United States
Environments! Protection
Agency
Office of
Radiation Programs
Washington, O.C. 20460
Radiation
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EPA 520/1-84-022-2
40 GFR Part 61
National Emission Standards
for Hazardous Air Pollutants
BACKGROUND INFORMATION DOCUMENT
(INTEGRATED RISK ASSESSMENT)
FINAL RULES FOR RADIONUCLIDES
VOLUME II
October 22, 1984
Office of Radiation Programs
U.S. Environmental Protection Agency
Washington, B.C. 20460
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CONTENTS
Page
1. INTRODUCTION 1-1
2. DEPARTMENT OF ENERGY FACILITIES
2.0 Introduction and Summary 2.0-1
References 2.0-9
2.1 Argonne National Laboratory 2.1-1
2.2 Brookhaven National Laboratory 2.2-1
2.3 Peed Materials Production Center 2.3-1
2.4 Fermi National Accelerator Laboratory 2.4-1
2.5 Hanford Reservation 2.5-1
2.6 Idaho National Engineering Laboratory 2.6-1
2.7 Lawrence Livermore National Laboratory 2.7-1
2.8 Los Alamos National Laboratory 2.8-1
2.9 Oak Ridge Reservation 2.9-1
2.10 Paducah Gaseous Diffusion Plant 2.10-1
2.11 Portsmouth Gaseous Diffusion Plant 2.11-1
2.12 Rocky Flats Plant 2.12-1
2.13 Savannah River Plant 2.13-1
2,14 Ames Laboratory 2.14-1
2.15 Bettis Atomic Power Laboratory 2.15-1
2.16 Knolls Atomic Power Laboratory 2.16-1
2.17 Lawrence Berkeley Laboratory 2.17-1
2.18 Mound Facility 2.18-1
2.19 Nevada Test Site 2.19-1
2.20 Pantex Plant 2.20-1
2.21 Pinellas Plant 2.21-1
2.22 Rockwell International Corporation 2.22-1
2.23 Sandia National Laboratories 2.23-1
2.24 Stanford Linear Accelerator Center 2.24-1
2.25 Reactive Metals, Inc. 2.25-1
2.26 Worldwide Impact of Selected Radionuclides 2,26-1
2.27 Future operations at DOE Facilities 2.27-1
2.27(A) Resumption of Operations at the PUREX Plant 2,27-1
2.27(B) Resumption of L-Reactor Operations at the
Savannah River Plant 2.27-3
111
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CONTENTS (Continued)
3. NRC-LICENSED FACILITIES AND NON-DOE FEDERAL FACILITIES
3.1 Research and Test Reactors
3.2 Accelerators
3.3 Radiopharmaceutical Industry
3.4 Radiation Source Manufacturers
3.5 Other NRG Licensees
3.6 Department of Defense Facilities
3.6A Armed Forces Radiobiology
Research Institute
3.6B U.S. Army Facilities
3.6C U.S. Navy Facilities
4. COAL-FIRED UTILITY AND INDUSTRIAL BOILERS
4.1 Utility Boilers
4.2 Industrial Boilers
5. URANIUM MINES
6. PHOSPHATE INDUSTRY FACILITIES
6.1 Phosphate Rock Processing Plants
6.2 Wet Process Fertilizer Plants
6.3 Elemental Phosphorus Plants
7. MINERAL EXTRACTION INDUSTRY FACILITIES
Metal Mines, Mills, and Smelters
7.1 Aluminum Industry
7,2 Copper Industry
7.3 Zinc Industry
7.4 Lead Industry
APPENDICES
Page
3.1-1
3.2-1
3.3-1
3.4-1
3.5-1
3,6-1
3.6A-1
3.6B-1
3.6C-1
4.0-1
4.1-1
4.2-1
5-1
6.1-1
6.2-1
6.3-1
7.1-1
7.2-1
7.3-1
7.4-1
A. ASSESSMENT METHODOLOGY
B. RADIONUCLIDE EMISSIONS TO AIR FROM FORMER MANHATTAN ENGINEER-
ING DISTRICT AND ATOMIC ENERGY COMMISSION SITES (FUSRAP)
C, RADON EMISSIONS FROM DEPARTMENT OF ENERGY- AND NUCLEAR
REGULATORY COMMISSION-LICENSED FACILITIES
D. DEPARTMENT OF ENERGY GOCO FACILITIES
A-l
B-l
C-l
D-l
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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)
final rules for sources of emissions of radionuclides pursuant to
Section 112 of the Clean Air Act.
This report 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) licensed(1) ancj non-DOS
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
emissions of radionuclides into air
(1^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.
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(3) A summary of emissions data
(4) Estimates of Che radiation doses and health risks to both
individuals and populations
Ends s i onj3_, Data
Insofar as possible, measured radiortuclide emission, data were used
to estimate health impacts. In the absence of measured data3 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 NRC annual effluent reports; and the data
for the other categories, such as coal-fired utility and industrial
boilers, uranium and ttonuranium mines, and the various extraction
industries, were usually obtained from special reports prepared under
contract with the EPA. Radon emissions from DOE- and NRC-licensed
facilities are considered separately in Appendix G of this volume.
Health Impact Assessment
The public health assessment includes estimates of the following
radiation exposures and health risks (see Chapter 8, Volume I, for more
detail);
(1) Dose-equivalent rates to the individuals at highest risk
(nearby individuals)
(2) Collective dose-equivalent rates to population groups
(3) Lifetime risks to nearby individuals its the exposed population
(4) The number of fatal cancers caused in the exposed, population
per year of facility operation
As s e ssm_ent Me thodo 1 ogy
DOE facilities were analyzed individually on a site-by-site
basis. Facilities in all of the other categories were grouped together
into source categories on the basis of similarity of activities or
operations and analyzed by defining a reference facility that
represents the source category. Doses were calculated using the
AIRDOS-EPA/BMTAB computer model developed by Oak Ridge National
Laboratory under contract to the EPA. These computations are based
upon current information on transport, uptake, and metabolic behavior
of the various radionuclides and are described in detail in two EPA
1-2
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reports (EPA79, Be8l). Appendix A of this volume contains a summary of
the parameters used for the AIRDQS-EPA calculations.
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. 20460.
Individuals
Dose-equivalent rates , radon concentrations, and radon decay
product exposures are presented for "nearby individuals." To select
the location for the nearby individuals, the lifetime risk for an
individual at offsite locations (at or beyond the perimeter of the
restricted area for each DOE facility) was calculated. Then, the
location providing the highest lifetime risk was selected and used for
assessing both the dose and risk for nearby individuals.
The dose equivalents presented for nearby individuals are 70-year
committed dose equivalents. This is also the dose-equivalent rate in
the 70th year following the start of exposure.
Radon decay product exposures presented for nearby individuals 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), unless otherwise indicated.
Re g iona 1 P opu I at i on
The term regional population refers to the population living
within a radius of 80 kilometers of a source. 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 are expressed in units of
person-rem/year and are the sum of the dose-equivalent rates for all
individuals considered in assessing releases from a source. Similarly,
collective radon decay product exposure rates are expressed in units of
person-working-levels. Further details of these calculations are
contained in Appendix A.
Lifetime Risk to Nearby Individuals and Number of Fatal Cancers
The lifetime risk to nearby individuals is the probability of
fatal cancer to an individual from a lifetime of exposure (70 years on
the average) to the concentrations of radionuclides estimated for that
individual.
1-3
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number of fatal cancers per year of operation is that
potential rtusnber of cancers in the population from one year s s release
of radlonuclides from the facility. These cancers are expected to
occur many years after the year in which the releases take place,
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""^ may be expressed as 1.23E-3; 3210, which is equal to
3,21 x 103s may be expressed as 3.21.E+3.
Metric system units have been used for reporting data, except in a
few instances where referenced data are presented in their original
customary units.
1-4
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REFERENCES
Be81 Begovich C. L., Eckerman K, F., Schlatter E, C., Ohr S. Y. 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,
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.
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Chapter 2; DEPARTMENT OF ENERGY FACILITIES
2-G rntroduct_ipn and Summary
The U.S. Department of Energy (DOE) operates diverse energy and
national defense programs involving research, development, and
production at a large nuraber of facilities located throughout the
United States, These facilities are owned (or leased) by the Federal
government and operated by contractors (so-called "GOCO"
facilities—Government Owned, Contractor Operated). DOE is granted
authority in the Atomic Energy Act of 1954* "to protect the public
health and safety from the operation of these facilities, including the
emission of radionuclides." This authority is implemented by
contractual agreements between DOE and its contractors. This obligates
the contractor to comply with all applicable health and safety
regulations and requirements of the Department of Energy.
As of 1980, there were 78 facilities in 24 statea which were
subject to DDE health and safety requirements (see Appendix D). This
report examin.es 25 of these facilities, selected on the basis of having
the most significant emissions as listed in the DOE Effluent
Information System,** Each of the 25 sites is described in terras of
its location, primary mission, major facilities that emit radio-
nuclides, and existing effluent control systems. The information in
these descriptions is taken from the annual environmental monitoring
reports prepared for each site. This information was supplemented,
when necessary, with information from environmental impact statements.
Airborne release data for all sites were obtained from the Department
of Energy's Effluent Information System and verified against the
airborne effluent data in the environmental monitoring reports. All
references are presented at the end of this section.
The worldwide impact of these facilities due to the emissions of
tritiutns carbon-14, krypton-85, and iodine-129 is also assessed.
Finally, the impacts of planned future operations at DOE facilities are
estimated.
*Section 161 of Public Law 83-703.
**Battelle-Colunibus and Shippingport Atomic Power Station, which were
included the draft report, have been, eliminated because they are no
longer DOE GOCO facilities.
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s ion
health and safety standards are contained in DOE Order 5430.1,
The requirements governing airborne releases of radionuclides to the
environment are expressed as concentration limits in air at the site
boundary. These concentrations can be related to emissions by
correcting for atmospheric dilution. A concentration limit, which is
established for each radionuclide by DOE, is the concentration of
radioactivity in air that causes a dose equivalent to the whole body,
gonadSj or bone marrow of 500 mrem per year, or 1500 mrem per year to
any other organ. In additions DOE requires that exposures to the
public be United to as small a fraction of the respective annual dose
limits as is reasonably achievable.
The radiation doses and risks of fatal cancers to individuals and
populations around Department of Energy facilities were estimated using
the methods discussed in Volume I. 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
entissionSj, 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
Principal
Radio-
nuclide
Ar-41
Kr-85
emissions
Quantity
(Ci/y)
0.4
6.7
Nearby individuals
Principal
organ
Pulmonary
Breast
Red marrow
Dose rate
( mrera/y )
<0.1
(Units of
10~6 >
0.0006 (0
.0002}
Brookhaven H-3 660 Pulmonary
National 0-15 36,000 Bone^3*
Laboratory Ar-41 170 Breast
Xe-127 2.3 Red marrow
Feed Materials D-234 0,11 Pulmonary
Production U-238 0.11 Bone^
Center Red Marrow
Kidneys
Fermi National H-3 0.42
Accelerator Oil 1500 Breast
Laboratory Red marrow
0.4
0.6
0.5
0.5
88(c>
26
1.8
12
0.7
0.6
0.7
3)
100 (100)
10
(4)
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
_ . , . . Radio-
Facility
nuclide
Quantity
(Ci/y)
Nearby individuals
Principal Dose rate
organ (rarem/y)
(Units of
Hanford Reservation
100 Area
200 Area
300-400 Areas
Idaho National
Engineering
Laboratory
Lawrence Liver-
more National
Laboratory
Los Alamos
National
Laboratory
(12 Technical
Areas)
Technical
Area 33
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
Oil
N-13
0-15
Ar-41
H-3
18
65,000
540
11,000
0.05
0.0004
450
400
2,500
59,000
0.055
2,600
170
170
1,100
130,000
25 , 000
200,000
1,400
6,100
Pulmonary
Bone(b)
Breast
Red marrow
Red marrow
Pulmonary
Pulmonary
Bone(b)
Pancreas
Pulmonary
Bone^b^
Thyroid
Breast
Red marrow
Red marrow
Kidneys
Intestine wall
Bone
Breast
Red marrow
2.2
2.4
2.2
2.2
< 0.1
< 0.1
1.4
1.5
1.4
< 0.1
< 0.1
0.12
< 0.1
< 0.1
1.3
1.3
1.6
11
9
11
0.5
0.7
0.7
40 (20)
0.2 (0.1)
30 (10)
0.5 (0.2)
30 (10)
200 (60)
10 (5)
1983 Emissions
TA-33
Oak Ridge
Reservation
_
H-3
Kr-85
1-131
Xe-133
U-234
-
11,000
6,600
0.6
32,000
0.12
Whole body
Pulmonary
Bone^b'
Thyroid
Kidneys
34
50(d)
7.6
9.3
5.4
800 (200)
100 (100)
See footnotes at end of table.
2.0-3
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Table 2-A. Summary of dose rates and risks to nearby individuals
for facilities with the largest emissions (Continued)
Facility
Paducah Gaseous
Diffusion
Plant
Principal
Radio-
nuclide
Tc-99
U-234
U-238
emissions
Quantity
(Ci/y)
0.006
0.01
0.04
Nearby individuals
Principal
organ
Pulmonary
Bone^b'
Thyroid
Kidneys
Dose rate
(mrem/y )
4.7(e)
7.1
0.2
3.6
(Units of
10 (10)
Portsmouth
Gaseous
Diffusion
Plant
Tc-99
Th-234
U-234
0.1
0.06
0,09
Pulmonary
Bone(b)
Thyroid
Kidneys
6.9(e)
11
2.0
5.1
20
(20)
Reactive
Metals, Inc.
Rocky Flats
Plant
Savannah River
Plant
U-234/5/8
H-3
U-234/5/8
Pu-239/40
H-3
Ar-41
Kr-85
Kr-88
Xe-133
1-131
1-129
0.00048
0.43
0.00003
0.000008
350,000
62,000
840,000
1,500
3,900
0.05
0.1.6
Pulmonary
Bone^b^
Kidneys
Intestine wall
Pulmonary
Bone'b->
Red marrow
Bone
Thyroid
Pulmonary
52 80 (80)
0.3
0.1
0.1
<0.1 0.02 (0.02)
<0.1
<0.1
2.3 40 (20)
4.9
2.2
(b)
(c)
(d)
(e)
location at point of highest dose equivalent. The risk estimates
in parentheses include a dose rate reduction factor of 2.5 for low-LET
radiations, as described in Chapter 8, Volume I, of this report.
Endosteal cells.
Lung clearance class for uranium: one-third D, one-third W,
one-third Y.
Lung clearance class for uranium: all uranium-238 and one-half
uranium-234, Y; one-half uranium-234, W.
Lung clearance class for uranium: all W.
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
Facility
Argonne
National
Laboratory
Brookhaven
National
Laboratory
Feed Materials
Production
Center
Fermi National
Accelerator
Laboratory
Hanford
Reservation
100 Area,
200 Area, and
300-400 Areas
Idaho National
Engineering
Laboratory
Lawrence Liver-
more National
Laboratory
Los Alamos
National
Laboratory
(12 Technical
Areas ) and
Principal
Radio-
nuc 1 ide
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
emissions
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 Fat*i can~
/ , \ cers/year of ^
organ (pers-rem/y) . Ca)
& r j operation
Pulmonary
Bone^b'
Breast
Red marrow
Pulmonary
Bone*-b^
Breast
Red marrow
Pulmonary
Bone^b^
Red Marrow
Kidneys
Bone(b)
Breast
Red marrow
Pulmonary
f i \ •*
| r\ j
Breast
Red marrow
Bone^b>
Thyroid
Red marrow
Pulmonary
Breast
Kidneys
Intestine wall
Red marrow
Bone(b)
Breast
Red marrow
<0.1 <0.001
<0.1
<0.1
<0.1
3 <0.001
3
3
3
440(c> 0.01 (O.OU
114
8
56
1.5 <0.001
1.2
1.4
11 0.003{<0.001)
13
10
11
0.3 <0.001
5.5
0.2
0.2
0.1
5.8 0.002 (<0.001)
7.5
5.6
63 0.01 (0.005)
53
61
Technical
Area 33
See footnotes at end of table.
2.0-5
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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
Biff us ion
Plant
Portsmouth
Gaseous Dif-
fusion Plant
Reac t ive
Metals, Inc.
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
U-234/5/8
H-3
U-234/5/8
Pu-239/240
H-3
Ar-41
Kr-85
Kr-88
Xe-133
1-131
1-129
11,000
6,600
0.6
32,000
0.12
0.006
0.01
0.04
0.1
0,06
0.09
0.00048
0.4
0.00003
0.000008
350,000
62,000
840 , 000
1,500
3,900
0.05
0.16
Regional population
T, • • i ^ «. Fatal can-
Principal Dose rate ,
/ , -^ cers/year of x
organ (pers-rem/y) , \a)
to operation
Pulmonary
Bone(b>
Thyroid
Kidneys
Kidneys
Pulmonary
Bone ^ b ^
Thyroid
Pulmonary
Bone^b'
Thyroid
Kidneys
Pulmonary
Bone(b)
Kidneys
Intestine wall
Pulmonary
Bone^b^
Red marrow
Pulmonary
Thyroid
212(d^ 0.008 (0.006)
22
15
15
6.7
3.4^^0.001
13
0.4
ll^e) <0.001
35
7.9
17
19.5 <0.001
0.1
0.1
0.04
0.1 0.001
0.2
0.01
103 0.03 (0,01)
120
£=.
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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
Bettis Atomic Power
Laboratory
Knolls Atomic Power
Lab. (Kesselring Site)
Knolls Atomic Power
Principal
organ
All organs
Pulmonary
(c)
Dose rate
(mrem/y)
0.001
0.004
0.08
(c)
Risk
(Units of
0.02
0.01
0.9
(c)
(a)
10"6)
(0.008)
(0.008)
(0.4)
Lab. (Knolls Site)
Knolls Atomic Power
Lab. (Windsor Site)
Lawrence Berkeley
Laboratory
Mound Facility
Nevada Test Site
Pantex Plant
Pinellas Plant
Rockwell International
Corp.
Thyroid
Pulmonary
Intestine
wall
0.003 0.04 (0.02)
1.6 9 (4)
0.2 4 (1)
0.002 0.03 (0.01)
0.005 0.008 (0.007)
0.3 5 (2)
0.00004 0.00006 (0.00002)
Sandia Laboratories Bone
Stanford Linear Bone^b'
Accelerator Center
0.0009
0.006
0.01
0.1
(0.006)
(0.04)
'a'Qff-site location at point of highest dose equivalent. The risk
estimates in parentheses include a dose rate reduction factor of 2.5
for low-LET radiations, as described in Chapter 8, Volume I, of this
report.
(b^Endosteal cells.
^'Kesselring and Knolls sites were assessed as a single combined site.
2.0-7
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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
Bettis Atomic Power
Principal
organ
Average all
Average all
Dose rate
( person-rem/y)
organs 0.004
organs 0.01
Fatal can-
cers/year of
operation^3'
<0.0001
< 0.0001
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
Stanford Linear
Accelerator Center
Average all organs 0.1
(b) (b)
Average all organs 0.001
Average all organs 0.7
Average all organs 8.9
Average all organs 0.001
Average all organs 0.0006
Average all organs 0.9
Average all organs 0.0001
<0.0001
(b)
<0.0001
0.0002 (0.00008)
0.003 (0.001)
<0.0001
<0.0001
0.0002 (0.0001)
<0.0001
Average all organs 0.003 <0.0001
) 0.03 <0.0001
cancers committed per year of operation of the facility. The
risk estimates in parentheses include a dose rate reduction factor of
2.5 for low-LET radiations, as described in Chapter 8, Volume I, of
this report.
ing and Knolls sites were assessed as a single combined site.
'c'Endosteal cells.
