C., February 1975.
IERL79 Industrial Environmental Research Laboratory, Control of
Particulate Emissions in the Primary Nonferrous Metals
Industries, NTIS Report No. PB-80-151822, Cincinnati, Ohio,
December 1979.
TRI81 Teknekron Research, Inc., Draft, Partial and Supplemental
Background Information Document—Primary Pyrometallurgical
Extraction Process, Report to Environmental Agency under Contract
No. 68-01-5142, USEPA Dosket Number A-79-11, May 1981.
7.4-6
-------
APPENDIX A
ASSESSMENT METHODOLOGY
-------
APPENDIX A: ASSESSMENT METHODOLOGY
CONTENTS
Page
A.I Introduction A-5
A.2 Reference Facility A-5
A.3 Generic Sites A-5
A.4 Source Characterization A-6
A.5 Environmental Pathway Modeling Computer Programs A-6
A.6 Individual Assessment A-ll
A.7 Collective Assessment A-ll
A.8 AIRDOS-EPA Parameters and Input Data A-12
A.9 DARTAB—Dose and Risk Tables A-15
References A-23
TABLES
A-l Characteristics of the generic sites A-7
A-2 Sources of food for the maximum individual A-12
A-3 Some site parameters used with AIRDOS-EPA A-13
A-4 Cattle densities and vegetable crop distributions
for use with AIRDOS-EPA A-l6
A-5 Site independent parameters used for AIRDOS-EPA
generic site assessments A-18
A-6 Element dependent factors used in AIRDOS-EPA assessments A-20
A-7 Weighting factors used for weighted sum dose equivalent A-22
A-3
-------
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 reference 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 individual and collective doses and risks due to
air immersion, ground surface exposure, inhalation, and ingestion of
radionuclides,
Assumptions made at each step were intended to be realistic without
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 Reference Facility
For each source category, a reference 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
reference 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 were used for the assessment.
A.3 Generic Sites
Generic sites were characterized for the purpose of assessing
different source categories. These sites were chosen by
A-5
-------
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, seven generic sites (designated A,
B, C, D, E, F, and G) 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 several 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 and suburban locations,
respectively. 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). Site G (near
Pocatello, Idaho) is representative of elemental phosphorous
processing sites. 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. 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.
A.5 Environmental Pathway Modeling Computer Programs
AIRDOS-EPA (Mo79) was used to calculate the individual and
collective radionuclide concentrations for these assessments. Decay
product concentrations (in working level units) associated with
radon-222 were calculated on the assumption of a 70 percent
equilibrium, a value considered representative of indoor exposure
conditions (Ge78).
A-6
-------
Table A-l. Characteristics of the generic sites
Site A—New York
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
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
12.1° C
1000 m
9.23E+5 persons
1.71E+7 persons
1.72E+5 head
1.17E+5 head
3.77E+4 ha
Site B—Missouri
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
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-G
60/01-64/12
102 cm
11.50 c
600 m
1.34E+4 persons
2.49E+6 persons
3.80E+4 head
6.90E+5 head
1.64E+4 ha
A-7
-------
Table A-l. Characteristics of the generic sites—continued
Site C--Florida
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
Average Mixing Height:
Population: (0-10 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
22.0° c
1000 m
1.55E+3 persons
1.41E+6 persons
2.75E+4 head
2.57E+5 head
1.39E+4 ha
Site D—Arkansas
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
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
14.8° C
600 m
1.18E+4 persons
5.92E+5 persons
1.19E+4 head
2.57E+5 head
2.94E+3 ha
A-8
-------
Table A-l. Characteristics of the generic sites—continued
Site E—New Mexico
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
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
13.2° C
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 Temperature:
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
8.10 c
700 m
0 persons
1.19E+4 persons
1.86E+3 head
1.47E+5 head
1.77E+4 ha
(WBAN=24033)
A-9
-------
Table A-l. Characteristics of the generic sites—continued
Site G—Idaho
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
Average Mixing Height:
Population: (0-10 km):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (0-80 km):
Vegetable Crop Area:
(0-80 km)
Pocatello (WBAN-24156)
A-F
54/01-62/12
27.4 cm
7.80 c
615 m
4.17E+4 persons
1.38E+5 persons
1.72E+4 head
1.45E+5 head
1.44E+5 ha
Air concentrations are ground level sector averages.
Dispersion is calculated from annual average meteorological data.
Depletion due to dry deposition and precipitation scavenging is
calculated for particulates and reactive vapors. AIRDOS-EPA does
not perform ingrowth calculations for airborne radionuclide chains.
The air concentrations are the basis for inhalation and submersion
dose calculations.
Ground surface and soil concentrations are calculated for
those nuclides subject to deposition due to dry deposition and
precipitation scavenging. A 100-year accumulation period was
generally used unless otherwise indicated. A general soil removal
rate of 0.02 y~l was assumed for deposited radionuclides.
Ingrowth, as well as decay and environmental removal, is calculated
for members of radionuclide decay chains. AIRDOS-EPA provides no
means for calculating resuspended air concentrations or subsequent
redeposition to the ground surface.
A-10
-------
The output from AIRDOS-EPA contains calculated radionuclide
Intakes and external exposure. This file is used as input to
DARTAB (Be81) to produce the dose and risk tables used in the
individual and collective assessments. The dose and risk
conversion factors used for these calculations are discussed
in Appendix B.
A.6 Individual 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 individuals living near each
facility within the source category. The location on the assessment
grid which provides the greatest lifetime risk (all pathways
considered) was chosen for the maximum individual.
2) The organ dose-equivalent rates in the tables are based on
the calculated environmental concentrations by AIRDOS-EPA. For
inhaled or ingested radionuclides, the conversion factors are the
70-year values calculated by RADRISK (Du80). 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 considered 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. We assumed that in an
agriculturally unproductive location, people would home-produce a
portion of their food comparable to residents of an urban area, and
so we used the urban fraction for such nonurban locations. The
fractions of home produced food consumed by individuals for the
generic sites are shown in Table A-2. Trial runs showed little
difference between assuming that the balance of the maximum
individual's diet comes from the assessment area or from outside the
assessment area.
A.7 Collective Assessment
The collective 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
stationary in time.
A-ll
-------
Table A-2. Sources of food for the maximum Individual
Food Urban/Low productivity Rural
(Sites A, B, E-G) (Sites C & D)
Fl F2 F3 Fl F2 F3
Vegetables
Meat
Milk
.076
.008
0.
0.
0.
0.