2.0-8
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REFERENCES
Department of Energy, Effluent Information System, Department of
Energy, Washington. B.C., 1981,
Argonne National Laboratory
Golchert N, W. , Duffy T. L. arid Sedlet J. , Environmental Monitoring at
Argonne National Laboratory — -Annual Report for 1981 (ANL-82-12), March
1982.
Moore E. B. , Control Technology for Radioactive Emissions to the
Atmosphere at U.S. Department of Energy Facilities, PNL-4621, Pacific
Northwest Laboratory, October 1.984.
Brjookhaven Nat ipnal Laboratory
Naidu J. R. and Qlmer L. L. , Editors, 1981 Environmental Monitoring
Report, Drookhaven National Laboratory, Safety and Environmental
Protection Division, April 1982.
Energy Research and Development Administration, Environmental
Statement, Brookhaven National Laboratory, Upton, New York, ERDA-1540.
Feed Materials Production Center
Fleming D. A., et al., Feed Materials Production Center, Environmental
Monitoring Annual Report for 1981, NLCO-1180, NLO, Inc., Cincinnati,
Ohio, May 1982.
Fermi Na t ional ^A^ce_lerator _Lab_or_atgry_
Dave M. J. and Charboneau R. , Baseline Air Quality Study at "Fermilab,
ANL Report, ANL/EES-TM-110, 1980.
Universities Research Association., Inc., Environmental Monitoring
Report for Fermi National Accelerator Laboratory, Annual Report for CY
1981, FERMILAB 8/22, Universities Research Association, Inc., Batavia,
Illinois, May 1982.
U.S. Atomic Energy Conmissions Environmental Statement, National
Accelerator Laboratory, Batavia, Illinois, WASH- 15 05, Washington, B.C.,
1971.
2.0-9
-------
REFERENCES (Continued)
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.
Energy Research and Development Administration, Final Environmental
Impact Statement, High Performance Fuel Laboratory, Hanford
Reservation, Richland, Washington, ERDA-155G, UC-2, 11, Washington, D.C. ,
1977.
Sula M. J. , McCormack W. D. , Dirkes R. L. , Price K. R. , Eddy P. A.,
Environmental Surveillance at Hanford for CY-81, PNL-4211, May 1982.
Idaho Na tional Enginner ing Laboratory
Department of Energy, Environmental Monitoring Program Report for Idaho
National Engineering Laboratory Site, IDO-12082 (81), Idaho Operations
Office, May 1982.
Energy Research and Development Administration, Final Environmental
Impact Statement, Waste Management Operations, Idaho National
Engineering Laboratory, ERDA-1536, Washington, B.C., September 1977.
Lawrence Livermore National Laboratory
Auyond M. , Griggs K. S., and Buddemeier R. W., Environmental Monitoring
at the Lawrence Livermore National Laboratory--198i Annual Report,
UCRL-50Q27-81, University of California, Livermore, California, March
1982.
Department of Energy, Environmental Impact Statement: Livermore Site,
Livermore, California, DOE/E IS-0028, September 1978.
Environmental Surveillance at Los Alamos During 1981, Los Alamos
National Laboratory Report, LA-9349-ENV (UC-41), April 1982.
Department of Energy, Environmental Impact Statement, Los Alamos
Scientific Laboratory Site, Los Alamos, New Mexico, DOE/E IS-0018 ,
Washington, B.C., December 1979.
Los Alamos National Laboratory, Environmental Surveillance at Los Alamos
During 1983, Los Alamos National Laboratory Report, LA-1010Q-ENV( UC-41) ,
April 1984.
2.0-10
-------
REFERENCES (Continued)
Union Carbide Corporation, Environmental Monitoring Report, Department
of Energy Oak Ridge Facilities, Calendar Year 1981, Y/UB-16, Oak Ridge,
Tennessee, May 1982.
Atomic Energy Commission, Environmental Statement: Radioactive Waste
Facilities: Oak Ridge National Laboratory, Oak Ridge, Tennessee,
WASH-1532, August 1974.
Energy Research and Development Administration, Final Environmental
Impact Statement; Management of Intermediate-Level Radioactive Waste:
Oak Ridge National Laboratory, Oak Ridge, Tennessee, ERDA-1553,
September 1977.
Pa d u c ah Gaseous Pi f f u si on PI an t
Union Carbide Corporation, Environmental Monitoring Report, Department
of Energy, Paducah Gaseous Diffusion Plant, Paducah, Kentucky, May 1982.
Atomic Energy Commission, AEC Gaseous Diffusion Plant Operations,
ORO-684, January 1972.
Portsmouth Gaseous Pi f fusion Plant
Acox T. A., et al., Portsmouth Gaseous Diffusion Plant Environmental
Monitoring Report for Calendar Year 1981, Goodyear Atomic Corporation,
Piketon, Ohio, April 1982.
U.S. Energy Research and Development Administration, Final
Environmental Impact Statement 4 Portsmouth Gaseous Diffusion Plant
Expansion, Piketon, Ohio, ERDA-155S, September 1977.
Rocky _FlaCs _Plant
Annual Environmental Monitoring Report, U.S. Department of Energy,
Rocky Flats Plant, January through December 1981, RFP-ENV-81, Rockwell
International, Energy Systems Group, April 1982.
U.S. Energy Research and Development Administration, Final
Environmental Impact Statement, Rocky Flats Plant Site, Golden,
Colorado, Vol. 1, ERDA 1545-1, Washington, B.C.
2.0-11
-------
REFERENCES (Continued)
Department of Energy, Environmental Monitoring in Che Vicinity of the
Savannah River Plant, Annual Report for 1981, DPSPU-8 2-30-1, E.I. du
Pont de Nemours and Company , Aiken, S. C. , 1981.
U.S. Energy Research and Development Administration, Final
Environmental Impact Statement, Waste Management Operations, Savannah
River Plant, Aiken, S. C. , ERDA-1537, Washington, D.C. , 1977.
Voss M. D. , Environmental Monitoring Summary for Ames Laboratory,
Calendar Year 1981, IS-4798, Ames Laboratory, Ames, Iowa, April 1982,
Voss M. D., Environmental Monitoring at Ames Laboratory, Calendar Year
1976, IS-4139, Energy Research and Development Administration, Ames
Laboratory, Ames, Iowa, 1977.
Bettis Atomic Power Laboratory
Annual Effluent and Environmental Monitoring Report for Calendar Year
1981, Bettis Atomic Power Laboratory, WAPD-RD/E(ESE)-5763 April 1982.
Kno 11 s At omi c ...... Power_Labora_tg_ry_
Knolls Atomic Power Laboratory, Environmental Monitoring Report,
Calendar Year 1981, KAPL-4148, General Electric Corporation, May 1982.
Lawrence Berkeley Laboratory
Lawrence Berkeley Laboratory, Annual Environmental Monitoring Report of
the Lawrence Berkeley Laboratory, 1981, LBL-14553, University of
California, Berkeley, California, June 1982.
Mound Facility
Farmer B. M. and Carfagno D. G. , Annual Environmental Monitoring
Report: Calendar Year 1981, MLM-2930, Monsanto Research Corporation,
Mound Facility, Miami sburg, Ohio, April 1982.
U.S. Department of Energy, Final Environmental Impact Statement, Mound
Facility, Miamisburg, Ohio, DOE/EIS-0014, Washington, D.C. , June 1978.
2.0-12
-------
REFERENCES (Continued)
Nevada Test Site
Black S. C., et al., OffsiCe Environmental Monitoring Report, Radiation
Monitoring Around U.S. Nuclear Test Areas, Calendar Year 1981.
Prepared for the U.S. Department of Energy by the EPA Environmental
Monitoring Systems Laboratory, EPA-6QO/4-DOE/DP/00539.
U.S. Energy Research and Development Administration, Final
Environmental Statement: Nevada Test Site, Nye County, Nevada,
ERDA-1551, September 1977.
Pantex Plant
Environmental Monitoring Report for Pantex Plant Covering 1981,
MHSMP-82-14, 1982.
Pinellas Plant
Pinellas Plant Environmental Monitoring Report, 1981, GEPP-EM-654,
General Electric Compnay, March 1982.
Rockwell International Corporation
Energy Systems Group, Environmental Monitoring and Facility Effluent
Annual Report 1981, ESG-82-21, Rockwell International, Canoga Park,
California, July 1982.
Sandia National Laboratories
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
Millard G. C., et al., 1981 Environmental Monitoring Report, Sandia
National Laboratories, SAND 82-0833,, April 1982.
Stanford Linear Accelerator Center
Annual Environmental Monitoring Report for Stanford Linear Accelerator
Center, January-December 1981, SLAC-249, Stanford University, Stanford,
California, May 1982.
2.0-13
-------
REFERENCES (Continued)
PUREXJPlant
Department of Energy, Draft Environmental Impact Statement, Operation
of PUREX and Uranium Oxide Plant Facilities, DOE/EIS-OQ89D, 1982.
L—Reactor Operation
Department of Energy, Environmental Assessment, L-Reactor Operation,
Savannah River Plant, DOE/EA-0195, 1982.
2,0-14
-------
2 • 1 • 1 general Descr lotion
Argonrie national Laboratory (ANL) occupies the central 6,88 kra^
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, biomedlcal, 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 Description of Facility
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 Rrgonne 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
(IPMS, 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
raulticurie quantities of the actinide elements. Two major facilities,
a 12.5 GeV proton accelerator (ZGS, the Zero Gradient Synchrotron) and
a 5 Mtf heavy water- enriched uranium reactor (CP-5) were not in
operation during 1981 and are awaiting decontamination and
decommissioning.
2,1.3. Radionuc 1 ide__Emiss ions and
Airborne emissions from Rrgonne Mational Laboratory for 1981 are
identified in Table 2,1-1. The emissions for years 1.979 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
HEPA filters and room temperature charcoal traps (Mo83). Calculations
of health impact were based on a single release point (stack height of
61 meters).
2.1.4 Health Impact As sessment o f Argonne National Laboratory
The estimated annual radiation dose rates from radionuclide
emissions in 1981 from Argonne National Laboratory are shown in Table
2.1-3. The nearby individuals are 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 fatal cancer from exposure to the radioactive emissions
from this facility are identified in Table 2.1-4. The risk estimates
include estimates which use a dose rate effectiveness factor of 2,5, as
described in Chapter 8, Volume I.
Table 2.1-1. Radionuclide emissions from Argonne National Laboratory,
1981
. . 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
-------
Table 2.1-2. Radionuclide emissions from Argonne National Laboratory,
1979 to 1981 (Cl/y)
Radionuclide 1979 i960 1981
Antimony- 125
Argon- 41
Tritium
Krypton-79
Krypton-85
Krypton~85ra
Xenon- 133
Xenon- 135
8.5E-5
7.1E+3
6.6B42
1.5E-4
9.0
1.4B-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 , 2B- 5
1 . 7B-5
3.8E-1
6.9E-7
0
6.7
0
0
0
Table 2.1-3. Radiation dose rates from radionuclide emissions
£rora Rrgonne National Laboratory
o Nearby individuals Regional population
(mrem/y) (person-rem/y)
Bndosteum
Red Marrow
Breast
Pulmonary
3 . 3E-5
3 . 1E-5
3.1B-5
3.1E-5
2.6E-3
2.5E-3
2.4E-3 .
2.5E-3
Table 2.1-4. Fatal cancer risks due to radionuclide emissions from
Argonne Mational Laboratory'3'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
ANL 6E-10 (2E-10) 6B-7 (3E-7)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.1-3
-------
Page Intentionally Blank
-------
2.2 BrookhavenNat ional_ Laboratory; Long Island, Mew York
2.2.1 Genera1 JDescript ion
Brookhaven National Laboratory (BNL) Is a large scientific
research facility located near the center of Long Island approximately
97 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 De_scr.lp_tion _ofFacility
a wide variety of scientific programs are conducted at
Brookhaven. The major facilities at the laboratory include several
accelerators, reactors, and groups of laboratories. The facilities
that release radioactivity to the atmosphere are described briefly
below.
The High-Flux Beam Reactor (HFBR) is a 60-HWCt), fully enriched,
heavy-water-moderateds -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/era^. 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, D20
decomposition products, D2® vapor, and tritiated heavy water vapor
(DTO).
The Alternating Gradient Synchrotron (RGS) is a 33-GeV proton
accelerator used for ultra-high energy particle physics research.
Protons originate in a 0.75 MeV Cockcroft-tfalton generator, are
accelerated by a 200-HeV 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 20Q-MeV linac also serves the Brookhaven Linac Isotope
Production Facility (BLIP) and the Chemistry Linac Irradiation Facility
(CLIP), 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
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. CLIP targets are shuttled into the beam via a
pneumatic target transfer system.
2.2-1
-------
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 (MRR) is a tank-type,
fully enriched, water-cooled and -moderated reactor. It operates
intermittently at power levels up to 3 MM (thermal) at a pressure of
246 g/em^. it is an integral part of the Medical Research Center and
is used for various research programs requiring irradiation facilities,
An intersecting storage ring accelerator, "ISftBELLE," is currently
under construction and will be operational sometime in the 1980s,
ISftBELLE will be a colliding beam machine, in which the collision of
two proton beams of 400 GeV will make available effective energies up
to 800 GeV. The machine will be used to conduct advanced studies in
high energy physics.
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 Radipnuc1Ide Emis s ions
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 (tj_/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 31 percent of the tritium and all xenon--127 and
2.2-2
-------
unidentified beta, gamma-ray releases. Thus, only small quantities oC
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
HF8R - Hot Laboratory
Hazardous Waste Management Area
MRC, Building-490
MRR, Building- 491
BLIP, Building-931
Height
On)
11
18
98
10
14
46
18
Table 2.2-2. Radionuclide emissions (Ci) from
Brookhaven National Laboratory by stack number, 1981
KaulO
nuclide s-1
Tritium 4.3
Beryllium- 7
Carbon- 14
Oxygen- 15
Phosphorus-32
Sulfur-35
Argon- 41
Iron- 59
Tin-113
iodine -125
Xenon- 127
Unid. beta, gamma
Stack Number
S-2 S~3 S-4 S-5
4.1E+2 2.4E+2 1.8E-1 1.1
2.6S-3
8. IE- 4
1.5B-4
5.7E-3
2.5E-4
2.6E-4
9 . 9E- 4
2.3
1.8B-4
S-6 S-7
6.6B-2
3.6B+4
1.7S+2
(a)
See Table 2.2-1 for stack identification.
2.2-3
-------
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), the Brookhaven Facility was modeled with 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).
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 1979 1980 1981
Argon-41
Beryllium- 7
Carbon-14
Iodine-125
Iron-59
Oxygen-15
Phosphorus-32
Sulfur-35
Tin-113
Tritium
Unidentified
beta + gamma
Xenon- 12 7
3.2E + 2
NR
NR
NR
NR
2.8E+4
NR
NR
NR
2.3E+2
1.7E-4
1.0
2.6E+2
NR
NR
NR
NR
2 . 5E +4
NR
NR
NR
5.5E+2
7.8E-5
1.6
1.7E + 2
2.6E-3
8. IE -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
MR None reported.
2.2.4 Health Impact Assessment ofBrogkhaven National Laboratory
The estimated annual radiation doses for this facility are
summarized in Table 2.2-4. The assessment was based on all emissions
in 1981 being combined into one point source. The nearby individuals
are located 1300 meters north-northwest from the hypothetical 18 m
stack. The population within the 80 km radius assessment area is about
4.6 million.
2.2-4
-------
The majority of the dose was due to oxygen-15 through the air
immersion pathway. The exposure to the regional population was
primarily due to tritium and oxygen-15.
The risks of fatal cancer as a result of exposure to the
radioactive emissions from this facility are listed in Table 2.2-5.
The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I.
Table 2.2-4. Radiation dose rates from radionuclide emissions
from Brookhaven National Laboratory, 1981
Nearby individuals Regional population
rgan (mrem/y) (person-rem/y)
Pulmonary
Red marrow
Breast
Liver
Endosteum
4.4E-1
5.4E-1
4.7E-1
4.1E-1
5.6E-1
3.1
3.3
3.2
3.0
2.9
Table 2.2-5. Fatal cancer risks due to radionuclide emissions from
Brookhaven National Laboratory, 1981^a'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
BNL 8E-6 (3E-6) 9E-4 (4E-4)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.2-5
-------
Page Intentionally Blank
-------
2 , 3
2.3.
The Feed Materials Production Center (FMPC) , operated by NLO,
inc., 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 SO
kilometer radius of FMPC is 2,6 million.
2.3.2 Description ....... of Facility
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 tetraf luoride. 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. The processes associated with each of the eight buildings
are as follows;
Plant 1
Plant 2
Plant 4
Plant 5
Plant 6
Plant 8
Plant 9
Pilot Plant
Material sampling and grinding;
Dry feeds digestion;
Uranium tetrafluoride production and
repackaging;
Metal production and slag grinding;
Metal machining;
Dumping and milling,*
Metal production, remelting, and machining;
Uranium and thorium metal and compound
production.
2.3-1
-------
2.3.3 EMiofiu^Llj^
Table 2.3-1 summarizes the radlonucllde emissions from FKPC in
1981 for each of the eight stacks and an on-site incinerator. Ex-
hausted air from these buildings is passed through scrubbers or cloth
type bag filters prior to release to building stacks. Only natural
uranium was released during 1981; no thorium was released during the
year.
Table 2.3-1. Radionucllde emissions from
Feed Materials Production Center, 1981 (Ci/y)
Source Uranium emissions (Ci/y)
Uranium-238 Uranium-234
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
3.3E-4
0.
6.26E-2
4.46E-2
0.
5.33E-3
0,
0
4.15E-4
Total 0.113 0.113
2-3.4 Health Impact assessmentof FMPC
For the health impact assessment, all releases were assumed to
originate from a single 10-meter stack at the center of the production
area. 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~234Hi. Uranium emissions are assumed to be one-third
Class Y, one-third class w, and one-third class D.
The estimated annual radiation doses from radionuclide emissions
from FMPC are shown in Table 2.3-2. The estimates of regional
population dose are for a regional population of 2,6 million. The
nearby individuals are located 810 m northeast of the release point at
the site boundary. The major pathway of exposure is inhalation,
the critical organ is the pulmonary.
2.3~2
-------
The estimated individual lifetime risk and the number of fatal
cancers per year of operation are shown in Table 2.3-3. The risk
estimates include estimates which use a dose rate effectiveness factor
of 2,5, as described in Chapter 8, Volume I.
Table 2.3-2. Radiation dose rates from radionuclide
emissions from the Feed Materials Production Center
„ Nearby individuals Regional population
Organ t i \ t / ^
(.rarem/y) (.per son-rein/y;
Pulmonary
Endosteum
Kidney
Red marrow
88
26
12
1.8
436
114
56
8
Table 2.3-3. Fatal cancer risks due to radionuclide
emissions from the Feed Materials Production Center
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
MFC 1E-4 1E-2
2.3-3
-------
Page Intentionally Blank
-------
2.4 FerRd_j|atio^
2.4.1 General Pescri^tion
The Fermi Mational Accelerator Laboratory (PMAL) 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 the university Research Associates,
Inc. The facility is composed of three basic elements: the
accelerator, experimental areas, and support facilities,
The primary purpose of FNftL is fundamental research in high energy
physics. In addition, cancer patients are treated using neutrons
released by the interaction of 66 MeV protons from the second stage of
the accelerator, ft 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 residen-
tial use. There are many municipalities in the vicinity, resulting in
a distinct pattern of high population concentration. Within a 3-km dis-
tance 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 FMAL is more than 7.5 million.
2.4.2 pescription of Fac il i ty_
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 10^.
2.4.3 Radlonuclide Emissions and Existing control Technology
Activation of air in measurable concentrations occurs 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, activation of air is usually
caused by secondary particles.
Radioactive gas, primarily carbon-11, is produced by interaction
of secondary particles with air. Monitoring is carried out by
2.4-1
-------
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 failed magnets by decomposing the epoxy adhesives at high
temperatures. Most of these magnets are radioactive. Thirty magnets
were debonded in 1981, and the total tritium release was approximately
5 mCi. Table 2,4-1 lists the activity, location, and stack heights of
the FNAL airborne releases for 1981. Table 2,4-2 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 are small during
acceleration, extraction, and transport to the experimental-area
targets.
The accelerator, beam-transport, and target systems are all within
well-shielded housingss while the beam travels in evacuated pipes, thus
reducing the activation of air.
2.4.4 Health Impact Assessment of Fermi Laboratories
The estimated annual radiation doses resulting from radionuclide
emissions in 1981 from the Fermi Laboratories are listed in Table
2,4-3. Nearby individuals are located 1300 meters north of the release
location. The predominant exposure pathway is that of air immersion.
The dose is primarily (greater than 99 percent) from carbon-11.
Table 2.4-4 lists the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional population
from these doses. The risk estimates include estimates which use a
dose rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I.
Table 2,4-1, Radionuclide emissions from
Fermi National Accelerator Laboratory, 1981
Source
Neutrino Area
Meson Area
Debonding oven
Radionuclide
Carbon-11
Tritium
Tritium
Emissions
(Ci)
1.5E+3
4.2E-1
5.0E-3
height = 10 meters.
2.4-2
-------
Table 2.4-2. Radionuclide emissions from
Fermi National Accelerator Laboratory, 1979 to 1981
(Ci/y)
Radionuclide 1979 1980 1981
Carbon-11
Tritium
4.QE+3
2.8E-1
1.3E+3
2.4E-1
1.5E+3
4.2E-1
Table 2.4-3. Radiation dose rates from radionuclide emissions from
Fermi National Accelerator Laboratory, 1981
Nearby individuals Regional population
Organ t i \ ( / i
0 (mrem/y ) (.person-rem/y)
Red marrow
Endosteum
Breast
6.7E-1
6.9E-1
5 . 8E -1
1.4
1.5
1.2
Table 2.4-4. Fatal cancer risks due to radionuclide
emissions from Fermi National Accelerator Laboratory,
Lifetime risk to Regional population
Source nearby individuals (Fatal cancers/y of operation)
FNAL 1E-5 (4E-6) 3E-4 (1E-4)
'a'Ihe risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2,4-3
-------
Page Intentionally Blank
-------
2.5.1.
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 11.3 kilometers from the river in the 200 Area.
The 300 Area, just north of the city of Richiand, contains the reactor
fuel manufacturing facilities and research and development
laboratories. The Past Flux Test Facility (FFTF) is located in the
400 Area approximately 8.8 kilometers northwest of the 300 Area, and
the Washington Public Power supply System (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 at the Hanford Reservation 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, fuel processing, and all site support
facilities
- UNC Nuclear Industries (UNC): N-reactor operation and
fuel fabrication
- Battelle's Pacific Northwest Laboratory (PNL): research in
biophysics and biomedicine and development of advanced waste
management technologies
- tfestinghouse Hanford Company {WHC): operation of Hanford
Engineering Development Laboratory (HEDL), including advanced
reactor development (principally the Liquid Metal Fast Breeder
Reactor Program).
2.5-1
-------
Hanforct Reservation 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.
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.
jLOO_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
Rrea include the N-Reactor and the 1706 Laboratory.
The N-Reactor is operated by UNC 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 M-Reactor operations,
Small-scale decontamination studies are also done at the laboratory.
200,Area
The 200 Rrea 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
200 East Rrea. Activities conducted in the area include waste
treatment and storage, equipment decontamination, plutonium and uranium
processing, and laboratory research.
2.5-2
-------
The Plant, in the 2QQ Area, is the
facility at Hartford. 1972 the has
held in is to operation no later
Rpril 1984 and continue through the year 2000. See Section 2.2T for a
discussion of the future operations of DOE facilities.
toother 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 plutoniura 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 flJR ami CR
vaults, and the numerous tank farms.
Major facilities in the 200 ¥est Area include the 1)03 plant, the
Z-Plant, and the Redox Plant. Uranyl nitrate hexahydrate solution
(UNH) is converted to 003 at the 003 Plant. The z-Plant has been
used to finish the processing of plutonium separated during the POREX
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 plutoniuis 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 PNL; facility tank farms; the 242T
waste evaporator; and the laundry facility.
300_ftrea
The 300 Rrea, 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 iciLoaeters north of Kichland and adjacent to
the east site boundary. The msjor facilities are the Hanford
Engineering Development Laboratory (HSDL), 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-Heactor. The Life sciences
Laboratory is operated by PNL; current programs include biophysical and
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, PNL operates two laboratories that conduct
2,5-3
-------
research in waste management techniques 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,
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,
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 Past 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.
2,5.3 RadionuclidejEmissions and Existing Control Technology
The airborne releases at Hanford Reservation are presented in
Table 2,5-1. 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 fron the northern site boundary at the location of
N~reactor. The 200 Area stack is 10 kilometers from the southern site
boundary anci 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.
All particulates released from Hanford operations are assumed to
be 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 front the M-Reactor are passed through KEPR 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
PURBX Plant account for the majority of the plutonium released in the
area. Airborne effluents from all 200 Area release points are passed
through acid scrubbers, deentrainers, fiberglass filters, and HBPA
filters prior to release. In addition, releases from the PUREX 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.
2.5-4
-------
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 HSPA 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 E-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.
2.5.4 Health Impact Assessment oftheHanfprd 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 2.5-2 and Table 2,5-3. The risk estimates in
Table 2.5-3 include estimates which utilize a dose rate effectiveness
factor of 2.5, as described in Chapter 8, volume 1. 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
nearby individuals for the 100 Area are 7500 m northwest of the source.
For the 200 Area, the nearby individuals are 16,000 m south of the
release point. The nearby individuals for the 300-400 Area are also
south of the facility, at a distance of 2000 meters.
2.5-5
-------
Table 2.5-1. Radionuclide emissions from the Hanford Reservation, 1981
Emissions (ci/y)
Had ionucl ides
Argon- 41
Arsenic-76
Carbon- 14
Barium-Lanthanum- 140
Cerium- 144
Cobalt-58
Cobalt -60
Cesiura-137
Cesium- 138
Europium- 154
Europium-155
Iron- 59
Tritium
iodine- 131
Iodine- 132
Iodine- 133
iodine- 135
Krypton- 85m
Krypton-87
Krypton-88
Manganese- 54
Manganese-56
Sodium- 2 4
Plutonium-239
Ruthenium- 103
Ruthenium-Rhodium- 106
Strontium- 89
Strontium-90
Strontium-91
Mo lybdenura- Techne t ium- 99m
Uranium-234
Uranium- 238
Xenon- 135
100 Area 200 Area 300-400 Area
6 . 5E+4
2.3E-2
3.2 4.5E-7
1.1E-1
7.9E-2
6.6E-3
1.6E-2 3.3E-7
8.9E-3 5.0E-2
1 . 1E+4
1.5E-1
2.5E-2
2.7E-3
1.8E+1
9.7E-2 3.0E-4
4.3
9.4E-1
1.6
2.5E+2
2.8E42
5.4E-«-2 4.5E-*-2
2.8E-3
4.6E-1
1.2E-1
6.4E-5 3.7E-4 2.2E-5
3.3E-3
4.2E-3
1.5E-3
4.8E-3 3.1E-3 8.8E-5
1.8E-1
2.5E-1
7.5E-5
7.5E-5
4.6E+2
2.5- 6
-------
Table 2,5-2. Radiation dose rates from radionuclide emissions from
the Hanford Reservation, 1981
Nearby individuals (mrem/y)
Orgsn ——••—— •*- — —— •* —
100 Area 200 Area(a) 300-400 Area
Red marrow
Endosteum
Pulmonary
Pancreas
Breast
2.2
2.4
2.2
2.1
2.2
2.0E-2
8.4E-2 (8.3E-2)
2.1E-2
9.2E-3
1.1E-2
1.2
1.5
1.4
1.4
1.3
Regional population (person-rem/y)
100 Area 200 Area^a) 300-400 Area
Red marrow
Endosteum
Pulmonary
Thyroid
Breast
Pancreas
6.9
8.8
5.9
5.8
6.2
5.5
3.8E-1
1.1
2.3E-1
1.8E-1
2.0E-1
2.0E-1
(3.6E-1)
(1.0)
(1.6E-1)
(1.9E-1)
(1.9E-1)
3.8
4.7
4.5
4.1
4.0
4.3
dose rates in parentheses are based on NRPB Publication R129;
see Chapter 7, Volume I, of this report.
Table 2.5-3. Fatal cancer risks due to radionuclide emissions from
the Hanford Reservation,
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
100 Area
200 Area
300-400 Area
4E-5 (2E-5)
2E-7 (1E-7)
3E-5 (1E-5)
2E-3 (7E-4)
6E-5 (2E-5)
1E-3 (5E-4)
'a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
¥olurae I, of this report.
2.5-7
-------
Page Intentionally Blank
-------
2.6 Idaho S§tiopal Engineering. .Laboratorf;^,,,Mi3sgg,L.__SQake River Plain
2.6,1
The Idaho National Engineering Laboratory (INEL) is a large R&D
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
3? klloraeters northwest of Blackfoot. As o£ 1981, 52 reactors had been
built, 17 of which were still operating or operable.
Current programs at INEL are conducted at various areas of the
site and are managed for DOE by four contractors: EG&G Idaho, Inc.;
Exxon Muclear Idaho company, Inc.; Argonne National Laboratory; and
aestinghouse 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, located in the Test Reactor Area (TRA); the Technical Support
Facility (TSF), located in the Test Area North (TAN); and the
Loss-oE--Fluid Test Facility, located in the Test Area North
-------
equipment components under conditions of high neutron flux. This
research is in support of DOE'S reactor developnent 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~l (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.
fhe 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.
ExxonNuclear Idaho Company, 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
fhe Argonne National Laboratory-west (ANL-W) currently has five
operational complexes: the Experimental Breeder Reactor Ho. 2
(EBR-II), the Transient Reactor Test Facility (TREAT), the Zero Power
Facilities (55PR-6, ZPR-9, and ZPPR), the Hot Fuels Examination Facility
(HKKK), and the Laboratory and Office (L&o) support complex. All of
these complexes provide support services for DOE'S Fast Breeder Reactor
(FBK) 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
2.6-2
-------
cores. The prototype reactors are also used as training centers for
naval reactor operators. Three operating reactors and the Expended
Core Facility (BCF) are located in this area. These include the Large
Ship Reactors (A.1V), the submarine Thermal Reactor CsiV), and the
Natural Circulation Reactor (S5G).
2.6,3 RadionucIide Eaissions _and Existing Contro1 Technology
Measurements of airborne releases at INEL have been consolidated
and are presented in Table 2.6-1. The majority of emissions are
attributable to the operation of the ATR and the ETR (dismantled in
1981) in the Test Reactor Area. These releases include argon-41, a
majority of reported isotopes of xenon, cesium-138, barium-139,
krypton-85, krypton-85ra, krypton-87, and rubidium-88. TREAT accounts
for the xenon~133 emissions, and activities at ICPP are responsible for
exhausting tritium and krypton-85. SBR-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 HBPA filters and,
occasionally, charcoal absorbers. Areas using such control
technologies include the zero Power Facilities, TREAT, MRP facilities,
PBP, and ARA-1.
2.6.4 Health, impact assessment of Idaho National EngineerinjLj^abp_ratQ.r.y
For the purpose of the dose/health effects assessment, it is
assumed that all particulates released from the site are respirable.
The assessment is based on all emissions being combined into one point
source midway between the TRA and ICPP 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 2.6-2. The nearby
individuals are located 19500 HI north of the assumed release point.
Air immersion is the major pathway contributing to the individual dose
equivalent rate.
The fatal cancer risks are given in fable 2.6-3. The risk
estimates include estimates which use a dose rate effectiveness factor
of 2.5, as described in chapter B, Volume I. The pathway with the
highest contribution to the fatal cancer risk is ingestion.
2.6-3
-------
Table 2.6-1. Radionuclide emissions from the Idaho
National Engineering Laboratory, 1981
Radionuclide
Silver- 110m
Argon- 41
Barium-131
Barium- 139
Ba r ium~ Lan t hanum- 140
Beryllium--?
Bromine- 82
Carbon- 14
Cerium~141
Cerium- 144
Cobalt- 57
Cobalt-58
Cobalt-60
Cesium- 134
Cesium-137
Cesium- 138
Chromium- 51
Bur opium- 152
Europium- 154
Europium- 155
Tritium-3
Hafnium- 181
Iodine- 129
Iodine- 131
Iodine- 133
Krypton- 85
Krypton- 85m
Krypton- 81
Krypton-Rubidium- 88
Manganese-~54
Niobium- 95
Promethium-144
Plutonium- 238
Plutonium- 239
Ruthenium- 103
Ruthenium-Rhodium- 106
Antimony- 122
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
8.6B-3
1.7E+1
2.8S-5
6.0E-7
7.7E-6
1.5E-6
4.0E+2
1.1E-5
3.7B-2
5.5E-2
2.0E-6
5,9E+4
2.2E+2
8.7E+2
8 . OE+2
7 . 3E-6
2.5E-5
3.7E-4
7.4E-5
1 . 8B- 5
1 . 4E-6
7.7E-2
1 . 2E--7
2.6-4
-------
Table 2.6-1, Radionuclide emissions from the Idaho
National Engineering Laboratory, 1981 (continued)
„ ,. , . , Emissions
RadlonucIide ,„. , ^
Antimony-125 1.9E-1
Strontium-90 4.1E-3
Tantalum-182 L.9E-7
Telluriuin-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
Zirconiura-95 1.9E-6
Table 2.6-2. Radiation dose rates from radionuclide emissions
from the Idaho National Engineering Laboratory
Nearby individuals Regional population
* (mrem/y) (person-rem/y)
Pulmonary
Endosteum
Red marrow
Breast
Thyroid
3.1E-2
3.1E-2
2.5E-2
2.4E-2
1.2E-1
1.7E-1
2.6E-1
1.9E-1
1.4E-1
5.5
Table 2,6-3. Fatal cancer risks due to radionuclide emissions from the
Idaho National Engineering Laboratory^3)
_ Lifetime risk Regional population
Source ^ ,.,..., , , , f • ^
to nearby individuals (.Fatal cancers/y of operation)
INEL 5E-7 (2E-7) 6E-5 (3E~5)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.6-5
-------
Page Intentionally Blank
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2.7 Lawrence Livermore National...Laboratory; Liverntore, California
2.7.1 GeneralDescription
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 km^ 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 kn»2 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.
2.7.2 Description of Facility
There are five principal facilities that release radioactivity
into the air at Lawrence Livermore Laboratory.
Light Isotope Jjandling^Facilitjy {Building 331J
Tritium is the principal nuclide released from this facility which
carries out research and development in the area of light isotopes,
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. The effluent is continuously
monitored.
Electron PositronLingarRccejerator (LINRC) (Building 194)
Operation of the 100 MEV LINRC for neutron physics research
produces activation of nitrogen, oxygen, and dust particles in the air
of the facility. The effluent stream is continuously monitored before
release to the atmosphere from a 30-meter high stack.
Decontamination Facility (Building419)
The radioactivity in air effluents originates from various
decontamination operations. Stack effluents are continuously sampled.
2.7-1
-------
So 1 id Wa s te Pi s pos a I Fa c i l_i ty_ ...... (_Bu_i_l jiug_ 2 92)
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. During
operations, the stack effluent is sampled.
2.7.3 Radionuclide Emissions and Existing Control Technology
Table 2.7-1 identifies radioactive emissions from the facilities
at Lawrence Livermore Laboratory in 1981. For the purpose of this
analysis, all emissions in 1981 are assumed to be released from
Building 194 from a 30-meter stack.
Radionuclide emissions for the period 1979 to 1981 are shown in
Table 2.7-2.
Tritium emissions from the Light Isotope Handling Facility
{Building 331), the Insulated Core Transfer Accelerator (Building 212),
and the Solid Waste Disposal Facility (Building 292) are released
without treatment. HEPA filters are used to reduce emissions of
radioactive particulates from the Electron Positron Linear Accelerator
(Building 194), the Decontamination Facility (Building 419), and the
Solid Waste Disposal Facility (Building 292). Activation products from
the Electron Positron Linear Accelerator are released without treatment.
2.7.4 Health Impact Ass_essment___g^__Ig_'j^ence Livermore National Laboratory
The estimated annual radiation doses resulting from radionuclide
emissions in 1981 from Lawrence Livermore National Laboratory are
listed in Table 2.7-3. Nearby individuals are 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 2.7-4 shows the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional population
from these doses. The risk estimates include estimates which use a
dose rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I.
2.7-2
-------
Table 2,7-1* Radionuciide 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
^Reported as "Unidentified Alpha."
(k)Re ported as "Unidentified Beta + Gamma."
Table 2.7-2. Radionuciide emissions from the
Lawrence Livermore National Laboratory, 1979 to 1981 (Ci/y)
Radionuciide
1979
1980
(a)Rep0rted as "Unidentified Alpha,"
^Reported as "Unidentified Beta + Ga.mma."
NR None reported.
1981
Argon-41
Tritium
Nitrogen-13
Oxygen- 15
Plutonium-239
-------
Table 2.7-4. Fatal cancer risks due to radiortuclide emissions
from the Lawrence Livermore National Laboratory,
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
LLNL 3E-5 (1E-5) 2E-3 (6E-4)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low~LET radiations, as described in Chapter 8,
Volume I, of this report.
2.7-4
-------
2 • 8 Los jllamos National Laboratory; Los ..... Rlamos, Mew Hexico
2.8.1 Gene r a l_ Desc r ip t ion
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, LJ\NL 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 bioraedical 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
HP- Site (TA-33), the South Mesa Site (TA-3), the Omega site (fA~2), and
several other technical areas. Programs at these sites include the
operation of an 800 MeV proton accelerator, laser and magnetic fusion
activities, and research reactors- -one of which is an 8 megawatt
reactor--at the Omega site, and experiments using a tandem Van de
Graaff accelerator.
2.8.2 Description of Facility
During 1981, effluents were released from more than 15 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.
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. Ol.her
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.
2.8-1
-------
This site has two primary research areas, DP west and DP East, DP
West is concerned with tritium research, DP Bast 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.
Tft-33, HP-Site
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 the Geosciences Division
related to the Hot Dry Rock Geothermal Project are also here.
Tft-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,
TJr-jlj. ..,W-S_ite
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.
Tfl-43, Health Research Laboratory
The Bioraedical 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.
TR-46, WR Site
applied photochemistry, which includes development of technology
for laser isotope separation and laser-enhancement of chemical
processes, is investigated at this site, solar energy research,
particularly in the area of passive solar heating for residences. Is
also done.
Tft"48. Radiochemistry Site
Laboratory scientists and technicians at this site study nuclear
properties of radioactive materials by using analytical and physical
2,8-2
-------
chenistry. Measurements of radioactive substances are made and "hot
cells" are used for remote handling of radioactive materials.
TR-50» Waste ManageBaent__SitLe
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.
Tfl-54, Waste Disposal Site
This is a disposal area for radioactive and toxic wastes.
TR-5 5, PIutqnlum Process ing Fac iIi t ies
Processing of plutonium and research in plutonium metallurgy are
performed here.
2.8.3 Radionuclide Emissions and Existing Control Technology
Radioactive airborne releases at Los Alamos are summarized in
Table 2.8-1. Emissions from all stacks within a Technical Area were
summed, and the curie quantities of each radionuclide discharged within
an Area are listed. Emissions include various isotopes of uranium and
plutonium, americium-241, and activation products {berylliura-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, all emissions were modeled 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 the other 12
TAs. The effluents listed in Table 2.8-1 were summed to provide the
radioactive source terms for the two point sources. These quantities
are listed in Table 2.8-2. All effluents are released from stacks with
assumed heights of 30 meters.