.924
.992
1.
.700
.442
.399
0.
0.
0.
.300
.558
.601
Fl and F2 are the home-produced fractions 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. Fractions are based on an analysis of
household data from the USDA 1965-1966 National Food Consumption
Survey (USDA72).
2) Average agricultural production data for the state in
which the generic site is located were assumed for all distances
greater than 500 meters from the source. For distances less than
500 meters no agricultural production is calculated.
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. Seventy-year dose
commitment factors (as for the individual case) are used for
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
for one year of plant operation.
A.8 AIRDOS-EPA Parameters and Input Data
Site independent parameter values used for AIRDOS-EPA are
summarized in Table A-5. Element dependent factors (Ba81) are
listed in Table A-6.
Mixing Height and Deposition
Table A-3 summarizes the mixing heights, rainfall rates, and
scavenging coefficients used for the generic sites. A dry
A-12
-------
deposition velocity of 0.0018 m/s was used for particulates and
0.035 m/s for reactive vapors (e.g., elemental iodine) unless
otherwise indicated.
Table A-3. Some site parameters used with AIRDOS-EPA
Average mixing Rainfall
Generic height rate
site
Site A
Site B
Site C
Site D
Site E
Site F
Site G
(m)
1000
600
1000
600
800
700
615
(cm/y)
102
102
142
127
20
20
27
Scavenging
coefficient
(s-1)
l.OE-5
l.OE-5
1.4E-5
1.3E-5
2.0E-6
2.0E-6
2.7E-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. Sites E through G are
relatively dry locations as reflected by the scavenging coefficients.
Meteorological Data
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. These data were
converted to AIRDOS format wind data using the utility program
listed in Appendix A of EPA80.
Dairy and Beef Cattle
Dairy and beef cattle distributions are part of the
AIRDOS-EPA input. A constant cattle density is assumed except for
A-l 3
-------
the area closest to the source or stack In the case of a point
source, i.e., no cattle within 500 m of the source. The cattle
densities are provided by State in Table A-4. 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~l and 200 kilograms of beef/animal slaughtered. A 180-day
grazing period was assumed for dairy and beef cattle.
Vegetable Crop Area
A certain fraction of the land within 80 km of the source is
used for vegetable crop production and is assumed to be uniformly
distributed throughout the entire assessment area with the exception
of the first 500 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-4) by State were obtained from the
production densities by assuming a production rate of 2 kilograms
(fresh weight)/year-square meter (NRC77).
Population
The population data for each generic site were generated by a
computer program, SECPOP (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 (GED) and the
geographic coordinates of the population centroid for the district.
In the Standard Metropolitan Statistical Areas (SMSA) the CED is
usually a "block group" which consists of a physical city block.
Outside the SMSAs the CED is an "enumeration district," which may
cover several square miles or more in a rural area.
There are approximately 250,000 CEDs in the United States
with an average population of about 800 persons. 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 personal 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. Hence, 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.
A-14
-------
A. 9 DARTAB—Dose and Risk Tables
The intermediate output files of ingestion and inhalation
intake and ground level air and ground surface concentrations of
radionuclides were processed by DARTAB (Be80) using RADRISK (Du80)
dose and risk conversion factors to produce the dose and risk
assessments for this report.
The internal dose conversion factors are for a 70-year dose
commitment. In general, the dose factors are calculated with
metabolic factors and methodology very similar to that used in
ICRP-30 (ICRP30). The principal differences are in some GI transfer
fraction (f^) values (e.g., uranium and transuranics) which have
been chosen to be representative of environmental rather than
occupational exposure settings.
Dose equivalents are calculated using a value of 20 for Q for
alpha radiation. A weighted mean of the dose equivalent to the
principal organs and tissues was calculated as an effective whole
body dose equivalent. The weighting factors (see Table A-7) are
based on the lifetime risk from a constant uniform low LET dose rate
to all organs and tissues. Note that the weighting factors are
unaffected by any change in the risk model which affects all organs
and tissues proportionally. The resultant quantity, identified as
the weighted sum in tables in the text, is risk equivalent for
external low LET exposures. That is, two different external
exposure conditions which provide different values for individual
organ and tissue dose rates but the same value for the weighted sum,
will provide, using the RADRISK methodology, the same lifetime
risk. The weighted sum is conceptually to the ICRP-26 effective
whole body dose equivalent (ICRP77) but with a more explicit set of
organs and tissues and a different set of weighting factors.
For intakes of radionuclides where the committed dose is
delivered over a considerable period of time or where both high and
low LET radiation are present, the weighted sum is no longer risk
equivalent. Our experience, however, is that the weighted sum is
between a factor of 1 and 2 greater than the true risk equivalent
low LET dose rate. Hence, the weighted sum provides a single
measure of dose equivalent which approximately corresponds to risk.
A-15
-------
Table A-4. Cattle densities and vegetable crop
distributions for use with AIRDOS-EPA
State
Alabama
Arizona
Arkansas
California
Colorado
Connecticut
Delaware
Florida
Georgia
Idaho
Illinois
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
3.34E4-1
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-16
-------
Table A-4. Cattle densities and vegetable crop
distributions for use with AIRDOS-EPA—continued
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
2.031+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-17
-------
Table A-5. Site Independent parameters used for AIRDOS-EPA
generic site assessments
Symbolic
variable
BRTHRT
T
DDI
TSUBH1
TSUBH2
TSUBH3
TSUBH4
LAMW
TSUBE1
TSUBE2
YSUBV1
YSUBV2
FSUBP
FSUBS
Description
Breathing Rate (cm3/!!)
Surface buildup time (days)
Activity fraction after washing
Time delay-pasture grass (h)
Time delay-stored food (h)
Time delay-leafy vegetables (h)
Time delay-produce (h)
Weathering removal rate
factor (h"1)
Exposure period-pasture (h)
Exposure period-crops or leafy
vegetables (h)
Productivity-pasture (dry
weight) (kg/m^)
Productivity-crops and leafy
vegetables (kg/nr)
Time fraction-pasture grazing
Pasture feed fraction-while
Value
9.17E+6
3.65E+4
0.5
0.0
2.16E+3
336.
336.
2.10E-3
720.
1.44E+3
.280
.716
0.40
QSUBF
TSUBF
UV
UM
UF
UL
TSUBS
pasture grazing 0.43
Feed or forage consumption
rate (kg-dry/day) 15.6
Consumption delay time-milk (d) 2.0
Vegetable utilization rate (kg/y) 176.