2.8-3
-------
The effluent control devices at LANL are determined by the types
of activities conducted at each 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 passed 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,
In 1983* the quantity of airborne activation products from the
linear particle accelerator at the TA-53 Meson Physics Facility
increased about 85 percent (about 213,000 curies more than in 1982) due
to higher operating levels. These activation products include the
following short half-life (2 to 20 minutes) radionuclides: carbon-11,
nitrogen-13, oxygen-14, oxygen~15, argon-41, gold-192, and mercury-195.
2.8.4 Health Impact As sessment o f Lo s Alamos Na tional Lab oratory
The health risk assessment performed for this facility is
summarized in Tables 2.8-3 and 2.8-4. The risk estimates include
estimates which use a dose rate effectiveness factor of 2.5, as
described in Chapter 8, Volume I. 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 nearby individuals with
respect to the TA33 source are located 930 m southwest of the stack,
while the nearby individuals with respect to the combined area source
are located 2100 m south-southwest of the hypothetical stack. The
population within the 80 km radius assessment area is 100,000 people.
Los Alamos* estimates the dose to the most exposed nearby
individuals to have increased to 34 mrem/y to the whole body during
1983, compared to 8.1 mrem/y in 1982. This increase is due both to the
higher operating levels of the linear particle accelerator at the TA-53
Meson Physics Facility and recent construction of a residential
dwelling near the site boundary, so that people are exposed for longer
time to higher levels of airborne activation products. A dose of 34
mrem/y corresponds to a lifetime risk of 8E-4. The risk would be 2E-4
if a dose rate reduction factor of 2.5 for low-LET radiation, as
described in Chapter 8, Volume I of this report is included.
*Los Alamos National Laboratory, Environmental Surveillance at Los
Alamos During 1983, Los Alamos National Laboratory Report,
LA-101QO-ENV(UC-41), April 1984.
2.8-4
-------
Table 2,8-1. Radionuclide emissions (Ci) from
Los Alaisos National Laboratory, 1981
Radionuclide
Tritium
Beryllium- 7
Carbon- 1 I
Nitrogen- 13
Technical Area
2 3 21 33 35
9.0E+2 1.11+2 6.1E+3
41 43
1 . 3E+2
Oxygen-15
Phosphorus-32
Argon-41
Iodine-131
3.0E+2
Radionuclide
4.4E-5
46
48
Technical Area
50
53
54
2.0E-5
Uranium-235
Uranium-238
Uranium- 235/238
Plutonium- 239
Plutonium-238/239
Americium- 241
MPP
1.8E-6
1.6E-4
5.3E-5
4.0E-5
1.7E-4
1 . OE-3
6.21-6
5.9E-6
2.9E-7
2.8E-6
2.7E-7 3.7E-7
55
Tritium
Beryllium-7
Carbon-11
Nitrogen-13
Oxygen-15
Phosphorus-32
Argon-41
Iodine-131
Uraniura-235
Uranium-238
Uranium-235/238
Plutonium-239
Plutonium-238/239
fljnericiuffl~241
MPP
1.4E-5
2.3E-6
1.3E-6 1.6E-6
1.2E-7
1.4E-3 2.3E-5
6.6
3.9E+1
1.3E+5
2.5E+4
2.0E-15
1.1E+3
9.0E-9 4.9E-8
4.8E-8
Kixed fission products.
2.8-5
-------
Table 2,8-2.
Rsdioouelide emissions (Ci) from Los Alamos
National Laboratory, 1981
Radionuclide
Tritium
Beryllium- 7
Garb on- 11
Nitrogen-13
Oxygen- 15
Phosphorus~32
Argon-41
Iodine-131
Uranium-235
Uranium-238
Uranium~235s -238
Plutonium-239
Plutonium-238, -239
Amer i c ium- 24 1
MFP
Technical Area All other
33 TAs^3*
6.1E+3 1.1E+3
3.9E+1
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
(^Technical Areas: 2, 3, 21, 35, 41, 43S 46, 48, 50, 53-55,
ties summed from Table 2.7-1,
MFP Mixed fission products.
Quant i-
Table 2.8-3. Radiation dose rates from radionuclide
emissions from the Los Alamos National Laboratory
From TA33 source
From all other sources
Organ
Endosteum
Red marrow
Breast
Nearby
individuals
(mrem/y)
5.4E-1
6.8E-1
6.8E-1
Regional
population
( person-rem/y )
1.4
1.8
1.8
Nearby
individuals
(mrem/y)
1.1E+1
1.1E+1
9.1
Regional
population
( person-rem/y )
6.2E+1
5.9E+1
5.1E+1
2.8-6
-------
Table 2,8-4,
Fatal cancer risks due to radioactive emissions from
the Los ftlamos National Laboratory^3'
Source
TR33
all other nHos
Nearby individuals
(Lifetime risk)
1E-5 (5E-6)
2E-4 (6B-5)
Regional population
(Total cancers/y of operation)
5E-4 (2E-4)
IB- 2 (5E-3)
risk estimates in parentheses include a dose rate reduction
factor of 2,5 for low-LET radiations, as described in Chapter 8,
volume I, of this report.
2.8-7
-------
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Page Intentionally Blank
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The Oak Ridge Associated Universities (GEM)} conduct research in
areas such as biological chemistry, immunology, nuclear siedicine, and
radiochemistry. Radionuclldes are handled in encapsulated or liquid
form and the potential for producing gaseous effluents is very small.
2.9.3 Radionuc 1. icte Emissions and ..Exist. ing Cont ro 1 . Technology
The central radioactive gas disposal facilities release tritium,
iodlne-131, and krypton and xenon from tadioisotope 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 HEPfi 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,
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,
OaJL Ridge ..... GaLseous_,,Dif fusion,,, Plant.
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.
¥-12 Plant
Many of the procedures conducted at the ¥-12 Plant release
participate activity into the room exhaust air. Laboratory and room
air exhaust systems are equipped with filtration systems which may
Include prefilters, HBPA filters, or bag filters,
2.9-2
-------
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-------
Table 2=9-2. Radionuclide emissions from the Oak Ridge Reservation,
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Carbon- 14
Tritium
Iodine-125
Iodine-131
Krypton-85
Plutonium-239^a^
Technetium-99
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Xenon- 13 3
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-L
8.3E-4
1.2E-4
4.1E-3
4.2E+4
1.2E-3
I.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
(^Reported as "Unidentified Alpha."
estimate.
2.9.4 Health Impact Assessment of Oak RidgeReservation
The health impact assessment resulting from radionuclide emissions
in 1981 from the Oak Ridge Reservation is listed in Tables 2.9—3 and
2.9-4. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I. The
nearby individuals are 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 uraniutn-234 and tritium.
Table 2.9-3. Radiation dose rates from radionuclide
emissions from the Oak Ridge Reservation, 1981
.. Nearby individuals Regional population
Organ t / \ t / i
(.mrem/yj (person-rem/yj
Pulmonary
Thyroid
Endosteum
Kidney
50
9.3
7.6
5.4
212
15
22
15
2.9-4
-------
Table 2,9-4. Fatal cancer risks due to radionuclide
emissions from the Oak Ridge Reservation, 1981^a'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Oak Ridge Reservation IE-4 (1E-4) 8E-3 (6E-3)
(a/The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.9-5
-------
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-------
2.10 Paducjah^asw
2.10,1 Genenai, _De scr i^tlon
The Paducah Gaseous Diffusion Plant (PGDP) is a uranium enrichment
gaseous diffusion plant with a uranium hexatluoride (UPg) manufactur-
ing plant and various other support, facilities. 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 kra^. 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
contained within a 3 km^ 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 13 kilometers southwest, Paducah,
Kentucky, a city oE 35,000, is located 19 kilometers east of the
plant. The population within a 80 km radius is 450,000.
2-10.2 Descri£tj.on_ _ojf jFaci. II
The primary plant, the diffusion cascade, contains a physical
process in which UPg is fed into the system, pumped through the
diffusion stages, and eventually is removed as UFg. 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,
The manufacturing building or Feed Plant uses hydrogen, anhydrous
hydrogen fluoride (HP), and uranium oxide (003) to produce the UFg
that is fed into the diffusion cascade.
The Uranium Recovery arid Chemical Processing Facility conducts
operations that Involve pulverizing and screening of uranium salts,
At the Metals Plant, depleted UFg from the Cascade is reacted
with HF to convert it to UP*^ which is more easily stored.
2.10.3 RadJLOTWcjj^^^
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
2.10-1
-------
stream. These contaminants, which consist of isotopes of uranium and
technetium-99, are released from the diffusion cascade building stack,
which is sampled regularly.
Gaseous emissions from fluorinatlon operations, which convert
Ul?4 to UPg, are passed through a series of waste treatment systems
that include cold traps, fluid bed absorbers, and sintered metal
filters. HEPA and bag filters are also used to treat other emissions
from the Feed Plant. Bag filters are used to reduce airborne emissions
from the Uranium Recovery and Chemical Processing Facility.
Radioactive material emissions are from two discharge points,
C-31G stack and vent O4QQ (Table 2.10-1). Releases for 1981 have
increased when compared to the average for 1979-1981, except for
technetium which has decreased (Table 2.10-2). All releases were
assumed to be at ground level from vent C-400 for calculational
purposes. Releases for 1982 from the C-4QQ 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 2.10-1,
Radionuclide emissions from the Paducah Plant,
(Ci/y)
1981
Radionuclide
Uranium- 23 4
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
l.OE-2
5.0E-4
3.0B-5
3.9E-2
_
Total
1981
l.OE-2
5.3E-4
3.QE-5
3.9E-2
6.1E-3
Table 2.10-2.
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
6.1E-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
2.10-2
-------
2.10,4 Health Imgact ;As_s_es_snrent__o_£ the_Paduc3h_Plant
The estimated annual radiation dose and fatal cancer risks from
plant emissions in 1981 are listed in Tables 2.10-3 and 2,10-4. The
risk estimates include estimates which use a dose rate effectiveness
factor of 2.5, as described in Chapter 8, Volume I. The nearby
individuals are located 1100 meters north of the release location. The
predominant exposure pathway is that of inhalation. The annual
radiation dose is primarily from uraniunv-234 and uranium-238.
Table 2.10-3. Radiation dose rates from radionuclide emissions
from the Paducah Plant, 1981
Nearby individuals Regional population
(tnrem/y) (person-rem/y)
Pulmonary
Endosteum
Thyroid
Kidney
Red marrow
4.7
7.1
2.0E-1
3.6
5.1E-1
3.4
1.3E + 1
4.3E-1
6.7
9.3E-1
Table 2.10-4. Fatal cancer risks due to radionuclide
emissions from the Paducah Plant,
Lifetime risk to Regional population
Source nearby individuals (Fatal cancers/y of operation)
Paducah Plant IE-5 (1E-5) 1E-4 (IE-4)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low~LET radiations, as described in Chapter 8,
Volume I, of this report.
2.10-3
-------
Page Intentionally Blank
-------
2.11 Portsmouth. gaseoiis_DiffiisiQn.,Plant.i Piketon,,^Ohio
2.11,1 General Description
The Portsmouth Gaseous Diffusion Plant is operated by Goodyear
Atomic Corporation, a subsidiary of the Goodyear Tire and Rubber
Company. The principal process in the plant is the separation of
uranium isotopes by gaseous diffusion. 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 ra
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: 1,000), 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.
2.11.2 Description of^Facility
The Portsmouth Gaseous Diffusion Plant consists of a 4020-Stage
Enrichment Cascade and a 60-Stage Purge Cascade for enriching UFg in
the U-235 isotope. The Portsmouth Cascade is housed in three
processing buildings (X-326, X-330, and X-333), and is the only
domestic enrichment plant capable of producing very highly enriched
uranium (97.65 percent U-235). A Cold-Recovery System is used to
recover UFg from the system and to purge light contaminants (air,
N2« HF, and coolant) from the diffusion cascade.
2.11.3 Radj.onucl.lde Emissions... andL.SjcistIng control Technology
The gaseous radioactive discharges for 1981 representing all
cold-recovery activities for the plant are shown in Table 2.11-1.
The total air emissions of radioactive material have decreased for most
radionuclides from 1979 to 1981. The most significant release point
for 1981 is X-326 Top Purge Vent. This release point discharged
2.11-1
-------
approximately 84 percent of the total plant releases as shown in
Table 2,11-1. Uranium emissions are assumed to be Class W.
Emissions from the Cold-Recovery System are passed through sodium
fluoride traps before release. The X-326 Purge Vent is equipped with
alumina traps to reduce airborne emissions.
2.11.4 Hea1thImpact Assessment of Fortsmouth Plant
The estimated annual radiation doses and fatal cancer risks
resulting from emissions in 1981 at the Portsmouth Plant are listed in
Tables 2.11-3 and 2.11-4. The risk estimates include estimates which
use a dose rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I. The nearby individuals are 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 2.11-1. Atmospheric emissions of radionuclides from
the Portsmouth Plant, 1981
Emissions
Source/Radionuclide (Ci/y)
Top Purge Cascade
X-326 Top Purge Vent
Protactinium-234M 3.7E-2
Technetium-99 l.OE-1
Thorium-234 3.7E-2
Uranium-234 8.5E-2
Uranium-235 2.5E-3
Uraniurn-236 3.4E-5
Uranium-238 1.4E-4
X-330 Cold Recovery System Vent
Protactiniura-234M 2.0E-2
Technetium-99 2.8E-3
Thoriura-234 2.0E-2
Uranium-234 9.7E-4
Uranium-235 4.7E-5
Uranium-236 1.1E-6
Uranium-238 5.5E-4
2.11-2
-------
Table 2,11-1. Atmospheric emissions of radionuclidles from
the Portsmouth Plant, 1981 (continued)
Emissions
Source/Radionuclide (ci/y)
X-333 Cold Recovery
X-333 Cold-Recovery System Vent
Protactiniura-234M 9.9E-4
fechne t ium-99 1.2E-3
Thorium-234 9.9E-4
Uranium-234 5.7E-4
Uraniura-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
Table 2.11-2. Radionuclide emissions from the Portsmouth Plant,
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
Protactinium-234M
Technet ium-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
1.1B-1
5.8E-2
8.7E-2
2.6E-3
3.6E-5
1.3E-3
2.11-3
-------
Table 2,11-3, Radiation dose rates from radionuclide
emissions from the Portsmouth Plant, 1981
o Nearby individuals Regional population
(mrem/y) (person-rem/y)
Pulmonary
Thyroid
Endosteum
Kidney
Red marrow
6.9
2
11
5.1
0.8
11
7.9
35
17
3
Table 2.11-4. Fatal cancer risks due to radioactive emissions
from the Portsmouth Plant, 1981(a)
_ Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Portsmouth Plant 2E-5 (2E-5) 6E-4 (5E-4)
(a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.11-4
-------
2.12=1
The Rocky Flats Plant (RPP) 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 uraniura is utilized in component fabrication and is treated by
many of the sarae 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 km^ security area which is situated on 26.5
te^ 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.
2.12.2 Pescrip_tiQn of^Facility
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 americiura-241.
2.12-1
-------
locky Flats Plant also conducts operations involving the handling
of uranium. Depleted uranium-alloy scrap is consolidated 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.
2.12.3 Radionuclide Emissions and Existing Control Technology
Atmospheric emissions from the Rocky Flats Plant are listed in
Table 2.12-1. 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 RFI* 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 northeastern 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 emissions for the years 1979 to 1981 is given
in Table 2.12-2.
Exhausts from buildings where plutonium and uranium are stored and
processed are passed through multiple banks of HEPA filters prior to
release to the atmosphere.
2.12-2
-------
Table 2.12-1. Radlonuclide emissions from
the Rocky Plats Plant, 1981
Source/Hadlonuclide
Emissions
(Ci)
Plutonium Analytical Laboratory
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
Fabrication assembly Building
Building 707-106 Plenum
Tritium
Plutonium-239, -240
Uraniunr233, -234, -238
Building 707-108
Tritium
Plutonium-239, -240
Uranium-233, -234, -238
2.0B-2
4.4E-7
4.1E-7
3.9E-3
4.7B-8
1.6E-7
2.5E-3
5.5E-8
9.2E-8
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
4.61-3
1.6E-7
2.8B-7
1.4B-2
5.5B-8
2.0E-7
2.6E-3
5.OE-8
3.8E-8
6.4E-3
1.2E-8
1.1E-8
4.3E-3
5.7E-8
8.7E-8
2,12-3
-------
Table 2.12-1. Hadionuclide emissions from
the Rocky Flats Plant, 1981 (continued)
/^ j- , j » 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
fritium 8.0E-2
Plutonium- 239, ~240 4.5E-6
Uraniura-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 771c-Room Plenum
Plutonium- 239, -240 8.9B-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
Uraniuffi-233, -234, -238 5.2E-8
Building 774-202
Tritium 1.8E-3
Plutonium-239, -240 7.8E-8
Uranium- 233, -234, -238 2.QE-8
Manufacturing Building
Building 776-250
Tritium 1.5E-2
Plutonium-239, -240 l,2B-7
Uranium- 233 , -234, -238 2.0E-7
Building 776-206
Tritium 1.2E-1
Plutonium-239, -240 5.0B-8
Uranium- 233, -234, -238 1.9E-7
2.12-4
-------
Table 2.12-1. Radionuclide emissions fros
the Rocky Flats Plant, 1981 (continued)
/*, j- tjj Emissions
Source/Radionuclide - .,
Manufacturing Building (continued)
Building 776-201/203
Tritium 8.4B-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.QE-8
Uranium-233, -234, -238 2.8E-8
Building 776-204
Tritium 1.5E-2
Plutonium-239, -240 1.1B-7
Uranium-233, -234, -238 5.6B-7
Building 776-251
Tritium 1.7E-8
Plutonium-239, -240 4.8B-8
Uraniura-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
Uraniura~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
Laundry
Building 778 Laundry
Plutonium-239, -240 7.4E-8
Uranium-233, -234, -238 4.5E-7
2.12-5
-------
Table 2,12-1. Radionuclide emissions from
the Rocky Flats Plant, 1981 (continued)
, ,. ,. . Emissions
Source/Radionucllde , .
\ w i. /
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 1.08-6
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
Uraniunr233, -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
Building 883-Duct A
Uranium-233, -234, -238 7.0E-6
Building 883-Duct B
Uranium-233, -234, -238 5.8E-6
2.12-6
-------
Table 2,12-1. Radionuclide emissions from
the Rocky Flats Plant, 1981 (continued)
... ... Emissions
Source/Hadlonuclide , ,.
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 1.51-8
Uranium-233, -234, -238 8.8E-1
Assembly Building
Building 991-985
Plutonium-239, -240 8.8S-9
Uraniuro-233, -234, -238 1.6E-7
991 Main
Plutonium-239, -240 3.2B-8
Uranium-233, -234, -238 8.2E-8
Table 2.12-2. Radionuclide emissions from the Rocky Plats Plant
1979 to 1981 (Ci/y)
Radionuclide 1979 1980 1981
tritium 8.0E-1 7.8E-1 3.9E-1
Plutonium-239, 240 5.4E-6 1.2E-5 8.2B-6
Uranium-234 9.0E-6
Uranium-238 2.5E-5
Uranium-233, 234,
and 238 3.0E-5
Uranium-233, 234 1.5E-5
Uranium-238 1.4B-5
2.12-7
-------
2.12*4 Health ImpactAssessment of Rocky Flats Plant
The estimated annual radiation doses and fatal cancer risks
resulting from radionuclide emissions in. 1981 from the Rocky Flats Plant
are listed in Tables 2.12-3 and 2.12-4. The nearby individuals are
located 2260 meters north northeast of the release location. The
predominant exposure pathway is that of inhalation. The doses are
primarily from uranium-233, -234, -238; and plutonium-239 and -240.