Milk utilization rate (kg/y) 112.
Meat utilization rate (kg/y) 85.
Leafy vegetable utilization
rate (kg/y) 18.
Consumption time delay-meat (days) 20.
A-18
-------
Table A-5.
Site independent parameters used for AIRDOS-EPA
generic site assessments (Continued)
Symbolic
variable
Description
Value
FSUBG
FSUBL
TSUBB
P
TAUBEF
MSUBB
VSUBM
Rl
R2
Produce fraction (garden of interest) 1.0
Leafy veg fraction (garden of
interest) 1.0
Soil buildup time (y) 100.
Effective surface density of soil
(kg/m2) 215.
Meat herd-slaughter rate
factor (d-1) 3.81E-3
Mass of meat of slaughter (kg) 200.
Milk production rate of cow (L/d) 11.0
Deposition interception fraction-
pasture 0.57
Deposition interception fraction-
leafy vegetables 0.20
A-19
-------
Table A-6. Element dependent factors used in AIRDOS-EPA assessments
Element
Ac
Ac
Ag
Am
As
Ba
Be
Bi
Ce
Cm
Co
Co
Cr
Cr
Cs
Eu
Fe
Ga
Hg
Ir
I
La
La
Mn
Mo
Na
Nb
Pa
Pb
Po
Po
Pr
Pu
P
Ra
Clearance
class
Y
W
Y
Y
W
D
W
W
Y
Y
W
Y
D
Y
D
Y
W
W
W
Y
D
W
Y
W
D
D
W
Y
W
W
D
Y
Y
D
W
Fl
0.10E-2
0.10E-2
0.50E-1
0.10E-2
0.30E-1
0.10
0.20E-2
0.50E-1
0.10E-3
0.10E-2
0.50E-1
0.50E-1
0.10
0.10
0.95
0.10E-3
0.10
0.10E-2
0.20E-1
0.10E-1
0.95
0.10E-3
0.10E-3
0.10
0.95
0.95
0.10E-1
0.10E-2
0.80E-1
0.10
0.10
0.10E-3
0.30E-4
0.80
0.20
(d?L)
2.0E-5
2.0E-5
3.0E-2
3.6E-5
6.2E-5
3.5E-4
9.1E-7
5.0E-4
2.0E-5
2.0E-5
2.0E-3
2.0E-3
2.0E-3
2.0E-3
5.6E-3
2.0E-5
5.9E-5
5.0E-5
9.7E-6
2.0E-6
9.9E-3
2.0E-5
2.0E-5
8.4E-5
1.4E-3
3.5E-2
2.0E-2
5.0E-6
8.7E-5
1.2E-4
1.2E-4
2.0E-5
5.3E-8
1.6E-2
5.9E-4
Ff
(d/kg)
1.6E-6
1.6E-6
1 . 7E-2
1.6E-6
2.0E-3
3.2E-3
l.OE-3
1.3E-2
1.2E-3
1.6E-6
1.3E-2
1.3E-2
2.4E-3
2.4E-3
1.4E-2
4.8E-3
4.0E-2
1.4
2.6E-1
1.5E-3
7.0E-3
2.0E-4
2 . OE-4
8.0E-4
8.0E-3
3.0E-2
2.8E-1
1.6E-6
9.1E-4
8.7E-3
8.7E-3
4.7E-3
1.9E-8
4.6E-2
5. OE-4
B!VI
1 . OE-2
l.OE-2
6.0E-1
9.8E-3
3.9E-3
6.1E-2
1.7E-3
6.0E-1
2.6E-2
1.3E-3
3.7E-2
3.7E-2
2.4E-2
2.4E-2
1.4E-1
l.OE-2
9.3E-3
l.OE-3
1.5
5.2E+1
2.0E-1
4.2E-3
4.2E-3
3.9E-2
3.4
2.1E-1
3.8E-2
l.OE-2
1.4E-1
4.2E-3
4.2E-3
l.OE-2
6.7E-3
4.4E+0
l.OE-1
Blv2
2.5E-3
2.5E-3
1.5E-1
1.5E-3
1.7E-2
2.0E-1
4.2E-4
1.5E-1
6.2E-3
1.7E-3
9.3E-3
9.3E-3
6.0E-3
6.0E-3
9.1E-3
2.5E-3
2.3E-3
2.5E-4
3.8E-1
1.3E+1
5.5E-2
1.1E-3
1.1E-3
9.8E-3
2.2E-1
5.2E-2
9.4E-3
2.5E-3
4.8E-3
2.6E-4
2.6E-4
2.5E-3
1.1E-3
1.1
2. OE-2
A-20
-------
Table A-6. Element dependent factors used in AIRDOS-EPA assessments
(Continued)
Element
Rb
Ru
Ru
Sb
Sn
Sr
S
Tb
Tc
Th
Th
Tl
U
U
Y
Zn
Zr
Clearance
class
D
W
Y
W
W
D
D
Y
W
W
Y
W
Y
D
W
W
W
Fl
0.95
0.40E-1
0.40E-1
0.50E-1
0.50E-1
0.20
0.95
0.10E-3
0.80
0.10E-2
0.10E-2
0.95
0.20E-2
0.50E-1
0.10E-3
0.50
0.20E-2
Fm
(d/L)
1.2E-2
6.1E-7
6.1E-7
2.0E-5
1.2E-3
1 . 1E-3
1.6E-2
2.0E-5
9 . 9E-3
5.0E-6
5.0E-6
2.2E-2
1.4E-4
1.4E-4
2.0E-5
l.OE-2
8.0E-2
Ff
(d/kg)
3.1E-2
1.8E-3
1.8E-3
4.0E-3
8.0E-2
3.0E-4
l.OE-1
4.4E-3
8.7E-3
1.6E-6
1.6E-6
4.0E-2
1.6E-6
1.6E-6
4.6E-3
3.0E-2
3.4E-2
BiVl
2.5E-1
1 . 7E-1
1.7E-1
1.1E-1
2.0E-2
2.4
2.4
l.OE-2
2.2E+2
6.3E-3
6.3E-3
1.0
2.1E-2
2. IE -2
1 . 1E-2
3.9E-1
6.8E-4
Biv2
6.3E-2
1.6E-2
1.6E-2
2.8E-2
5.0E-3
2.2E-1
5.9E-1
2.6E-3
1.1
3.5E-4
3.5E-4
2.5E-1
4.2E-3
4.2E-3
4.3E-3
9.8E-2
1.7E-4
A-21
-------
Table A-7. Weighting factors used for weighted sum dose equivalent
Organ or
tissue
Lung (Pulmonary)
Breast (Muscle)
Red marrow
Endosteum
Stomach
Small intestine
Upper large intestine
Lower large intestine
Liver
Kidneys
Bladder
Pancreas
Thyroid
Thymus
Spleen
Testes
Ovaries
Uterus
Risk(a>
(x 10~5)
0.608
0.399
0.326
0.031
.087
.017
.035
.069
.156
.035
.035
.121
.085
.017
.017
.017
.017
.017
Weighting
factor
0.2911
.1911
.1561
.0148
.0416
.0081
.0168
.0330
.0747
.0168
.0168
.0579
.0407
.0081
.0081
.0081
.0081
.0081
Total
2.089
1.0000
(a)lndividual lifetime risk for a low LET dose rate of 1 mrad/y
(see Appendix-B).