Table 2.12-3. Radiation dose rates from radionuclide
emissions from the Rocky Flats Plant, 1981
Organ
Nearby individuals
(inrem/y)
Regional population
(person-rem/y)
Endosteum
Pulmonary
Liver
Red Marrow
1.5E-2
1.2E-2
2.8E-3
1.2E-3
1.6E-1
1.3E-1
2.9E-2
1.2E-2
Table 2.12-4. Fatal cancer risks due to radioactive emissions from
the Rocky Flats Plant, 1981
Source
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Rocky Flats Plant
2E-8
3E-6
2.12-8
-------
2.13 Sa.yan.naJL River, p lant j hiken. South Carolina
2*13,1 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 for 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.
2.13.2 Pescrip_tion ......... of Facility
SRP facilities are grouped into five major areas according to
their operational functions in the plutonium recovery process. These
areas and the major activities performed there include:
100 ftrea - 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 -_ Muc lear Product ^lon^
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 (D20) 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
P area, and uranium and other special nuclear materials are recovered
in the H Area.
2.13-1
-------
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 ftlloy 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.
40(1 Area--JteayY.. 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 D20 is then shipped to the 100
Area where it is used both as moderator and primary coolant in the
production reactors.
7QO_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,
2.13-2
-------
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.
2.13,3 Rad ionuc 1 ide Miss ions and_ Bxls t ing Cpn.tr o l^JTechno log^
Annual emissions for all facilities at SRP are summarized by
operational area in Table 2,13-1. 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 2.13-1).
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 Rrea -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
2,13-1). Control technologies employed include either scrubbers,
fiberglass filters, high-efficiency sand filters, or oxidation and
moisture trapping.
300 Rrea - 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.
2.13-3
-------
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 o£ 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 cobalt-60 is the only release of this nuclide reported
Cor the site. All discharges from processing areas are filtered
through at least two stages of HEPA filtration and a multi layered sand
trap before discharge from a 50-iaeter stack.
Summary of Radioactive amiss ions at..... SRP
The separations facilities and the reactor areas are responsible
for the majority of radioactive releases at SRP. The production
reactors release virtually all of the noble gases discharged at SRP and
one- third of the tritium (see Table 2.13-1). 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 2.13-2 indicates the releases for 1979 to 1981,
SRP occupies a large area of 710 square kilometers. However,
population densities in the vicinity of the site are relatively low.
For this reason, 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.
Table 2.13-1. Radionuclide emissions from the
Savannah River Plant, 1981 (Ci/y)
Radionuclide
Americiunr241
-------
Table 2.13-1. Radionuclide emissions from the
Savannah River Plant, 1981 (Ci/y) (continued)
Radionuclide
Curium- 244
Cobalt- 60
Cesium- 134
Cesium- 137
Tritium
Iodine- 129
Iodine- 131
Krypton-85
Kryj»ton-85m
Krypton-87
Krypton-88
Niobiura-95
Plutonium- 238
Plutonium- 239^
Ruthenium- 103
Ruthenium- 106
strontiun-90
Uranium-234
Uranium-238
Xenon- 13 1m
Xenon- 13 3
Xenon- 135
zirconium- 95
100
_
—
_
-
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
"
Area
200 300 400 700
1.6E-4 -
- 8.9E-5
6.4E-4 ~
3.1E-3 -
2.3E+5 - 2.0E43 1.5E-H
1.6E-1 - - 5.0E-6
3.7E-2 - - 3.2E-3
8.4E+5 -
6.4E-2 -
4.57E-3 -
201? T "3 AX? "7 —
* O C» O J * OC* * ~ ~
IOC* O
. *3C» £. ~ ~~
7.8E-2 -
3.1E-3 - - 5.0E-6
6.1E-3 -
61 t? O
. la o _ _ _
6.4 -
-
_
1*7!?— O —
. t & £.
Total
1.6E-4
8.9E-5
6.4E-4
3.1E-3
3.5E+5
1.6E-1
4.7E-2
8.4E+5
1.3E+3
8.7E+2
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
2.5E43
1.7B-2
one-half that activity designated as "Unidentified Alpha."
one-half that activity designated as "Unidentified Beta +
Gamma .
2.13-5
-------
Page Intentionally Blank
-------
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 toes, Iowa. The site occupies 16.2
hectares in Story County.
2.14.2 Description of FactIity
The Rmes 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 RLRR was terminated on December 1, 1917. Decommissioning began
January 3, 1978, and was completed on October 31, 1981. ftt present,
varied research programs involving small amounts of radionuclides are
carried out at the site.
2.14.3 Radionuclide Emissions
Prior to decommissioning, the major airborne releases were tritium
and argon-41 from the flLRR. Tritium was the major radionuclide
released during the 1981 decommissioning activities, fable 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 HealthImpactassessment of Ames Laboratory
The estimated annual radiation doses and fatal cancer risks from
radionuclide emissions from ALRR are listed in fables 2.14-2 and
2.14-3. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I.
These estimates are based on a regional population of 630,000, The
nearby individuals are located 750 meters north of the facility. The
major pathway of exposure was ingestion.
2.14-1
-------
Table 2.14-1. ladionuclide emissions frora tones Laboratory, 1981
Radlonucilde
Cobalt-60
Tritium
Unidentified alpha
Unidentified beta 4 gamma
Zinc-65
Emissions
(ci/y)
2.2E-7
4.5
1.6E-7
2.7E-6
2.4B-7
Table 2.14-2.
Radiation dose rates from radionuclide emissions
from Ames Laboratory for 1981
Organ
Nearby individuals
(mreia/y)
Regional population
(person-rem/y)
Endosteura
Pulmonary
1 . IE- 3
9.6E-4
4.8E-3
4. OB- 3
Table 2.14-3, Fatal cancer risks due to radionuclide emissions
frora Ames Laboratory, 1981
Source
Lifetime risk
to nearby individuals
Regional population
(Fatal cancers/y of operation)
ALRR Stack
2E-8
(7E-9)
1B--6 (4E-7)
estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in chapter 8,
Volume I, of this report.
2,14-2
-------
2,15 BettisAtomic Power
2.15.1 General. P_escr.l£tlp_n
The Bettis Atomic Power Laboratory is operated for the Department
of Energy by the Westinghouse Electric Company. It is sited on an
0.8 square kilometer tract in West Mifflin, Pennsylvania, approximately
12 km southeast of Pittsburgh. The facility designs and develops
nuclear reactors for the DOE Naval Reactors Program. The population
within 80 kilometers of the site is 3.2 million.
2.15.2 Description.of Facility
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.
2.15.3 JRadionuclide Emissions, .and...Existing. Cont.rpJL^echnQlogY
Airborne emissions data for Bettis are presented in Table 2.15-1.
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.
Table 2.15-1. Radionuclide emissions from
Bettis Atomic Power Laboratory, 1981
Era iss ions
Radionuclide (Ci/y)
Tritium 3.0B-5
Iodine-129 2.5B-7
Iodine-131 8.4E-7
Krypton-85 1.6E-1
Ant imony-125 5.8B-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}
2.15-1
-------
2.15.4 Health _Igpact:
The entire site is modeled 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 respi-
rable. Actual site boundary distances were used for the location of
the nearby individuals.
Table 2.15-2 lists the estimates of Che annual radiation doses
resulting from radionuclide emissions. The nearby individuals are
located 410 meters north of the release point. The major pathway
contributing to the individual dose equivalent rate is inhalation.
Table 2.15-3 lists estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional population
from these doses. The risk estimates include estimates which use a
dose rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I. Inhalation is the predominant pathway contributing to the
fatal cancer risk.
Table 2.15-2. Radiation dose rates from radionuclide emissions
from the Bettis Atomic Power Laboratory
Nearby individuals Regional population
(rarem/y) (person-rem/y)
Pulmonary 3.9E-3 3.8E-2
Thyroid 1.5E-3 3.2E-3
Etidosteum 8.6E-4 4.9E-3
Red marrow 5.SE-4 3.9E-3
Table 2.15-3. Fatal cancer risks due to radionuclide
emissions from the Bettis Atomic Power Laboratory^3'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
BAPL 1E-8 (8E-9) 2E-6 (1E-6)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.15-2
-------
2.16 KrTOllSjjyiojri^ Kessei,riiifl,, and,, .Windsor _
Sites;Schenectady, New York
2.16.1 Gene r a 1 peso r ipt. ion
The Knolls Atomic Power Laboratory (KRPL) is operated for the
Department of Energy by the General Electric company. The facilities
of KAPL are located on three separate sites: Knolls, Kesselring, and
Windsor. KAPL is one of the two laboratories operated for the DOE
Naval Reactors Program.
Knolls andKesselring 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 3.2 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 8 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.
2.16.2 Description_of PaciI tty
Facilities at the Knolls site are utilized in the development of
naval reactors. Nuclear power plant operators are trained at the
Kesselring and Windsor sites. Pressurized water reactors are located
at both the Kesselring and Windsor site.
2.16.3 RadiQnuclide Emissions andExistingControl Technology
The chemistry, physics, and metallurgy laboratories at the Knolls
site are the only potential emitters of radionuclides to the
atnosphere, while effluents from reactor operations are the only source
of radioactive emissions at the Kesselring and Windsor sites.
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 charcoal filters.
2.16-1
-------
The exhaust systems of the reactors at both the Kesselring and
Windsor sites are fitted with HEPA filtration systems to control
participate 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 2.16-1.
Table 2.16-1. Radionuclide emissions from
Knolls Atomic Power Laboratory, 1981
Radionuclide
Argon- 41
Broraine-82
Carbon- 14
Cobalt- 60
Cesium- 137
Iodine- 131
Krypton- 83m
Krypton-85
Krypton- 85m
Krypton- 87
Krypton- 88
Manganese-54
Plutonium- 239
Sulfur-35
Antimony- 125
Strontiura-90
Uranium- 23 4
Uranium- 23 5
Uranium- 236
Uranium-238
Xenon-13lro
Xenon- 133m
Xenon- 133
Xenon- 135
Xenon- 138
Knolls and
Kesselring
sites
3.8
3.3E-4
1.8B-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
2.9B-5
8.7E-7
5 . 7B--8
9.0E-10
2.5E-4
1.4E-3
4.2E-2
4.0E-2
1.3E-3
BSaissions (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
3.7B-4
l.OE-2
9.5E-3
2.16-2
-------
2.16.4 He §1 _th, , Impac t Rsse s smen_t g£ ..... KRPL
airborne particles released are assumed to be respirable. The
assessment is based on all releases for the Knolls and Kesselrlng sites
being combined at a central point at the Knolls site. A release height
of 10 meters was assumed for all effluents, ftctual site boundary
distances were used for the Knolls site and the Windsor site. Table
2.16-2 presents the dose rates from radionuclide emissions at these
sites.
Knolls and Kesselring Sites
For the Knolls and Kesselring sites, the nearby individuals are
located 300 meters north of the release point. Ingestion is the major
pathway of exposure.
Windsor Site
For the Windsor site, the nearby individuals are located 110 m
south of the release point. Inhalation is the major pathway of
exposure .
Table 2.16-2. Radiation dose rates from radionuclide emissions
from the Knolls and Kesselring Sites
Organ
Bndosteum
Red marrow
Breast
Pulmonary
.Nearbyindividuals
Knolls and Kesselring sites
Windsor site
Endosteura
Red marrow
Breast
Pulmonary
1.4E-1
7.8E-2
5.0E-2
4.7E-2
6.6E-3
3.6E-3
2.4E-3
1.5E-3
Regional population (person-rem/y)
Knolls and Kesselring sites Windsor site
2.3E-1
1.4E-1
l.OB-1
1.3E-1
2.8E-3
1.7E-3
1.3E-3
9.6E-4
The lifetime risk to the nearby individuals and the total number of
fatal cancers per year of operation of these sites are listed in Table
2.16-3. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in chapter 8, Volume I. Ait
immersion is the major pathway of exposure for these estimates.
2.16-3
-------
Table 2.16-3. Fatal cancer risks due to radionuclide emissions
from Knolls Atomic Power Laboratory'3^
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Knolls and Kesselring
sites 9E-7 (4E-7) 3E-5 (1E-5)
Windsor site 4E-8 (2E-8) 3E-7 (1E-7)
estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.16-4
-------
2 , 1 '/ . i Genera.! __Descrip_t ion
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 roost of the residents of the
greater metropolitan San Francisco Bay area.
Lawrence Berkeley Laboratory is a large raultifaceted research
laboratory conducting programs of pure and applied research in
physical, biological, and environmental sciences.
2.11.2 Descr i
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 H1LAC, the 224-centimeter Sector- Focused Cyclotron,
and the 46? centimeter Cyclotron.
The Tritium Facility was designed to accommodate kilocurie
quantities of tritium as a labeling agent for chemical and biomedical
research. Radiocheniical and radiobiological studies in many
laboratories typically use millicurie quantities of various
radionuclides .
2.17.3 Radioniic I id_e Emiss tons
Radionuclide emissions during 1981 at Lawrence Berkeley Laboratory
are shown in Table 2.17-1,
Table 2.17-1. Radionuclide emissions from
Lawrence Berkeley Laboratory, 1981
Radionuclide • Emissions
(Ci/y)
Carbon-14 3.6F.-2
CobriH-60 4.0E-5
Tritium 70.4
Iodine-125 5.7R-4
Plutonium 239 2.5B-9
Strontium- 90 4.0E-5
2.17-1
-------
2. IV. 4 H§L§ItJ^ISIR^
fable 2.17-2 lists the estimates of the annual radiation doses
resulting from radionuclide emissions. The nearby individuals are
located 100 meters east of the assumed release point. The predominant
exposure pathway is ingestion.
Table 2.17-3 gives the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers per year of operation. The
risk estimates include estimates which use a dose rate effectiveness
factor of 2.5, as described in Chapter 8, Volume I. Ingestion is the
major pathway for population exposure.
Table 2.17-2. Radiation dose rates from radionuclide emissions
from the Lawrence Berkeley Laboratory
Nearby individuals Regional population
Organ . , „ , / .
(mrem/y) (person-rem/y)
Thyroid 1.6 7.7E-1
Table 2.17-3. Fatal cancer risks due to radionnclide
emissions from the Lawrence Berkeley Laboratory^
Lifetime risk Regional population
Source
to nearby individuals (Fatal cancers/y of operation]
T,m, 9E-6 C4E-6) 2E~ 4 (8E-5]
risk estimates in parentheses include a dose rate reduction
factor of 2,5 for low-LET radiations, as described in Chapter B,
Volume I, of this report.
2.17-2
-------
2.18,1 Generaj,_Descr_iEtiofi
Mound Facility is located in Miamisburg, Ohio, approximately 16
kilometers southwest of Dayton. Hound 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.
2-18.2 pes.crlp.tion of Facility
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
there are stored and transferred to the Stf Building. At the SW Build-
ing, operations involve disassembly, analysis and development of
nuclear components containing tritium, and the recovery of tritium
wastes.
Plutonium-238 was released in airborne effluents from H, PP, R,
SM, W>, TOR, 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, TOR, and 41 Buildings
involve radioactive waste disposal processes.
2.18,3 Radi.Qnuc.lide Emissions and Existing Control Technolo_gy_
Table 2.18-1 identifies radioactive emissions from nine buildings
at the Mound Facility in 1981.
Total emissions are assumed to be released from the SW Building
with an effective stack height of 61 meters. Table 2.18-2 compares the
radioactive emissions from Hound for the years 1979 to 1981.
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. At all other facilities, particulate
radioactivity is removed from process air streams by HEPA filters.
2.18-1
-------
Table 2.18-1. Radionuclide emissions from the Mound Facility, 1981
Source
Em \g_s_io ns (C i/ y )
Tritium
Plutonium-238
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
6.13E+2
3.80E + 2
3.24E+3
4.29E+3
1.1E-10
1.21E-6
3.55E-7
6.49E-6
4.20E-8
4.14E-8
1.07E-7
2.5 OE -8
2_.,31E-9
8.28E-6
Table 2.18-2, 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
2.18.4 Health Impact Assessment of the MoundFacility
The estimated annual radiation doses resulting from radionuclide
emissions in 1981 from the Mound Facility are listed in Table 2.18-3.
The nearby individuals are located 1,500 meters north-northeast of the
assumed release point (SW Building). Ingestion is the major pathway of
exposure, and nearly all of the dose is attributable to tritium.
2.18-2
-------
Table 2.18-4 gives the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional population
from these doses. The risk estimates include estimates which use a
dose rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I. The regional population within an 80 kilometer radius of the
site is 2.9 million, Ingestion is the major pathway for population
exposure.
Table 2.18-3. Radiation dose rates from radionuclide emissions
from the Mound Facility, 1981
„ Nearby individuals Regional population
Organ / i \ r i \
(.mrem/yj (,person-rem/y)
Intestine wall
Endosteum
Kidneys
2.5E-1
1.5E-1
1.9E-1
11.4
7.1
9.0
Table 2.18-4. Fatal cancer risks due to radionuclide
emissions from the Mound Facility, 198l(a'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Mound Facility 4E-6 (1E-6) 3E-3 (lE-3)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.18-3
-------
Page Intentionally Blank
-------
2.19 Nevada Yest Site^ ^M^e, CountY..,. Nevada.
2,19.1 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 DOB's nuclear weapons research and development
complex. Programs at NTS include nuclear weapons development, proof
testing and weapons safety, nuclear physics programs, and studies of
high-level waste management. Primary activities at NTS are centered
around the testing of weapons. Tests are conducted at the site for DOE
contractors (e.g., Lawrence Liveraore Laboratories, Los Alamos
Scientific Laboratory, Reynolds Electrical Engineering, and for the
Department of Defense). Since 1962, all nuclear weapons tests have
been conducted underground.
2-19.2 Description of Facility
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. 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).
2.19.3 Radionuclide emissions and Existingcontrol.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 arid tunnel
ventilations are presented in Table 2.19-1. In addition to the
monitored releases, the source terras 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. Experiments with waste disposal and fuel
storage may possibly release radionuclides, but no releases have been
reported for these operations.
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. 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.
2.19-1
-------
Table 2.19-1. Radionuclide emissions front Nevada Test Site in 1981
Radionuclide Emissions
(Ci/y)
Tritium 534
Iodine-131 0.05
Xenon-133 2700
Xenon-133m 29
Xenon-135 142
2.19.4 HealthImpact Assessment of the Nevada Test Site
The estimated annual individual radiation dose equivalents from
radionuclide emissions from the Nevada Test Site are shown in Table
2.19-2. The nearby individuals are located 34,000 meters south of the
assumed release point located near the center of the test site. Air
immersion is the major pathway for the individual dose equivalent rate.
Table 2.19-3 lists the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional
population. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I.
Ingestion is the major pathway contributing to the fatal cancer risk.
Table 2.19-2. Radiation dose rates from radionuclide emissions
from the Nevada Test Site
o Nearby individuals Regional population
Crarem/y) (person-rem/y)
Red marrow 2.IE-3 I.IE-3
Thyroid 1.8E-3 2.IE-3
2.19-2
-------
fable 2.19-3. Fatal cancer risks due to radioactive
emissions from the the Nevada Test
Lifetime risk Regional population
j™» __.„„,_ ** *. *«
to nearby individuals (Fatal cancers/y of operation)
NTS 3E-8 (1E~8) 3B-7 (IE- 7}
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low- LET radiations, as described in Chapter 8,
Volume I, of this report.
2.19-3
-------
Page Intentionally Blank
-------
2.20 PantexP1 ant; _AmariIlo, Texas
2.20.1 GeneralDescription
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 27 kilometers northeast of Amarillo,
Texas.