A-22
-------
REFERENCES
At74
Ba8l
Be81
EPA80
Ge78
Ho72
ICRP26
ICRP30
Mo79
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.
Baes C. F. Ill, and R. D. Sharp, A Director of Parameters
Used in a Series of Assessment Applications of the
AIRDOS-EPA and DARTAB Computer Codes, ORNL-5710, Oak Ridge
National Laboratory, Oak Ridge, Tennessee, March 1981.
Begovich C. L., K. F. Eckerman, E.G. Schlatter, S.Y. Ohr,
and R. 0. Chester, 1981, DARTAB: A program to combine
airborne radionuclide environmental exposure data with
dosimetric and health effects data to generate tabulation
of predicted impacts. ORNL/5692, Oak Ridge National
Laboratory, Tennessee, August 1981.
Environmental Protection Agency, Radiological Impact Caused
by Emissions of Radionuclides into Air in the United
States, EPA 520/7-79-006, EPA, Office of Radiation
Programs, Washington, D.C., December 1980.
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.
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.
International Commission on Radiological Protection, ICRP
Publication No. 26, Pergamon Press, N.Y., January 1977.
International Commission on Radiological Protection, ICRP
Publication No. 30, Pergamon Press, N.Y.
Moore R. E., C. F. Baes, III, L. M. McDowell-Boyer, A. P.
Watson, F. 0. Hoffman, J. C. Pleasant, C. W. Miller, 1979,
AIRDOS-EPA: A Computerized Methodology for Estimating
Environmental Concentrations and Dose to Man from Airborne
Releases of Radionuclides, EPA 520/1-79-009, EPA Office of
Radiation Programs, Washington, D.C. 20460, December 1979.
A-23
-------
REFERENCES (Continued)
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.
A-24
-------
APPENDIX B
THE BASIS FOR RISK ESTIMATES
- RADRISK CODE -
-------
APPENDIX B: THE BASIS FOR RISK ESTIMATES
- RADRISK CODE -
CONTENTS
Page
Introduction B-5
B.I The RADRISK Code B-6
B.2 Risk Estimates for Inhaled Radon and Radon
Decay Products B-9
References B-13
TABLES
B-l Number of premature deaths due to chronic radiation exposure
by type of cancer B-7
B-2 Risk factors for high and low LET radiation by type of
cancer B-8
FIGURES
B-l Excess fatal lung cancer in various miner groups by
cumulative exposure B-10
B-3
-------
APPENDIX B: THE BASIS FOR RISK ESTIMATES
Introduction
There are two kinds of risks from the low levels of ionizing
radiation characteristic of exposures to radionuclides released into
the environment. The most important of these is cancer, which is fatal
at least half the time. The other risk is the induction of hereditary
effects in descendants of exposed persons; the severity ranges from
fatal to inconsequential. We assume that at low levels of exposure the
risk of cancer and hereditary effects is in proportion to the dose
received, and that the severity of any induced effect is independent of
the dose level. While the probability of a given type of cancer
increases with dose, such a cancer induced at one dose is equally as
debilitating as that same type of cancer induced at another dose. For
these effects, we assume that there is no completely risk-free level of
radiation exposure.
The risks and effects on health from low-level ionizing radiation
were reviewed for EPA by the National Academy of Sciences in reports
published in 1972 and in 1980 (NAS72a, NASSOb). We use these studies
and others to estimate the risks associated with the radiation doses
calculated in this report.
The individual lifetime risk is estimated for the "most exposed
individuals"; these are the people at the location of highest lifetime
risk. The risk to the individual is the risk of premature death from
cancer due to the radiation dose. The risk calculation considers all
important radionuclides, pathways, and organs of the body.
i
The risk to an individual can be related to other parameters. For
example, we can determine which part of the risk is due to each
radionuclide moving through a specific pathway or which organ is at
highest risk. This information is helpful when deciding which control
strategies will be the most effective.
The risk to populations is also estimated; that is, the number of
future effects on health that are committed for each year that the
B-5
-------
source operates. The dose is not necessarily delivered to people
during the years of release because radionuclides with long half-lives
may take a long time to move through environmental pathways to people.
Like the individual lifetime risk, the total risk to populations
can be related to other parameters, such as organ, radionuclide, or
exposure pathway.
B.I The RADRISK Code
The estimates of cancer risk are calculated using a computer code
called RADRISK. In RADRISK, the group at risk is a hypothetical cohort
of 100,000 people, all born simultaneously and subject to the same
risks throughout their lives. Each member is assumed to be exposed at
a constant rate to a unit concentration of radionuclides. For each
radionuclide and for each pathway, the code calculates the number of
premature deaths due to radiation and the number of years of life lost
due to these deaths (Table B-l).
When radionuclides are inhaled, they enter the lung. The ICRP
Task Group lung model is used to predict where in the lung they go and
how fast they are removed to other parts of the body. Depending on the
particle size and solubility in lung fluids, there is removal of some
of this material to the gastrointestinal (GI) tract and absorption by
the blood. A GI tract model is used to estimate how much of the
material reaching the tract is absorbed by the blood.
After absorption by the blood, radionuclides are distributed among
the organs according to uptake and metabolic information supplied to
RADRISK. Most of this information is taken from ICRP-30 (ICRP30)
supplemented with additional information summarized by EPA (Su81).
Dose rates are calculated with the help of models that simulate the
biological processes involved when radionuclides enter and leave organs.
Cancers do not appear immediately after exposure. There is a
latent period before the cancers are observed; the length, usually
years, varies with the type of cancer. Thereafter, there is a
specified "plateau" period when there is a finite probability of
cancer. The plateau period varies with the type of cancer.
Lifetime probabilities for many types of cancer, in many organs,
are followed and risks calculated by the RADRISK code. At the same
time, competing risks unrelated to the radiation exposure are accounted
for. The risk factors for high and low LET radiation by type of cancer
are listed in Table B-2. We believe these factors are accurate to an
order of magnitude only; therefore, risk estimates are reported to only
one significant figure.
A more detailed description of RADRISK can be found in
ORNL/TM-7745, "Estimates of Health Risk from Exposure to Radioactive
Pollutants" (Su81).
B-6
-------
Table B-l. Number of premature deaths due to chronic radiation
exposure by type of cancer
Type of
cancer
Leukemia
Bone
Lung
Breast
Liver
S tomach
Pancreas
Lower large
intestine
Kidneys
Bladder
Upper large
intestine
Small intestine
Ovaries
Testes
Spleen
Uterus
Thymus
Thyroid
Latency
(years)
2
5
10
15
15
15
15
15
15
15
15
15
15
15
15
15
15
2
Plateau
(years)
25
304
110
110
110
110
110
110
110
110
110
110
110
110
110
110
110
45
Number of premature deaths
in cohort from chronic
1 mrad/y exposure'3/
0.326
0.031
0.608
0.399
0.154
0.087
0.121
0.069
0.035
0.035
0.035
0.017
0.017
0.017
0.017
0.017
0.017
0.085
(a)Low-LET.
B-7
-------
Table B-2. Risk factors for high and low LET radiation by
type of cancer
Type of
cancer
Leukemia
Bone
Lung
Breast
Liver
S tomach
Pancreas
Lower large
intestine
Kidneys
Bladder
Upper large
intestine
Small intestine
Ovaries
Testes
Spleen
Uterus
Thymus
Thyroid
Risk
Low-LET radiation
(Deaths/106 rad
person-y at risk)
2.3
0.2
3.0
2.3
0.9
0.5
0.7
0.4
0.2
0.2
0.2
0.1
0.1
0.1
0.1
0.1
0.1
0.4(a)
factors
High-LET radiation
(Deaths/106 rad
person-y at risk)
46
4
30
2.3
9
5
7
4
2
2
2
1
1
1
1
1
1
0.40.04 for 131I and longer-lived radioiodine.
B-8
-------
B'2 Risk Estimates for Inhaled Radon and Radon Decay Products
An estimate of the health risk from inhaling radon and its
snort-lived decay products is done separately. The units of exposure,
Working Level and Working Level Month, are unusual and do not fit into
the RADRISK computer code. The risks due to exposure to radon and
radon decay products have been calculated independently of the RADRISK
code.
Risk of Lung Cancer from Inhaling Radon Decay Products
The high incidence of lung cancer mortality among underground
miners exposed to radon decay products is well documented (EPA79a,
Ar79, Ar8l). Uranium miners are particularly affected, but lead, iron,
and zinc miners exposed to relatively low levels of radon decay
products also show an increased lung cancer mortality that correlates
with exposure to radon decay products. The type of lung cancer most
frequently observed, moreover, is relatively uncommon in the general
population.
Risk estimates for the general public based on these studies of
miners are far from precise. First, and most important, the relatively
small number of miners at risk causes considerable statistical
uncertainty in estimating the risk of lung cancer (see Figure B-l).
Second, most miners were exposed to much higher levels of radon decay
products than usually occur in the general environment. Third, the
miners' exposure levels are uncertain. Fourth, significant demographic
differences exist between miners and members of the general public—the
miners were healthy males over 14 years old, many of whom smoked.
However, we believe that information from the studies of miners
provides useful estimates, if not precise predictions, of the risks to
the general population from radon decay products.v^)
Since the miners have not all died, the number of eventual excess
lung cancers must be projected from current data by using mathematical
models. There are two ways to do this. One method, called the
relative risk model, yields the percent increase in the normal
incidence of cancer per unit of exposure. The other, called the
absolute risk model, yields the absolute numerical increase in cancers
per unit of exposure. In the relative risk model, it is assumed that
the increased risk is proportional to the age-dependent natural
incidence of the disease for each year an individual remains alive
following exposure. In the absolute risk model, it is assumed that the
added risk is constant each year an individual remains alive following
exposure, i.e., independent of natural incidence.
(l)See "Indoor Radiation Exposure Due to Radium-226 in Florida
Phosphate Lands" (EPA79a) for greater detail of such an analysis.
B-9
-------
80
70
|- 60
CO
o:
UJ
o
< 50
§
UJ
-j
CO
40
D
00
30
20
10
Ai:
()
O CZECH-URANIUM
D SWEDEN-LEAD, ZINC (A), IRON (R.J)
A UNITED STATES-URANIUM
f CANADA-URANIUM
I 95% CONFIDENCE LIMITS
RD
I
100 200 300 400 500
CUMULATIVE WORKING LEVEL MONTHS
GOO
700
Figure B-l. Excess fatal lung cancer in various miner groups
by cumulative exposure (Ar79).
B-10
-------
When using the relative risk model, we conclude that a 3-percent
increase in the number of lung cancer deaths per WLM is consistent with
data from the studies of underground miners. However, because of the
differences between adult male miners and the general population
(EPA79a), we estimate that the risk to the general population may be as
low as 1 percent or as high as 5 percent. To develop absolute risk
estimates, we use the estimate of 10 lung cancer deaths per WLM for 1
million person-years at risk as reported by the National Academy of
Sciences (NAS76).
In 1978, Land and Norman (La78) reported that, in Japanese A-bomb
survivors, radiation-induced lung cancers had a temporal distribution
of occurrence similar to naturally-occurring cancers. Further, they
concluded that the cumulative distribution of radiation-induced lung
cancer across time after exposure was consistent with either a relative
risk model of cancer incidence or with an age-specific absolute risk
model. Also in 1978, Smith and Doll (Sm78) reported the risk of cancer
developing at most "heavily irradiated" sites in ankylosing spondylitic
patients treated with x-rays was directly proportional to the risk of a
tumor in the absence of radiation; in other words, a relative-risk
response. In the most recent report on the Japanese A-bomb survivors,
Kato and Schull (Ka82) repeated the observation that radiation-induced
lung cancer develops only after the survivors attain the age at which
this cancer normally develops. The evidence in these three reports
points to relative-risk or age-specific absolute risk models as the
most appropriate models for radiation-induced lung cancer.
Recent information from China provides additional evidence.