The Pantex Plant is split into numerous zones and some zones 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
259,000. This includes Amarillo, located 30 kilometers to the
southwest with a population of 150,000, and Pampa, 65 kilometers to the
northeast with a population of 23,000.
2,20.2 Description of Facility
The primary mission at Pantex involves assembling, 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 for 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.
2.20-1
-------
2.20.3 Radionue, 1 ide Em i s s ions a rid Ex is t i n.gCon tr o 1 Te ch no I ogy
Airborne emissions from Pantex for 1981 are given in Table 2,20-1.
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 and 1980 are also summarized in Table 2.20-1.
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 cotttrol technologies are appropriate
to the releases which result from the test firings, so atmospheric
dilution is relied upon.
Table 2.20-1. Radionuclide emissions from Pantex Plant
1979 to 1981 (Ci/y)
Radionuclide
Tritium
Uranium— 238
1979
2.QE-2
3.QE--5
1980
l.OE-1
5.01S-5
1981
9.5E-2
L.OE-5
2.20.4 Health Impact Assessment^ for the Pantex Plant
For the purposes of dose/health effects assessment, it is assumed
that all particles released are respirable. The assessment is based on
all emissions in 1981 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 in 1981 from the Pantex Plant are listed in Table 2.20-2.
The nearby individuals are located 1,350 meters north of the release
point. The major pathway contributing to the individual dose
equivalent rate is inhalation.
Table 2.20~3 gives the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional
population. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I. The
pathway contributing primarily to the fatal cancer risk is inhalation.
2.20-2
-------
Table 2.20-2. Radiation dose rates from radionuclide emissions
from the Pantex Plant, 1981
„ Nearby individuals Regional population
(mrem/y) (person~rem/y)
Pulmonary 4.6E-3 2.6E-3
Kidneys 3.9E-5 3.9E-5
Table 2.20-3. Fatal cancer risks due to radionuclide
emissions from the Pantex Plant, 1981^a'
„ Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Pantex Plant 8E-9 (7E-9) 7E-8 (6E-8)
'a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.20-3
-------
Page Intentionally Blank
-------
2.21
2.21.1 General
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
o£ the nation's weapons program. Major operations include the design,
development, and manufacture of special electronic and mechanical
nuclear weapons components. The population within 80 km is
approximately 1.9 million.
2.21.2 Description of Facility
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.
2.21.3 Radionuclide Emissions and Existing 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 given in Table 2.21-1.
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 silica gel.
Table 2.21-1. Radionuclide emissions from Pinellas Plant, 1981
Emissions (Ci/v)
Radionuclide
Main Stack
Laboratory Stack
Building 800
Tritium gas
Tritium oxide
Krypton-85
Carbon- 14
129.2
115.3
3.7
—
89.7
75.4
—
8.5B-5
2.81
4.63
—
_.,
2.21-1
-------
2.21.4 Halfampa
The estimated annual individual radiation dose equivalents from
radionuclide emissions from the Pinellas Plant are shown in Table
2.21-2. The nearby individuals are located 470 meters west of the
release point. Ingestion is the major contributor to the individual
dose equivalent rate.
The risks of fatal cancer are shown in Table 2.21-3. The risk
estimates include estimates which use a dose rate effectiveness factor
of 2.5, as described in Chapter 8, Volume I. The inhalation pathway
contributes most of the fatal cancer risk.
Table 2.21-2. Radiation dose rates from radionuclide emissions
from the Pinellas Plant
_ Nearby individuals Regional population
(mrem/y) (person-rem/y)
Kidneys 2.5E-1 8.9E-1
Intestine wall 3.0E-1 1.1
Table 2.21-3. Fatal cancer risks due to radioactive
emissions from the Pinellas Plant'3*
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Pinellas Plant 5E-6 (2E-6) 2E-4 (1E-4)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.21-2
-------
2.22 Rockwell. InternatiQnal.i. Sarita Susa_rta_.,California
2.22,1 General Description
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 station
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 NEC- 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.
2.22,2 Description ofFacility
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.
2.22.3 Radionuclide Emissions and Existing Control Technology
Table 2.22-1 compares radioactive releases for the years
1979-1981. The 1981 release information is used in the health impact
assessment section.
HEPA filters are used to remove particulates from the effluent
from the Radioactive Material Disposal Facility.
The total emissions are assumed to originate from Buildings 21 and
22, with an effective stack height of 30 meters.
2.22-1
-------
Table 2.22-1. Radionuclide emissions from Che 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.
2.22.4 Health Impact Assessment of Rockwell International
The estimated annual radiation doses resulting from radionuclide
emissions in 1981 from the DOE facilities at Santa Susana are listed in
Table 2.22-2. The nearby individuals are located 180 meters north of
the assumed release point (Buildings 21 and 22). Ingestion is the
predominant exposure pathway.
Table 2.22-3 gives the estimates of the lifetime risk to nearby
individuals and the number of fatal cancers per year of operation. The
risk estimates include estimates which use a dose rate effectiveness
factor of 2.5, as described in Chapter 8, Volume I. Ingestion is the
primary pathway for population exposure. The regional population
within 80 kilometers of the site is 8 million.
Table 2.22-2. Radiation dose rates from radionuclide emissions
from the Rockwell International Plant, SSFLS 1981
_ Nearby individuals Regional population
Organ / / % / / \
vmrem/y) (person-rem/y)
Endosteum 4.1E-5 1.2E-3
Red marrow 2.1E-5 5.8E-4
Thyroid 2.8E-7 7.9E-6
2.22-2
-------
Table 2,22-3. Fatal cancer risks due Co radionucllde
emissions from Che Rockwell International Plant, SSFL, 198l(a)
_ Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Rockwell 6E-11 (2E-11) 2E-8 (9E-y)
'a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.22-3
-------
Page Intentionally Blank
-------
2,23 SdndiaLjIational, L^borMpX-^J;j-,...BlfeM£Hg,j!fl-u^' ^ew Mexico
2.23.1 General Description
Sandia National Laboratories (SNL) is a nuclear ordnance
laboratory with locations in Albuquerque, New Mexico, and Livermore,
California. The Livermore site is discussed in Section 2,1 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).
2.23.2 DescriptiQn ofFacility
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 two Sandia Pulsed Reactors and the Annular Core Research
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 storage site.
2.23.3 RadionuclideEmissions and Existing Control Technology
Airborne releases from operations at SNL, Albuquerque, are
summarized in Table 2.23-1.
Table 2.23-1. Radionuclide emissions from
Sandia National Laboratories, 1981
Radionuclide Emissions
(Ci/y)
Argon-41 6.84
2.23.4 Health Impact AssessmentofSandia National Laboratories
The entire site is treated as a single ground level point source
release located centrally within the facility. Actual site boundary
distances were used in the calculations.
2.23-1
-------
Tables 2,23-2 and 2.23-3 list the estimated annual radiation doses
and fatal cancer risks from radionuclide emissions from Sandia National
Laboratories at Albuquerque. The risk estimates include estimates
which use a dose rate effectiveness factor of 2.5, as described in
Chapter 8, Volume I. The nearby individuals are located 3200 meters
west-northwest of the source. Air immersion contributes essentially
all of the observed dose equivalent rate and fatal cancer risk.
Table 2.23-2. Radiation dose rates from radionuclide emissions
Sandia National Laboratories
Nearby individuals Regional population
" (mrem/y) (person-rem/y)
Endosteum
Breast
Red marrow
8.5E-4
8.1E-4
8.0E-4
3.6E-3
3.4E-3
3.4E-3
Table 2.23-3. Fatal cancer risks due to radionuclide
emissions from Sandia National Laboratories^
Lifetime risk Regional population
•<- *-
to nearby individuals (Fatal cancers/y of operation)
SNL 1E-8 (6E-9) 9E-7 (4E-7)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.23-2
-------
2.24 .S.tanf.ord.J.Anear ftccelerator Center; Stanford, California
2.24.1 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.
2-24.2 Descriptionof PacilitY
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 energies of
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 SPSAR 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.
2.24.3 Radionuclide Emissions and Existing Control Technology
Airborne radioactivity produced as a result of SLAG operations
and respective half-lives of the radionuclides are listed in Table
2.24-1. 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.
2.24-1
-------
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.
Table 2.24-1. Radionuclide half-lives and 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
2.24.4 Health Impact assessment of Stanford LinearAccelerator
The estimated annual radiation doses and fatal cancer risks
resulting from radionuclide emissions from Stanford Linear accelerator
are listed in Tables 2.24-2 and 2.24-3. The risk estimates include
estimates which use a dose rate effectiveness factor of 2.5, as
described in Chapter 8, Volume I. The nearby individuals are located
250 meters south of the release location and the predominant exposure
pathway is air immersion.
Table 2.24-2. Radiation dose rates from radionuclide emissions
from Stanford Linear Accelerator
Nearby individuals Regional population
9an (mrem/y) (person-rem/y)
Endosteum 5.6E-3 3.6E-2
Breast 5.3E-3 3.4B-2
Red Marrow 5.3E-3 3.4E-2
2.24-2
-------
Table 2.24-3. Fatal cancer risk due to radionuclide emissions
from Stanford Linear Accelerator^*)
Lifetime risk to Regional population
nearby individuals (Fatal cancers/y of operation)
1E-7 (4E-8) 9E-6 (4E-6)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.24-3
-------
Page Intentionally Blank
-------
2.25
2.25.1 General Descriptioii
Reactive Metals , Inc. (RMI), is located in northeastern Ohio in
the City and County of Ashtabula approximately 80 km northeast of
Cleveland, 65 km north of Warren, and 80 km north of Youngstown, the
closest major population centers. According to the 1970 U.S. Census,
the population of Ashtabula County is 102,000 and the population within
80 km of the facility is about 1.6 million.
2.25.2 Process
Reactive Metals operates an extrusion plant which fabricates
uranium rods and tubing from ingots for use as fuel elements in nuclear
reactors. The ingots are first extruded by a press into either rods or
tubing, cooled, and then sectioned by abrasive sawing. Scrap material
is fed to a pyrophoric incinerator to form a uranium oxide.
Air from each stage of the fabrication process is exhausted
through a separate stack. Stacks 1, 2, and 3 exhaust air from the
extension press tunnel, the press exit area, and the cooling table
area, respectively. Air from the abrasive saws is exhausted from Stack
4. The only stack with filtration, Stack 5, exhausts the incinerator.
This stack has a Roto-Clone Type N Air Scrubber.
2.25.3 Radionuc lide Emiss ions
The only radioactive material released to the air from RMI is
insoluble natural uranium. Radionuclide emissions from Reactive Metals
in 1981 are listed in Table 2.25-1.
2.25.4 Healjh Impact Assessment
To evaluate the health impact from the operation of RMI} releases
from the facility were assumed to be from a single 10-meter stack. The
released material was assumed to be equal quantities of uranium-234 and
uraniuai-238 in equilibrium with daughters thorium-234 and
protactinium-234m, all in an insoluble form.
The estimated annual radiation doses and fatal cancer risks
resulting from emissions at RMI are listed in Tables 2.25-2 and
2.25-3. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I.
These estimates are based on a regional population of 1.6 million. The
nearby individuals are located 120 meters north of the release point,
The critical organ is the pulmonary and the predominant exposure
pathway is inhalation.
2,25-1
-------
Table 2.25-1. Radionuclide emissions from Reactive Metalss Inc.
Source
Stack 1
Stack 2
Stack 3
Stack 4
Stack 5
Radionuclide
Natural uranium
Natural uranium
Natural uranium
Natural uranium
Natural uranium
Emissions (Ci/y)
3.56E-4
1.49E-4
1.18E-3
3.Q3E-3
6.79E-5
Table 2.25-2. Radiation dose rates from radionuclide
emissions from Reactive Metals, Inc., 1981
_ Nearby individuals Regional population
Organ / / \ / i \
(mrem/y) tperson~rem/y)
Pulmonary 51.8 19.5
Endosteura 0.27 0.12
Kidneys 0.14 0,06
Intestinal wall 0.06 0,04
Table 2.25~3. Fatal cancer risks from radionuciide
emissions from Reactive Metals, Inc., 198l'a^
. Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
RMI 8E-5 (8E-5) 4E-4 (4E-4)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations , as described in Chapter 8,
Volume I, of this report,
2.25-2
-------
2,26 Worldwide Impact o_f___.Selec^tgd E3_d_ionucl_ide_s
Some radioiiuclides 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 2.26-1).
Table 2.26-1. Emissions of selected radionuclides from DOE facilities
which may lead to worldwide impact
Source
(a)
Emissions (Ci/y)
Carbon-14 Iodine-129 Krypton~85
Argonne National Laboratory
Brookhaven National Laboratory
Hanford Reservation
Idaho National Engineering
Laboratory
Oak Ridge Reservation
Savannah River Plant
KAPL
LBL
Pinellas
Shippingport
0
8. IE -4
3.2
1.7E-1
1.2E-3
6.9E+1
1.8E-1
3.6E-2
_
7.2E-2
0
0
0
3. 71 -2
0
1.6E-1
-
-
_
-
6.7
0
250
5.9E+4
6.6E+3
8.4E+5
1.4E-1
-
3.7
_,
Combined releases for all DOE
facilities
7.2E+1
2.0E-1
9.0E+5
'a'DOE facility having significant releases of selected radionuclides.
Garbon-14
By combining the emission of 72 Ci per year and the dose
equivalent conversion of 700 person-rem per Ci released, a worldwide
dose equivalent of 50,400 person-rem were committed from 1981 emissions
of carbon-14. Similarly, the estimate of fatal cancers due to these
emissions (using 0.1 fatal cancers per Ci—Table 2.26-2) is 7. Those
effects would be observed during the time it takes carbon-14 to decay
away, or over approximately 40,000 years.
2.26-1
-------
Table 2.26-2, Estimated radiation doses and fatal cancers from
emissions of selected radionuclides from DOE facilities
to the world population
r, ,. , ., World population
Radionuclide
(person-rera/Ci) (Fatal cancers/Ci release)^3'
Carbon-14 7E+2^) lE-l^J (4E-2)
Krypton-85 4E-3^d) lE-6^e>f^ (4E-7)
Iodine-129 2.8E+5^g) 8E+l^f) (3E+1)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I3 of this report.
equivalent recorded by red marrow and endosteal cells,
(United Nations Scientific Committee on the Effects of Atomic
Radiation, Sources and Effects of Ionizing Radiation, Annex C, 1977,
p. 120).
'c'Health effects integrated over all time (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).
equivalent is received by the skin (UNSCEAR, Sources and
Effects of Ionizing Radiation, Annex C, 1977, p. 121).
'e'National Council on Radiological Protection, Krypton-85 in the
Atmosphere, Report No. 44, 1975.
'f'Assumed 200 fatal cancer per million person-rem received.
'•S-'Kocher, D. C. , A Dynamic Model of the Global Iodine Cycle and
Estimation of Dose to the World Population from Releases to the
Environment, Environment International, Vol. 5, 15-31, 1981.
Iodine-129 and Krypton-85
The worldwide health impact of emissions of iodine-129 and
krypton-85 are of similar concern. In 1981, 0.20 Ci of iodine-129 and
900,000 Ci of krypton-85 were released from operations at all DOE sites.
The committed collective dose equivalent due to iodine-129 was
56,000 person-rem; for krypton-85, 3600 person-rem.
Health effects conversion factors taken from Table 2.26-2 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 16 fatal cancers and for the krypton-85 this
yielded an estimated 0,9 fatal cancers. Both of these calculated
values are based on an assumption of 280 fatal cancers per million
person-rem received by the world population.
2,26-2
-------
2.27 "ura^
2.27(A) Resumption of operations at the PUREX_JPlant
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 has been calculated for
operation of the PUREX plant in the 200 Area of the Hanford site.
Process Pe script ion
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-OQ89D) indicates
that PUREX will be reactivated in 1984 for additional reprocessing of
N-Keactor 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 is given in Table 2.27(A)-1.
Radionuclide Emissions and Existing Control Technology _ aj: Purex
Table 2.27(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 and fatal cancer risks from
resumed operation of the PUREX Plant are given in Tables 2.27(A)-2 and
2.27(A)-3. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I. The
nearby individuals receiving the highest dose equivalent are assumed to
be located 16,000 m south of the source. The major pathway contributing
to the individual dose equivalent rate is air immersion.
2.27-1
-------
Table 2.27(A)-1. Estimated radionuclide emissions 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 2,27(A)-2 Estimated radiation dose rates from resumed
operation of the PUREX plant
Organ
Red marrow
Endosteuin
Pu Imonary
Liver
Thyroid
Nearby individuals Regional population
(mrem/y) (person~rem/y) \a'
2.1
4.9
2.1
1.0
1.5
63
1.3E+2
30
15
1.4E+2
(59)
(122)
(14)
(131)
dose rates in parentheses are based on NRPB Publication R129;
see Chapter 7S Volume I, of this report.
Table 2.27(A)-3. Estimated fatal cancer risks from resumed operation
of the PUREX plant(a>
Lifetime risk Regional population
Source ,-,-.,, /„ -, i <• • \
to nearby individuals (.Fatal cancers/y of operation)
PUREX Plant 2E-5 (9E-6) 6E-3 (3E-3)
'a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.27-2
-------
2 . 2 7 ( B ) J-££H5jy3^
The U.S. Department of Energy has proposed resumption of operation
of the L-Reactor at Savannah River Plant.
Pr oc e s s De script ion
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.
to_diQnuclide Emissions From L -React or ^Opera^tjoRS^
Table 2.27(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.
o f .. ,rthe L-Iteactor
The estimated dose rates from resumption of the L-Reactor are
given in Table 2.27(B)-2 for the nearby individuals at the location of
highest risk. This location is 9,780 meters south of the release
location. Ingest ion is the major pathway for dose equivalent rate.
The fatal cancer risks from resumption of the L-Reactor are given in
Table 2»27(B}-3. The risk estimates include estimates which use a dose
rate effectiveness factor of 2.5, as described in Chapter 8, Volume I.
2,27-3
-------
Table 2.27(B)-1 Estimated radionuclide enisslons front
resumption of L—Reactor operations at
the Savannah River Plant
Radionuclides Emissions (Ci/yr)
Tritium
Carbon- 14
Argon-41
Kryptori-85im
Krypton- 8 7
Krypton-88
Xenon-133
Xenon- 13 5
Iodine-129
Iodine-131
Plutonium-239
Ame r i c ium- 24 1
Strontium-90
5.5E+4
1.2E+1
2.0E+4
6.0E+2
5.4E+2
8.0E+2
1.7E+3
1.4E+3
l.OE-4
4.1E-3
5.0E-7
5.0E-7
l.OE-4
Table 2.27(B)-2. Estimated radiation dose rates
from resumption of the L-Reactor, Savannah River Laboratory
_ Nearby individuals Regional population
(mrem/y) (person-rem/y)
Intestine wall 4.5E-1 19.6
Red marrow 3.8E-1 15.9
Thyroid 3.7E-1 15.3
2,27-4
-------
Table 2.2?(B)-3. Fatal cancer risks due to radionuclide emissions
from resumption of the L-Reactor, Savannah River Laboratory^3'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
L-Reactor 7E-6 (3E-6) 4E-3 (2E-3)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
2.27-5
-------
Page Intentionally Blank
-------
Chapter 3: NRC LICENSED FACILITIES AND
NON-DOE FEDERAL FACILITIES
3.1 Research and Test Reactors
Description
This category consists of those land-based reactors licensed by
the Nuclear Regulatory Commission which 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 fields such as physics, biology, and
chemistry; and education. Power levels range from near zero to 10 MW.
3.1.3 Control Technology
No effluent controls for argon-41 or tritium in the form of water
vapor are used on research and test reactors. Some facilities use
filters to remove the small quantities of fission products which may be
present; others do not (Co83).
3.1.4 Radiotiuc 1 ide Emi ss ions
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.
3.1-1
-------
Some research and test reactors are not required to submit data on
air emissions of radionuclides to the Nuclear Regulatory Commission
(NRG). However, many 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 (Co83).
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 facility with the
highest emission rates as shown in Table 3.1-1 was chosen to be the
reference facility. The emission rates used in Table 3.1-2 were for a
prior year, however. The actual stack height (50 m) of that facility
was used. Other parameters used in the analysis were chosen to be
representative of a major metropolitan area in the northeastern United
States.
3.1.6 HealthImpact Assessment of Reference Facility
The estimated annual radiation doses from the reference facility
for nearby individuals and population groups are shown in Table 3.1-3.
Fatal cancer risks to nearby individuals and to the regional population
are presented in Table 3.1-4. The nearby individuals are located 1000
meters north of the stack. The risk estimates include estimates which
use a dose-rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I.
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
total impact of research and test reactors was estimated as follows:
a. Emissions of argon-41 from reactors are roughly
proportional to their power level. The reactors
were grouped according to four power level ranges
(0.1-0.5 MW; 0.5-1 MW; 1-5 MW; and greater than 5
MW). The average emission rate was determined for
each group using the data in Table 3.1-1. Reactors
having a power level of less than 0.1 MW have
negligible emissions,
b. The metropolitan areas where reactors are located
were classified according to the population density
of the standard sites used in AIRDOS (see Appendix
A). All the reactors were classified as being in
standard sites A, B, and D. Only one reactor was
classified as being in Site A.
3.1-2
-------
c. The number of fatal cancers per year from the
reference reactor was estimated for each of the
three standard sites using AIRDOS.
d. The health impact may be estimated for a reactor at
each site by assuming that the impact of the reactor
and that of the reference reactor are in the same
ratio as their emission rates. Using this
relationship and the number of reactors in each
power level group at the three sites, the impact
from all seventy reactors may be estimated. The
total estimated impact is approximately 0.06 that of
the reference reactor.
3.1.8 Existing Emission Standardsand Air Pollution Controls
Research and test reactors licensed by NRC are subject to the
requirements of 10 CFR 20, Appendix B, Table II, which places limits on
air emissions to unrestricted areas. Argon-41 is limited to an air
concentration of 4 x 10~8 microcuries per milliliter above
background, and tritium is limited to an air concentration of 2 x
10"' microcuries per milliliter.
3.1.9 Supplemental Control Technology
Emissions of tritium in the form HT can 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.
Table 3.1-1. Radionuclide emissions from research and test reactors
Design
type
1. Heavy water
2. Tank
3. Heavy water
4. Heavy water
5. Pool
Power
(kW)
10,000
10,000
5,000
5,000
5,000
Radionuclide
Argon-41
Tritium
Argon-41
Tritium
Argon-41
Tritium
Argon-41
Emissions
(Ci/y)
465.0
155.0
2504
16
N/A
8560
22
350
N/A Not available.
3.1-3
-------
fable 3.1-1. Radionuclide emissions from
(Continued)
research and test reactors
Design,
type
6. Pool
7. Pool
8. Pool
9. Pool
10. TRIGA
11. TRIGA
12. Pool
13. TRIGA
14. TRIGA
15. TRIGA
16. Pool
17. Pool
18 . TRIGA
19. TRIGA
20. TRIGA
21. TRIGA
22. TRIGA
23. TRIGA
24. TRIGA
25. TRIGA
26. TRIGA
27. TRIGA
28. TRIGA
29. TRIGA
30. TRIGA
31. TRIGA
32. TRIGA
33. TRIGA
34. TRIGA
35. Pool
36. Graphite/water
37. Light water
38. TRIGA
39. TRIGA
40. Graphite/water
Power
(kW)
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
250
250
250
250
200
100
100
100
100
100
Radionuclide
Argon-41
Noble gas
Radio iodine
Particulate
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Argon-41
Argo n-4 1
Argon-41
Particulate
Argon-41
Argon-41
Argpn-41
Argon-41
Argon-41
Argon-41
none
Argon-41
None
Argon-41
Tritium
none
Argon-41
Argon-41
Argon-41
Argon-41
Einissions
(Ci/y)
247.0
47
0.021
0.01
N/A
6
0.09
2.1
N/A
9.2
7
2.9
14
10
41
2
0.001
2.6
1.8
1.2
1.0
0.003
0.016
0.0
0.06
N/A
N/A
0.0
0.002
N/A
0.002
0.0
3.1
33
N/A
0.001
N/A
68.2
N/A Not Available.
3.1-4
-------
Table 3.1-1.
Radionuclide emissions from research and test reactors
(Continued)
Design
type
41,
42.
43.
44.
45.
46.
47.
48.
49.
50.
51.
52.
53.
54.
55.
56.
57.
58.
59.
60.
61.
62.
63.
64.
65.
66.
67.
68.
69.
70.
Graphite /water
Graphite/ water
TRIGA
Special
TRIGA
Graphite/water
Pool
Pool
Pool
Homogeneous
Pool
Special
Special
Tank
Homogeneous
Homogeneous
Homogeneous
Homogeneous
Homogeneous
Homogeneous
Homogeneous
Homogeneous
Homogeneous
Tank
Homogeneous
Homogeneous
Homogeneous
Pool
Pulse
Pulse
Power
(kW)
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
Argon-41
Argon-41
none
none
none
none
none
none
Krypton-85
none
none
none
none
none
none
Argon-41
Eniiss ions
(Ci/y)
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
0.0
N/A
0.0
N/A
N/A
N/A
0.0
13
N/A Not available.
3.1-5
-------
Table 3.1-2. Reference facility
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
_ Nearby individuals Regional population
(mrera/y) (person~rem/y)
Average of all organs 1.0 340
Table 3,1-4. Fatal cancer risks due to radionuclide emissions from
the reference facility^3-'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Research and test
reactor 2E-5 (8E-6) 0,1 (4E-2)
risk estimates in parentheses include a dose rate reduction
factor of 2,5 for low- LET radiations , as described in Chapter 8,
Volume I, of this report.
3.1-6
-------
REFERENCES
Co83 Corbit C. D., Herrington W. N., Higby D. P., Stout L. A., and
Corley J. P., Background Information on Sources of Low-level
Radionuclide Emissions to Air, PNL-467Q, Prepared for EPA
under U.S. DOE Contract by Battelie Memorial Institute,
September 1983.
3.1-7
-------
Page Intentionally Blank
-------
3.2 Rcce_lerators
3.2.1 OeneraJLJDescrJj>t ion
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, Dynaraitron, resonant
transformer, and insulating core transformer.
Incremental acceleration machines are accelerators whose electric
field strength varies with time. This type of accelerator incrcar.es
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
accel«rutors 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 type is as follows: linacs, 50 percent; neutron generators (of
several different designs), 17 percent; Van de Graaff, 15 percent;
. ?.-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 HeV).
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 Cont ro1 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. The treatment of exhaust streams
prior to release is usally accomplished by high-effiency particulate
air (HEPA) filters, preceded by prefliters. In some cases, adsorptive
filters are necessary to remove specific types of gases. Examples
include activated charcoal and molecular sieves, which are usually
preceded (in line) by a particulate filter (Co83).
3.2.4 Radiortucj.ide! 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.
3.2-2
-------
3.2,6 HeaLth_..Impac.t Assessment
The health impact assessment for the reference facilities was made
for a mid-western, suburban site. The nearby individuals are located
1000 meters from the stack, and there are approximately 2,5 million
persons in the regional population.
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.
The risk estimates include estimates which use a dose rate effective-
ness factor of 2.5, as described in Chapter 8, Volume I.
Table 3.2-1.
Nuclear reactions responsible for some airborne
radioactivity
Reaction
(T,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- 1 1
Nitrogen- 13
Oxygen- 15
Carbon- 1 1
carbon- 14
Oxygen- 15
Nitrogen- 13
Carbon- 14
Argon- 41
Threshold
energy
(Mev)
10.5
15.7
18.7
11.3
18.0
20.0
Nft
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
3.2-3
-------
3,2-2. Estimated annual emissions from typical
particle accelerators (EPA79)
Radio-
nuc 1 ide
Carbon- 11
Nitrogen-13
Oxygen- 15
Tritium
Carbon- 14
Argon- 41
18 MeV
100 MeV Electron 6 MeV
Cyclotron Llnac Van de Graaff'a^
(Ci) (Ci) (Ci)
2.0E-3
4.0E-2
1.0
1
l.OE-9
l.OE-4
(a)Tritiura 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)
3.2-4
-------
Table 3,2-4, tanual radiation doses due to radioactive
emissions from typical accelerators (EPA80)
Nearby
Type of individuals Population
accelerator (rarem/y) (person-rem/y)
6 MeV
Van de Graaff 1.1E~4 5.9E-4
18 MeV
Electron linac 4.2E-8 3.1S-7
100 MeV
Research cyclotron 9.6E-5 5.11-6
Table 3.2-5. Lifetime risks to nearby individuals and number
of fatal cancers due to radioactive emissions from
typical accelerators (SPA80)
-------
3.2.7 Total Heal_th__Jjigact
The estimated total number of fatal cancers caused by all non-DOS
accelerators may be 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
Accelerators are regulated by the individual States. All of the
States have adopted standards equivalent to the Radiological
Concentration Guides given by the Nuclear Regulatory Commission in 10
CFR 20, Appendix B, Table II. The guides for carbon-14, argon-41, and
tritium are: IE-7 microcuries/ml, 4E-8 microcuries/ml, and 2E-7
microcuries/ml, respectively. The guide for isotopes with half-lives
less than two hours is 3E~6 microcuries/ml.
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.
3.2-6
-------
REFERENCES
BRH78 Bureau o£ Radiological Health, 1918, Report of State and
Local Radiological Health Programs, Fiscal Year 1971. HEW
Pub, No. 78-8034, FDA, Department of Health, Education and
Welfare, Rockville, Md. 20852.
Co83 Corbit C, D., Herrington W. N., Higby D. P., Stout L. A., and
Corley J, P., Background Information on Sources of Low-level
Radionuclide Emissions to Air, PWL-467Q, Prepared Cor EPA.
under U.S. DOE Contract by Battelle Memorial Institute,
September 1983.
EPA79 Environmental Protection Agency, ft 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
-------
Page Intentionally Blank
-------
3. 3 Radjiopjiarmaceu t ica I Indu s t_r y
3.3.1 General Description
Increasing medical and research demands for radioactive chemicals
have resulted in the evolution of a large radiopharraaceutical
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 radioimrnunoassay (RIA) kits (TI79). 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 (TI79).
Suppliers
Industrial suppliers produce 65 different, generally-used
radionuclides (EPA80). Major suppliers of radiopharmaceuticals and
medical isotopes are listed in Table 3.3-1 (TI79). 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 NRG 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 using radiopharmaceuticals (TI79).
Radionuclides used in diagnostic and therapeutic procedures are listed
in Table 3.3-2 (FDA76, NRC79),
3.3-1
-------
Table 3.3-1. Major suppliers of radiopharmaceuticals and medical
Isotopes, excluding nuclear pharmacies (TI79)
Location
Supplier
Product
California
Emeryville
Glendale
Vallecitos
Van Nuys
San Ramon
Davis
Irvine
Richmond
Florida
Miami Lakes
Georgia
East Point
Medi-Physics, Inc.
(home office)
Medi-Physics, Inc.
General Electric Company
Nuclear Med. Sves.,Ine.
Gammaceutics
University of California
ICN Pharmaceuticals
Bio-Rad Laboratories
Medi-Physics, Inc.
Medical Research
Foundation, Inc.
Illinois
Arlington Heights Amersham Corporation
Rosemont
Medi-Physics, Inc.
Indium-Ill, Iodine-123,
Gallium~67, Rubidium-81/
Krypton-Sim generators,
Xenon-133, Technetium-99m.
TechnetiuHi-99m-
labeled compounds.
Xenon-133.
Groups I, II, & IV.
Iodine- 12 3.
Iodine-123.
RIA
Iodine-125, Cobalt-57,
RIA kits.
Technet ium-99m~
labeled compounds.
Yttrium-90 microspheres.
Cobalt-58 as cyanocobalaniin,
Selenium~75 as
selenoraethionine,
Iodine-125 as fibrinogen.
Technetium-99m as per-
technetate. Kits for
preparation of Tc-99in
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 (TI79) (continued)
Location
Supplier
Product
Ind iana
Indianapolis
Elkhart
Massachusetts
Billerica
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.
Kits for preparation of
Tc-99in-labeled DTPACc),
and pyrophosphate.
Iodine-125 RIA kits.
Kits for preparation of
Tc-99m-labeled
pyrophosphate, DTPA.
Thallium-201, Galliuin-67,
Iodine-131, Iodine-125
Seleniura-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
Ma11inckr od t, Inc.
Diagnostic Products Div,
University of Missouri
See footnotes at end of table.
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).
3.3-3
-------
Table 3.3-1. Major suppliers of radiopharmaceuticals and medical
isotopes, excluding nuclear pharmacies (TI79) (continued)
Location
Supplier
Product
New Jersey
Princeton
S. Plainfield
Ohio
Cincinnati
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.
Medi-Physics, Inc.
Procter and Gamble Co,
Iodine-123, Gallium-67, Tc-99m,
Indium-Ill, Rb~81/Kr-81m
generators.
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.
10 CFR 35.100, Schedule A.
Radioimmunoassay.
Diethylenetriarnine pentaacetic acid.
3.3-4
-------
Table 3,3~2. Major radlopharmaceuticais and
their uses (FDA76, NRC79)
Radionuclide
Use
Phosphorus-32
Gallium-67
Rubidium-Si
Technetium-99m
lodine-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
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) technetiura-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).
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 3.3-3.
3.3-5
-------
Table 3.3-3. Estimated quantities of radionuclides received and
used by hospitals, 1977 (TI79)
Radionuclide Received Used
900-1500
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-1255 and technetium~99m have the greatest.
potential for release as airborne effluents from sewage treatment
plants (TI79).
3.3.2 Proce s s De s cription
Radionuclides used in the radiopharmaceutical industry are
produced by irradiation of target materials (or fuel) in a reactor or
accelerator, and by radioisotope generators.
Sjjp_p_li_e_rs_
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 3,3-4.
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 (TI79).
Table 3.3-4. Nuclear reactions used in radioisotope production
Reaction Examples
(1) Neutron-gamma (n,y ) ^Co + n ^ OUGO + y
(2) Neutron-proton (n,p) 32g + n -* 32p +
(3) Neutron-alpha (n,a) ^^Cl + n -*- ^2p H
3,3-6
-------
In a reactors the main steps in radionuclide production are as
follows (Ba66):
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
nuclide.
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 nuclides 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-1233 iron-52s mercur'y-199m, carbon-11,
nitrogen-13, and oxygen-15'.
Typical nuclear pharmacy productio'n 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.
UjS_er_s
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.
3.3-7
-------
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.
lechnetium-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-ReceivingFacijJLt:ies
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 (TI79).
It is estimated that approximately 21 percent of the sewage
treatment facilities in the U.S. employ incineration or pyrolysis for
sludge treatment (1179). 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.
3.3.3 Control Technology
Types of effluent controls employed by producers of
radiopharmaceuticals depend on the type and amount of each nuclide
handled in the facility (LeSO). 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
3.3-8
-------
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 (TI79). 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 (technetiutn-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 NRG 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 (TI79).
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 radiopharaaeeutical 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 3.3-5.
The NRG conducted a survey of over 3000 by-product material
licensees in late 1980 to collect annual radioactive effluent emissions
data (NRC81). Three hundred and eight-five industrial licensees
responded to the survey. Table 3,3-6 summarizes emissions data for the
facilities manufacturing radionuclides.
A report prepared for EPA includes average release rates for
radiopharmaceutical manufacturers and radiopharmacies (Cob83). Because
large releases from a single manufacturing facility are included, re-
leases should not be considered typical. For this reason, emissions
from this facility are listed separately in Table 3.3-7.
3.3-9
-------
Table 3.3~5. Radionuclide emissions from six major radiopharma-
ceutical producers (Coa82, EPA80, Fr82a, Fr82b, Ro82a, Ro82b)
Producing
Plant
A
B
C
D
E
F
Emissions
Iodine-125 Iodine-131
1.8E-2 3.9E-4
2.2E-6
-
_
l.OE-2 7.6E-2
2.6E-3 3.1E-2
(Ci/y)
Technetiura-99m
-
-
4.14E-3
4.5E-3
-
-
Table 3.3-6. Summary of reported atmospheric emissions of
radionuclides from 385 industrial facilities (NRC81)
Source
Iodine-131
Iodine-125
Xenon~133
Molybdenum-99
Technetium-99m
Tritium
Number of
facilities
using
nuclide
11
55
6
4
2
66
Number of
facilities
reporting
releases
4
25
4
4
1
21
Mean
1.8E-4
1 . 7E-3
7.0
8.31-6
3.2E-6
5.1E+1
Emissions
Maximum
4.6E-4
2.0E-2
2.3E+1
3.0E-5
3.2E-6
7.4E+2
(Ci/y)
Minimum
3.0E-5
3.0E-8
2.0E-2
1.5E-7
3.2E-6
l.OE-4
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. A survey conducted by Battelle
Memorial Institute to update the emissions was generally in agreement
with the values listed in the table (Cob83).
3.3-10
-------
Table 3.3-7. Radionuclide emissions from a large radlo-
phanaaceutical producer (Cob83)
Radionuclide
Emissions (Ci/y)
Krypton-83tn
Krypton-85
Krypton-85m
Krypton-87
Iodine-125
todine-131
Xenon-133
Xenon- 13 3m
Xenon-135
Argon-41
6.1E+2
2.3
1.7E+3
1.6E+2
2,3
3.4
1.9E+4
2.2E+3
1.1E+4
1.2E+3
Table 3.3~8» Summary of reported atmospheric emissions of radio-
nuclides from 860 government and public medical facilities (NRC81)
Number of
„ facilities
Source
using
nuclide
Iodine-131
Iodine-125
Xenon-133
Ho lybdenum~99
Technetium-99m
Cobalt-60
346
270
229
268
73
112
Number of
facilities
reporting
releases
25
19
142
3
2
6
Emissions (Ci/y
Mean
2.9E-3
1.7E-3
4.6E-1
1.0
2.8E-1
1.3E-2
Maximum
S.OE-2
9.5E-3
6.4
3.0
5.0E-1
7.2E-2
Minimum
2.0E-8
l.OE-8
2.0E-5
l.OE-8
5.2E-2
l.OE-7
3.3-11
-------
Radioactive airborne emissions resulting from sludge drying and
incineration at a sewage treatment plant were studied (1179) 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.
3.3.5 Ref;erence Fac i 1 ities_
Radiopharmaceutica1 Supp1ier Facility
The radiopharraaceutical supply industry can be characterized as
generally urban, with suppliers located near their major users,
hospitals (TI79). 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 Hiolybdenum-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
3.3-5 and 3.3-6.
Emissions from the reference facility are controlled by charcoal
beds and HEPA filters.
UserFacility
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.
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
3.3-12
-------
Table 3.3-9. Reference facilities of typical suppliers and
users of radiopharmaceuticals
Parameter
Value
Product line:
Emission controls
Stack parameters:
Size :
Volume of administrations
Emission controls:
Sewage Treatment Plant
Process:
Iodine-131s iodine~125j xenon-133,
technetium-99m, raolybdenum-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 HE PA
filters. Release through building
ventilation roof vents.
Vent height: 10 m
Sludge drying and incineration
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 (TI79).