Shi-quan and Xiao-ou (Sh82) have reported that in Chinese tin miners
exposed to radon and its decay products, the lung cancers develop at
the age at which lung cancer normally develops. Those who started
mining at age 8 or 9 had an induct ion-la tent period about 10 years
longer than those who started mining at age 19 or 20. In view of these
observations that a simple absolute risk model is inappropriate for
estimating the risk of lung cancer due to radon decay products, we do
not use it.
A comparison of risks calculated using a relative model and an
age-specific absolute risk model (BEIR III, NAS80b) showed both models
give numerically similar results (860 cases/10^ person-WLM versus 850
cases/10^ person-WLM) (RPC80). Because of the similarly in risk
estimates, we use relative risk estimates for exposure to radon decay
products.
To estimate the total number of lung cancer deaths from increased
levels of radon decay products, we use a life-table analysis (Bu81).
This analysis uses the risk coefficients just discussed. It also takes
into account the length of time a person is exposed and the number of
years a person survives other potential causes of death. The result is
expressed as the number of premature lung cancer deaths that would
occur due to lifetime radiation exposure of 100,000 persons. We
assume, further, that injury caused by alpha radiation is not
repairable, so that exposed persons remain at risk for the balance of
their lifetimes.
B-ll
-------
To summarize, we estimate that a person exposed to 0.01 WL (0.27
WLM/y) over a lifetime incurs a 1.7 percent (1 in 60) additional chance
of contracting a fatal lung cancer. This estimate is made using the
relative risk model and assuming children are no more sensitive than
adults. If exposure to radon decay products during childhood carries a
three times greater risk, this estimated lifetime relative risk would
increase by about 50 percent (EPA79a). For comparison, a lifetable
analysis for the same population not exposed to excess radiation yields a
2.9-percent chance of lung cancer death.
Even though the risk of lung cancer due to radon decay products
varies with age, it is sometimes convenient to express these risks on an
average annual basis. We calculate a person's average annual risk from a
lifetime of exposure by dividing the lifetime risk estimates given above
by an average lifespan of 71 years. (Note that this is not the same as
applying the risk coefficient for 71 years, since the lifetable analysis
accounts for other causes of death.) Based on the risk model and
assumptions just described for lifetime exposure, we estimate an average
of 2.4 lung cancer deaths per year for each 100 person-working-levels of
such exposure. "Person-working-levels" is the population's collective
exposure; that is, the number of people times the average concentration
of radon decay products (in working levels) to which they are exposed.
Radiation risk can also be stated in terms of years of life lost due
to cancer death. In the relative risk model, the distribution of ages at
which lung cancer caused by radiation occurs is the same as that for all
lung cancer in the general population. Since lung cancer occurs most
frequently in people over 70 years of age, the years of life lost per
fatal lung cancer—14.5 years on the average—is less than for many other
fatal cancers.
Our assessments are for current conditions because we use recent
population data. If the population lifestyle, medical knowledge,
and other patterns of living affecting mortality remain unchanged, then
these rates of lung cancer death should persist for the indefinite
future. We believe this is a prudent assumption.
B-12
-------
REFERENCES
Ar79
Ar81
Bu81
EPA79a
Ka82
ICRP79
La78
NAS72a
NAS72b
NAS76
Archer V.E., "Factors in Exposure Response Relationships of
Radon Daughter Injury," in Proceedings of the Mine Safety and
Health Administration Workshop on Lung Cancer Epidemiology and
Industrial Applications of Sputum Cytology, November 14-16,
1978, Colorado School of Mines Press, Golden, Colorado, 1979.
Archer V.E., "Health Concerns in Uranium Mining and Milling,"
J. Occup. Med. 23:502, 1981.
Bunger B.M., Cook J.R. and Barrick M.K., "Life Table
Methodology for Evaluating Radiation Risk: An Application Based
on Occupational Exposures," in Health Physics, 40:439-455,
1981.
Environmental Protection Agency, "Indoor Radiation Exposure
Due to Radium-226 in Florida Phosphate Lands," EPA
520/6-78-013, Office of Radiation Programs, Washington, D.C.,
July 1979.
Kato H. and Schull W.J., "Studies of the Mortality of A-Bomb
Survivors. 7. Mortality, 1950-1978: Part 1. Cancer
Mortality." Radiat. Res. 90:395-432 (1982).
International Commission on Radiological Protection, "Limits
on Intakes of Radionuclides for Workers," A Report of
Committee 2 of the ICRP, Pergamon Press, Oxford, 1979.
Land C.E. and Norman J.E., "Latent Periods of Radiogenic
Cancers Occurring Among Japanese A-Bomb Survivors," pp. 29-47
in Late Biological Effects of Ionizing Radiation, Volume I,
IAEA, Vienna, 1978.
National Academy of Sciences, The Effects on Populations of
Exposure to Low Levels of Ionizing Radiation, Report of the
Advisory Committee on the Biological Effects of Ionizing
Radiation, PB-239 735/AS, NAS, National Technical Information
Service, Springfield, Virginia, 1972.
National Academy of Sciences, "Water Quality Criteria,"
EPA-R3-73-033, USEPA, Washington, D.C. 1972.
National Academy of Sciences, "Health Effects of Alpha
Emitting Particles in the Respiratory Tract," Report of Ad Hoc
Committee on "Hot Particles" of the Advisory Committee on the
Biological Effects of Ionizing Radiations, EPA Contract No.
68-01-2230, EPA 520/4-76-013, USEPA, Washington, D.C. October
1976.
B-13
-------
REFERENCES (Continued)
NASSOa National Academy of Sciences, "Drinking Water and Health,"
Volume 3, NAS, National Academy Press, Washington, D.C., 1980.
NASSOb National Academy of Sciences, "The Effects on Population of
Exposure to Low Levels of Ionizing Radiation," Committee on
the Biological Effects of Ionizing Radiations, NAS, National
Academy Press, Washington, D.C., 1980.
Sh82 Shi-quan S. and Xiao-on Y., "Induction-Latent Period and
Temporal Aspects of Miner Lung Cancer," unpublished report (in
English), 1982.
Sm78 Smith P.G. and Doll R., "Age- and Time-Dependent Changes in
the Rates of Radiation-Induced Cancers in Patients with
Ankylosing Spondylitis Following a Single Course of X-ray
Treatment," pp. 205-218 in: Late Biological Effects of
Ionizing Radiation, Volume I., IAEA, Vienna, 1978.
Su81 Sullivan R.E., et al., "Estimates of Health Risk from Exposure
to Radioactive Pollutants," ORNL/TM-7745, Oak Ridge National
Laboratory, Oak Ridge, Tennessee, 1981.
B-14
-------
APPENDIX C
CALCULATION OF RADON-222 CONCENTRATIONS
-------
5643B 3/17/83
APPENDIX C: CALCULATION OF RADON-222 CONCENTRATIONS
The annual average radon-222 concentrations in air at various
distances in each of 16 wind directions (i.e., sector segments) from an
underground uranium mine vent emitting 1 kCi/y of radon-222 are presented
in Table C-l. These concentrations were calculated for a 3-meter vent
height using meteorological data from Grants, New Mexico (see Table A-l
of Appendix A).
Using data from Table C-l, the relationship between radon-222
concentration and distance from a mine vent can be expressed as a power
function. For distances between 0.2 and 3 kilometers, the radon-222
concentration averaged over all wind directions, can be described by the
following expression:
Cj = 0.11 Q(Xj)-1-72
where
Cj = radon-222 concentration in air in pCi/L at location j
Xj = distance in kilometers from mine vent to location j
Q = radon emission rate in kCi/y
Since underground uranium mines emit radon-222 to air through
multiple vents rather than a single vent, the data in Table C-l were used
to estimate the radon-222 concentrations in air at various distances from
a reference underground uranium mine (see Table 5-9) with 5 vents
distributed as shown in Figure 5-4 and each emitting 1 kCi/y of
radon-222. Examples 1 and 2 show how the data from Table C-l were used
to make these calculations.
It should be emphasized that the radon-222 concentration in air at
any specific location near a uranium mine with multiple vents is highly
dependent upon the spatial distribution of the vents with respect to the
location of interest and the wind frequency distribution. The data in
Table 5-9 illustrate the levels which could occur in a given situation.
C-3
-------
For other situations (i.e., different spatial distribution of the vents,
wind frequencies, etc.), the radon-222 concentrations in air could be
higher or lower than the values shown in Table 5-9.
Example 1
Calculation of radon-222 concentrations in air at receptor
location using dispersion factors based on the average of all
wind directions
Source
Vent 1
Vent 2
V.^nt 3
Vent 4
Vent 5
Distance^3'
(km)
1.3
1.7
2.1
1.1
0.5
Radon-222
(pCi/L)
0.07
0.04
0.03
0.09
0.37
Total
0.60
(^Distance from vent to receptor location 0.5 km in southeasterly
direction from Vent 5.
'k'Radon-222 concentrations from Table C-l for average of all wind
directions.
Source
Vent 1
Vent 2
Vent 3
Vent 4
Vent 5
Example 2
Calculation of radon-222 concentrations in air at receptor
location using dispersion factors based on wind frequency
from each vent to the receptor location
Distance^3)
(km)
1.3
1.7
2.1
1.1
0.5
Direction(b)
S
SSW
S
SW
SE
Radon-222
(pCi/L)
0.07
0.01
0.03
0.02
1.32
Total
1.45
'a'Distance from vent to receptor location 0.5 km in southeasterly
direction from Vent 5.
^'Direction from vent to receptor location.
'c'Radon concentration from Table C-l for distance and direction from
each vent.
C-4
-------
Table C-l. Annual average radon-222 concentrations in air at
selected distances from an underground uranium mine vent
emitting 1 kCi/y
Distance
(meters)
100
150
200
300
400
500
800
1000
1500
2000
3000
4000
5000
8000
10000
15000
20000
30000
40000
50000,
Radon-222 air concentrations (pCi/L)
N
1.46
1.20
8.89E-1
5.05E-1
3.18E-1
2.17E-1
9.38E-2
6. 24E-2
3.22E-2
2.01E-2
1.03E-2
6. 97E-3
5.10E-3
2.64E-3
1.94E-3
1.21E-3
8. 62E-4
5. 35E-4
3.79E-4
2.89E-4
NNW
1.96
1.54
1.13
6.38E-1
4.00E-1
2.73E-1
1.18E-1
7.84E-2
4.04E-2
2.53E-2
1.30E-2
8.73E-3
6.39E-3
3.31E-3
2.43E-3
1.51E-3
1.08E-3
6. 69E-4
4.74E-4
3.62E-4
NW
4.60
3.26
2.31
1.27
7.92E-1
5.38E-1
2.31E-1
1.53E-1
7.85E-2
4. 88E-2
2.49E-2
1.66E-2
1.21E-2
6.27E-3
4.59E-3
2.83E-3
2.02E-3
1.25E-3
8.83E-4
6.73E-4
WNW
3.22
2.29
1.63
8.96E-1
5.57E-1
3.78E-1
1.63E-1
1.08E-1
5.53E-2
3.44E-2
1.76E-2
1.17E-2
8.54E-3
4.41E-3
3.22E-3
1.99E-3
1.42E-3
8.74E-4
6.18E-4
4.71E-4
C-5
-------
Table C-l. Annual average radon-222 concentrations in air at
selected distances from an underground uranium mine vent
emitting 1 kCi/y (Continued)
Distance
(meters)
100
150
200
300
400
500
800
1000
1500
2000
3000
4000
5000
8000
10000
15000
20000
30000
40000
50000
Radon-222 air
W
6.22E-1
3.96E-1
2.70E-1
1.44E-1
8.85E-2
5.97E-2
2.54E-2
1.68E-2
8.52E-3
5.26E-3
2.66E-3
1.76E-3
1.27E-3
6. 58E-4
4. 82E-4
2.95E-4
2.10E-4
1.29E-4
9.09E-5
6.91E-5
WSW
1.51E-1
1.23E-1
9.14E-2
5.20E-2
3.27E-2
2.23E-2
9.66E-3
6.42E-3
3.30E-3
2.06E-3
1.06E-3
7.16E-4
5.25E-4
2.73E-4
2.00E-4
1.25E-4
8.93E-5
5.54E-5
3.93E-5
3.00E-5
concentrations (pCi/L)
SW
5.98E-1
4.29E-1
3. 06E-1
1.69E-1
1.05E-1
7.13E-2
3.06E-2
2.03E-2
1.04E-2
6.44E-3
3.28E-3
2.19E-3
1.60E-3
8.31E-4
6.09E-4
3.78E-4
2.70E-4
1.67E-4
1.18E-4
8. 97E-5
SSW
5.09E-1
3.94E-1
2.88E-1
1.62E-1
1.02E-1
6.93E-2
2.98E-2
1.98E-2
1.02E-2
6.34E-3
3.25E-3
2.18E-3
1.59E-3
8.29E-4
6.07E-4
3.78E-4
2.70E-4
1.67E-4
1.18E-4
9.02E-5
C-6
-------
Table C-l. Annual average radon-222 concentrations in air at
selected distances from an underground uranium mine vent
emitting 1 kCi/y (Continued)
Distance
(meters)
100
150
200
300
400
500
800
1000
1500
2000
3000
4000
5000
8000
10000
15000
20000
30000
40000
50000
Radon-222 air concentrations (pCi/L)
S
2.77
2.18
1.60
9.01E-1
5.65E-1
3.85E-1
1.66E-1
1.10E-1
5.66E-2
3. 53E-2
1.81E-2
1.22E-2
8.89E-3
4.62E-3
3.39E-3
2.12E-3
1.52E-3
9.45E-4
6.71E-4
5.13E-4
SSE
5.50
4.23
3.08
1.73
1.08
7.36E-1
3.17E-1
2.10E-1
1.08E-1
6.74E-2
3.46E-2
2.32E-2
1.70E-2
8.81E-3
6.46E-3
4.03E-3
2.89E-3
1.80E-3
1.28E-3
9.82E-4
SE
8.89
7.30
5.43
3.08
1.94
1.32
5.72E-1
3.80E-1
1.96E-1
1.23E-1
6.31E-2
4. 25E-2
3.11E-2
1.62E-2
1.18E-2
7.42E-3
5.33E-3
3. 32E-3
2.36E-3
1.81E-3
ESE
5.96
4.91
3.65
2.07
1.30
8.89E-1
3.84E-1
2.56E-1
1.32E-1
8.27E-2
4.26E-2
2.87E-2
2.11E-2
1.10E-2
8.03E-3
5.04E-3
3.62E-3
2.26E-3
1.61E-3
1.23E-3
C-7
-------
Table C-l. Annual average radon-222 concentrations in air at
selected distances from an underground uranium mine vent
emitting 1 kCi/y (Continued)
Distance
(meters)
100
150
200
300
400
500
800
1000
1500
2000
3000
4000
5000"
8000
10000
15000
20000
30000
40000
50000
Radon-222 air
E
2.74
2.19
1.61
9.12E-1
5.72E-1
3.90E-1
1.69E-1
1.12E-1
5.77E-2
3.61E-2
1.85E-2
1.25E-2
9.11E-3
4. 73E-3
3.47E-3
2.16E-3
1.55E-3
9.63E-4
6.84E-4
5.22E-4
ENE
1.92
1.38
9.81E-1
5.41E-1
3.37E-1
2.29E-1
9.84E-2
6. 54E-2
3.36E-2
2.09E-2
1.07E-2
7.18E-3
5.23E-3
2.71E-3
1.98E-3
1.23E-3
8.74E-4
5.41E-4
3.84E-4
2.93E-4
concentrations (pCi/L)
NE
1.44
1.04
7.43E-1
4.11E-1
2.56E-1
1.74E-1
7.48E-2
4.97E-2
2. 54E-2
1.58E-2
8.10E-3
5.42E-3
3.95E-3
2.05E-3
1.50E-3
9.26E-4
6.59E-4
4. 07E-4
2.88E-4
2.19E-4
NNE
1.01
7.36E-1
5.27E-1
2.92E-1
1.82E-1
1.24E-1
5.34E-2
3.55E-2
1.82E-2
1.13E-2
5.81E-3
3.89E-3
2.84E-3
1.47E-3
1.07E-3
6.64E-4
4.73E-4
2.92E-4
2.06E-4
1.57E-4
Average
2.71
2.10
1.53
8.61E-1
5.39E-1
3.68E-1
1.58E-1
1.05E-1
5.41E-2
3.38E-2
1.74E-2
1.17E-2
8.52E-3
4.42E-3
3.24E-3
2.02E-3
1.45E-3
8.98E-4
6.38E-4
4.88E-4
C-8
-------
TECHNICAL REPORT DATA
friease read Instructions on the reverse before completing)
NO.
EPA 520/1-83-001
2.
3. RECIPIENT'S ACCESSION NO.
TITLE AND SUBTITLE
Draft Background Information Document
Proposed Standards for Radionuclides
5. REPORT DATE
March 1983
6. PERFORMING ORGANIZATION CODE
7. AUTHOR(S)
8. PERFORMING ORGANIZATION REPORT NO.
9. PERFORMING ORGANIZATION NAME ANQ ADDRESS
U.S. Environmental Protection Agency
Office of Radiation Programs
Washington, D.C. 20460
10. PROGRAM ELEMENT NO.
11. CONTRACT/GRANT NO.
12. SPONSORING AGENCY NAME AND ADDRESS
13. TYPE OF REPORT AND PERIOD COVERED
14. SPONSORING AGENCY CODE
15. SUPPLEMENTARY NOTES
16. ABSTRACT
This report presents background information that supports the Environmental Protec-
tion Agency's (EPA's) proposed emission standards for radionuclides pursuant to
Section 112 of the Clean Air Act. An analysis of public health impacts from the
following source categories is presented: (1) Department of Energy (DOE) facilities,
(2) Nuclear Regulatory Commission (NRC)-licensed and non-DOE Federal facilities,
(3) coal-fired utility and industrial boilers, (4) uranium mines, (5) phosphate
industry facilities, and (6) mineral extraction industry facilities. For each
source category, the following information is presented: (1) a general description
of the source category, (2) a brief description of the processes that lead to the
emission of radionuclides into air, (3) a summary of emissions data, and (4) esti-
mates of radiation doses and health risks to both individuals and populations.
This is a draft document. After a public comment period, it will be revised as
necessary and issued in final form when the standards are promulgated.
17.
KEY WORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
b.lDENTIFIERS/OPEN ENDED TERMS
c. COSATI Field/Group
Clean Air Act
Radionuclides
Radon
DOE Facilities (Department of Energy)
Nuclear Regulatory Commission licensed
Facilities
Uranium mines Phosphate Industry
18. DISTRIBUTION STATEMENT
EPA Form 2220-T (R.x. 4-77) PREVIOUS EDITION is OBSOLETE
19. SECURITY CLASS (ThisReport)
PURITY CLASS (Thli
Unclassified
419
lURITY CLASS IThffpage)
Unclassified
20. SECURITY CLA:
22. PRICE
------- |