3,3.6, Health Impact Assessment of Reference Radiopharmaceutical
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). Nearby
3,3-13
-------
Table 3,3-10, Radionucltde emissions from reference
radiopharraaceutical industry facilities
„ /n i- i • j Emissions
Source/Radiotiuclide , . , ^
Iodine-125 2.0E-2
Iodine-131 7.6E-2
Xenon-133 2.3E+1
Techn.etium-99m 4.5E-3
Iodine-125 9.5E-3
Iodine-131 5.0E-2
Xenon~133 6.4
Sewage Tre a tme nt P 1 ant
Iodine-131 5.0E-4
Technetium-99m 8.0E-4
Table 3.3-11. Radiation dose rates from radionuclide emissions
from the reference radiopharmaceutical industry facilities
_ Nearby individuals Regional population
Organ /• / \ t / \
(mretti/y) \person-rent/y)
Thyroid " 3.2E-1 2.5
Thyroid 3.7E-2 1.9E-1
S^ewaee^_t£e^_tinent_p_la|it_
Thyroid ~~~~ ~~~~~ 8.0E-4 7.4E-3
3.3-14
-------
individuals are located 500 meters from the supply facility, 500 meters
from the medical facility, arid 500 meters from the sewage treatment
plant .
Table 3,3-12 presents estimates of the lifetime risk to nearby
individuals and the number of fatal cancers to the regional populations
from these doses. The risk estimates include estimates which use a
dose rate effectiveness factor of 2.5, as described in Chapter 8S
Volume I.
3,3.7 jjgjLkth. ..jjEE*!- --- -A^ £:§JT§J!L§ILL-2J— ^.B-^P •*• ^...•'•.P j^adiophiarmaceutical_
In a recent survey of radiopharmaceutical users, EPA identified
those facilities which have the largest radionuclide emission rates and
estimated the resulting dose to nearby persons (JFA84). Most of the 32
facilities contacted in the survey were large medical centers. Twenty-
three facilities cooperated with the survey and gave useful informa-
tion. Iodine-125 is the radionuclide of concern at most facilities.
At some facilities, however, xenon-133 or iodine-131 is the radio-
nuclide of concern. All of the facilities, with the possible exception
of one , have emissions that result in doses of less than 10 mrem/y to
any organ* A more accurate calculational technique than that used in
this survey may produce dose estimates of less than 10 mrem/y for this
facility, however.
Table 3,3-13 lists the estimated doses from radionuclide emissions
from the radiopharmaceutical production facility. The emissions from
this facility are listed in Table 3.3-7. The nearby individuals are
located 1500 meters from the facility.
The estimates of the lifetime risk to nearby individuals and the
number of fatal cancers for the regional population resulting from
these doses are listed in Table 3.3-14. The risk estimates include
estimates which use a dose rate effectiveness factor of 2.5, as
described in Chapter 8, Volume I.
3.3,8 Total Health Impact of the Radiopharmaceutical Industry
For all segments of the radiopharmaceutical industry, the
estimated total health impact may be obtained as follows.
The estimated total health impact caused by all radiopharma-
ceutical suppliers is based on the assumptions that (1) emissions of
1-125, I™131, Xe-133, and Tc~99m reported for industrial facilities in
a survey by NRG (NR.C81) are from radiopharmaceutical suppliers; and
(2) the number of industrial licensees in non-Agreement States, for
which data were not available, is approximately equal to the number of
licensees in Agreement States.
3.3-15
-------
Table 3.3-12. Fatal, cancer risks due to radlonuclide emissions from
the reference radiopharmaceutical industry facilities^3-'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
2E-7 (1E-7) 2E--5 (9E-6)
_Medigal_facili.ty
2E-8 (1E-8) 8E-7 (4E-7)
Sewage treatment ..... plant_
2E-10 (2E-10) 3E-8 (3E-8)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8S
Volume I, of this report.
Table 3.3-13. Radiation dose rates from radionuclide emissions
from a large radiopharmaceutical producer
Nearby individuals Regional population
(mrem/y) (person-rem/y)
Thyroid 3,3 4.1E+1
Data presented in the NRG survey (NRC81) showed that approximately
15 percent of industrial licensees in the survey handled 1-125, 3
percent handled 1-131, and less than 2 percent handled Xe-133 and
Tc~99m. Based on these figures and the above assumptions, the total
numbers of suppliers in the United States 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.
3.3-16
-------
Table 3,3-14. Fatal cancer risks due to radionucllde emissions from
from a large radlopharmaceutical producer^3'
,, Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Radiopharmaceutical
producer 6E-6 (3E-6) 7E-3 (3E-3)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low—LET radiations, as described in Chapter 8,
Volume I, of this report.
Assuming that available average emissions data (Tables 3.3-5 and
3.3~6) 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 Te-99m, 0.13 Ci/y.
Based on these emissions, releases from the reference facility
(Table 3.3-10) are 2.4 percent of the national total for 1-125, 7.7
percent for 1-131, 7.4 percent for Xe-133, and 3.5 percent for Tc-99m.
Assuming that the reference facility also causes equal percentages of
total health impact, the impact from all radiopharraaceutical suppliers
may then be calculated.
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 percent of
the total 1-125 emission, 1.4 percent of the total 1-131 emission, and
1.7 percent of the Xe-133 emission. Assuming that the reference
facility contributes equal percentages to the total health impact, the
impact from all medical facilities may be estimated.
3.3-17
-------
Sewage Tregtment PIant_s_
It has been estimated that approximately 4000 sewage treatment
plants in the U.S. employ sludge incineration or pyrolysis (TI79).
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
may then be calculated.
3-3.9 Existing Emj.ssion Standards and Air Pollution Controls
Suppliers and users of radiopharmaceuticals are either NRG or
Agreement State 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.
3.3.10 Supp 1 ementa 1 _Con.troj._ Techno1ogjr
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
(TI79). 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
3.3-18
-------
reductions achieved by adding supplementary controls to capture patient
xenon exhalations at the reference medical facility are shown in
Table 3.3-15.
Airborne radioactive iodine emissions may be controlled by using
an activated charcoal filter in an iodination box in conjunction with a
fume hood (DM80). An iodination box is used at some facilities for all
procedures involving the use of 1 mCi or more of radioiodine. Basically,
it is a box with two 5-inch-diameter portholes and a front opening door
for access during experimental work. With the filter filled with
activated charcoal, initial collection efficiencies between 90 and 100
percent have been measured.
The cost of an iodination box is $700-$2,000. The cost to adapt a
futne hood for charcoal filter use is $1000-^2,000. The annual costs to
replace the filter are approximately $70 for an iodination box and
approximately $5000 for a fume hood.
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 3.3-15.
3.3-19
-------
Table 3.3-15. Costs and risk reductions at adding supplemental
controls to reference radiopharraaceutical industry facilities
Type
of
control
Level of
control
Annual
cost
($1000)
(a)
Fatal cancer risk
reduction factor
Medical facility
No xenon
controls^)
Add retention
bags or xenon
traps
Sewage treatment plant
Dry cyclone and
scrubber^)
99.9
25.0
1E-3
Add HEPA filter
with preheater
and charcoal filter
99(0)
50.0
0.1
not include capital costs.
(b'Typical existing controls.
'c'Particulates.
iodines.
3.3-20
-------
Ba66 Baker P. J., Reactor Produced Radionuclides, in AEG Symposium
Series No. 6, Radioactive Pharmaceuticals, Conf 651111, April
1966.
Coa82 Cole L. W., Environmental Survey of the Manufacturing Facility,
Medi-Physics, Inc., Arlington Heights, Illinois, Oak Ridge
Associated Universities, Oak Ridge, Tennessee, January 1982.
Cob83 Corbit C. D., Harrington W, N., Higby D. P., Stout L. A., and
Corley J. P., Background Information on Sources of Low-level
Radionuclide Emissions to Air, PNL-4670, Prepared for EPA under
U.S. DOE Contract by Battelle Memorial Institute, September
1983.
DM80 Dames and Moore» Airborne Radioactive Emission Control
Technology. Prepared for EPA under Contract No. 68-01-4992,
1980.
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, B.C., Reprinted 1980,
FDA76 Food and Drug Administration, A Pilot Study of Nuclear Medicine
Reporting Through the Medically Oriented Data System, HEW (FDA)
76-8045, June 1976.
Fr82a Frame P. W., Environmental Survey of the New England Nuclear
Corporation, Boston, Massachusetts, Oak Ridge Associated
Universities, Oak Ridge, Tennessee, April 1982.
Fr82b Frame P, W., Environmental Survey of the New England Nuclear
Corporation, Billerica, Massachusetts, Oak Ridge Associated
Universities, Oak Ridge, Tennessees April 1982.
JFA84 Jack Faucett Associates, Impact of Proposed Clean Air Act
Standards for Radionuclides on Users of Radiopharmaceuticals.
Prepared for EPA under contract, October 1984.
Le79 Leventhal L., et al., Radioactive Airborne Effluents from the
Radiophannaceutical 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 II3 Jerusalem, Israel, 1980.
3,3-21
-------
REFERENCES
(Continued)
NRC79 Nuclear Regulatory Commission, Private Communication from
G. Wayne Kerr, NRC, Washington, D,C. , 1979.
NR.C81 Nuclear Regulatory Commission, A Survey of Radioactive Effluent
Releases from Byproduct Material Facilities, NUREG-0819, Office
of Nuclear Material Safety and Safeguards, NRC? Washington,
D.C., 1981.
Ro82a Rocco B, P., Environmental Survey of the Medi-Physics Facility,
South Plainfield, New Jersey, Oak Ridge Associated
Universities, Oak Ridge, Tennessee, January 1982,
Ro82b Rocco B. P., Environmental Survey of the E. R, Squibb and Sons
Facilitys New Brunswick, New Jersey, Oak Ridge Associated
Universities, Oak Ridge, Tennessee, March 1982,
TI79 Teknekron, Inc., Draft Final Report, A Study of Airborne
Radioactive Effluents from the Radiopharrnaceutical Industry,
EPA Contract No. 68-01-5049, March 1979,
3.3-22
-------
3 • 4 SMllLy£D_Souxc^_Jla£yy^a<;tuyneTs
3-4.1 GejD§Jial_Descxi£tjLon
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 (EP&19).
3.4.2 P roce s s : Desc r ij). t ion
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 NKC to have inventories of
radioactive materials in quantities ranging from ten Ci to as high as
100,000 Ci.
3.4.3 Emission ControlSystems
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.
3.4.4 Radionuclide Emissions
Each radiation source manufacturer handles a unique combination of
radionuclides; therefore, each site has unique emission characteristics.
Table 3.4-1 shows radionuclide emission data on eighteen manufacturing
sites; these data were taken from reports submitted to NRC.
3.4.5 Refer_enc^e__Facility
For this analysis, a reference facility was created by summing all of
the radionuclides emitted by the eighteen sites listed in Table 3.4-1.
Other parameters used in the analysis were assumed to be those of an
3.4-1
-------
Industrial zone In a suburban area adjacent to a major city in the
raidwestern United States. Table 3.4-2 describes the parameters of the
reference facility.
3.4.6 HealthImpactAssessmentofReferenceFacility
The estimated annual radiation doses from the reference facility for
individuals and population groups are shown in Table 3,4-3, Cancer risks
to nearby individuals and committed population fatal cancers are presented
in Table 3.4m4. Nearby individuals are located 500 meters north of the
source. The risk estimates include estimates which use a dose rate
effectiveness factor of 2.5, as described in Chapter 8, Volume I.
Because of the way in which the reference facility was artificially
created, the risk to nearby individuals estimated for the reference
facility is much higher than the actual 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 3.4-1.
3.4.7 Total Hea1th Impact
The estimated number of fatal cancers caused by all radiation source
manufacturers is the same as the reference facility, because of the way in
which the reference facility was created.
3,4.8 Exis t ing Em i s sion Standards and Air Pollution Contro1s
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.
3.4-2
-------
Table 3.4-1,
Radioriticlide emissions from radiation source
manufacturers (Co83)
Site Radionuclide
R 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
(Cl/y)
0.0
1.3
3B-1
5B-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
fable 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
3.4-3
-------
fable 3.4-3. Radiation dose rates £rom radionuclide emissions
from the reference radiation source manufacturer
Nearby individuals Regional population
Organ (mrem/y) (person-rem/y)
Average of all organs 0.22 8.4
Table 3.4-4. Fatal cancer risks due to radionuclide emissions
from the reference radiation source manufacturer^
Lifetime risk Regional population
Source
to nearby individuals (Fatal cancers/y of operation)
Reference facility 4E-6 (2E-6) 2E-3 (8E-4)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LEf radiations, as described in Chapter 8,
Volume I, of this report.
3.4-4
-------
REFERENCES
Co83 Corbit C. D., Harrington W. N., Higby D. P., Stout L. A., and
Corley J. P., Background Information on Sources of Low-level
Radionuclide Emissions to Air, PNL-4670, Prepared for EPA
under U.S. DOE Contract by Battelle Memorial Institute,
September 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.
3.4-5
-------
Page Intentionally Blank
-------
3.5 Oth£rjffietiees (Co83)
This section Includes NRC licensed laboratories,, low—level waste
disposal sites, and NRC-licensed mineral and metal processing
facilities.
3.5.1 General De_s_cri£ticm
NRC-licensed laboratories include test, research, and development
laboratories in industry, government agencies, and academic and
research institutions. Approximately 700 laboratories are licensed by
Agreement States to handle radioisotopes in an unsealed form. It is
assumed that an equal number of NRC licensees handle unsealed
radioisotopes j resulting in a total number of about 1,400 laboratories
that are possible sources of low-level radioactive airborne emissions.
Waste Disposal Sites
There are six commercial low-level radioactive waste disposal
sites, but only three of the sites, located at Barnwell, South
Carolina, Beatty, Nevada, and Richland, Washington^ are operational.
The remaining three, located at Maxey Flats , Kentucky, Sheffield,
Illinois, and West Valley, New York, are no longer operational.
The operational sites accept low-level radioactive wastes in a
stabilized form, but not special nuclear materials, transuranics, and
spent reactor fuels. Wastes accepted for disposal by shallow-land
burial must meet specific site acceptance criteria. The majority of
these wastes come from three sources: power-reactor operations,
laboratory research, and medical facilities.
Minja^aJ^jinj^^
Facilities which extract metals from thorium- and uranium-bearing
ores are licensed by NRC or an Agreement State. Six facilities,
located in California, Florida, Illinois, New Mexico, and Pennsylvania
(2 facilities), are licensed by NRC, and four facilities, located in
Alabama, Colorado, Oregon, stnd Tennessee, are licensed by Agreement
States. At facilities licensed by NRC, columbiutn and tantalum followed
by rare earth extraction, processes are the principal sources of
radioactive materials that require control under the present provisions
of 10 CFR 40. Two of the State-licensed facilities use thorium in
their manufacturing process and two process ore to recover rare earths
and refractory metals, respectively.
3.5-1
-------
3.5.2
Lab or a t or iej?
Laboratory facilities at a single site vary from a small multi-
purpose single laboratory up to 300 individual laboratories, located
within several buildings, at a major university. The smaller testing
laboratories tend to specialize in the limited use of radionuclides for
one purpose, such as soil testing or weld testing. Both academic and
industrial laboratories use byproduct materials in basic research and
development; radioactively labeled chemicals are used to trace a
metabolic or physical pathway through a system. Medical research
laboratories conduct basic chemical and applied radionuclide research
related to a broad spectrum of diseases and health problems.
Government laboratories may use radionuclides for specific purposes,
such as food and drug testing, water and air quality, and ocean and
fisheries monitoring. Thus, the testing and industrial laboratories
tend to use larger quantities but a more limited variety of
radionuclides than academic and other research laboratories.
A wide variety of radionuclides are found in laboratory work; the
most frequently encountered nuclides are tritium, carbon- 14, xenon- 133,
iodine-125, and iodine-131. The annual usage of any one radionuclide
rarely exceeds 10 Ci , and typically is less than 0,5 Ci .
Was t e Pi s po sal Sites
The disposal sites typically consist of a large fenced area of
about 100 ha. Operations buildings for decontamination , maintenance,
and waste preparation, are typically located at one end of the site,
Wastes are usually buried in the transport containers in which
they arrive at the site which minimizes radionuclide emissions to the
air .
Mi ner al and Me ta 1 Pro cess ing Faci litle s_
In general, most Agreement State and MC licensed facilities are
processing uranium- and thorium-bearing ores for either refractory
metals, their oxides (zirconium, columbium/niobium, tantalum and
hafnium) or for rare earths (cerium, praesodymium, neodymium,
dysprosium, ytterbium, etc.). Thorium is being used in licensed
facilities to manufacture welding rods and to cast machine parts.
The industrial processes used in licensed facilities may vary from
wet chemical and solvent extraction to high temperature sintering and
smelting. Raw ore storage, as well as sludge lagoons, drying beds, or
other waste storage facilities, may also be sources of radon arid thoron
emissions .
3.5-2
-------
3.5.3
The primary airborne emission controls employed by laboratories
are HEPA filters installed in fume hoods, hot cells, and glove boxes,
Laboratories which use one radionuclide predominately will frequently
have specific controls for that nuclide, such as activated charcoal
traps for xenon and iodine removal.
A catalytic recombiner followed by moisture removal is the
principal technology for removal of gaseous tritium from airborne
effluent streams. Chemical scrubbers may be used for removal of
carbon- 14.
Wa s_te |__Di s po_s aj^_S i t e s
Currently, the operating burial sites use compacted soil covering
to contain radioactive materials placed in the trenches. Despite
having up to 2.4 in of soil cover, some radionuclides may permeate
through the cover and enter the atmosphere. These low-level releases
may be in various chemical or physical forms. No emission controls,
beyond use of overburden, are currently used to minimize such releases.
Miner a 1_ and Mental Process ing Facilities
Information on controls to reduce airborne emissions of
radionuclides from NRC licensed facilities processing uranium- and
thorium-bearing ores is not available.
3.5.4
Data for 168 laboratories, including industrial, academic „
government, medical, and engineering, were obtained from two surveys of
byproduct users. Table 3.5-1 is a summary of the annual airborne
releases reported by these facilities. For purposes of population
exposure calculations , these emissions can be assumed to be at ground
1 eve 1 «
Haste Di.sposal___Si tes_
Radionuclide emissions from a nonoperational low-level waste
disposal are summarized in Table 3.5-2. To reduce subsurface migration
of radionuclides at this facility, the groundwater is pumped from sump
wells in the trenches to an evaporator. The water is evaporated and
the vapor is exhausted from an unfiltered 10-m stack. As can be seen
in Table 3.5~2, the primary radionuclide of interest is tritium, which
is emitted from the trenchwater evaporation system,
No data on operational sites were available,
3.5-3
-------
Table 3.5-1. Radionuclide emissions from laboratories (Co83)
Number of
facilities
103
45
9
35
35
Radionuclide
Tritium
Garbon-14
Krypton-85
Iodine-125
lodine-131
Xenon—133
All
Minimum Average Maximum Total
2.8E-1
6.9E-3
1.6E-1
4.0E-3
3.0E-4
8.0E-1
2.5E+1
1.1E-1
1.4
4.2E-2
5,9E-3
l.OE+1
2.9E+1
3.1E-1
1.4
1.4E-1
1.1E-2
1.6E+1
4.7E+1
Table 3,5-2. Radionuclide emissions from a nonoperational
low-level waste disposal site (Co83)
Radionuclide
Tri t ium
Carbon™14
Cobalt-58
Cobalt-60
Strontium-90
Ces ium"134
Cesium-137
Plutotiium-238
Plutonium~239
Trenches
8E-KL
5
Evaporator
6E+3
1.9E-4
5.8E-4
4.6E-4
2.IE-4
8.3E-3
1.1E-4
2.0E-6
The NRC-licensed ore processing facilities are not required to
report airborne radionuclide emissions. States having licensed
facilities uniformly report that airborne radionuclide levels are well
below values that require reporting. The limited data available
indicate some elevated radon levels in the immediate vicinity of sludge
lagoons; any effect on off-site radon levels was not obvious.
3.5-4
-------
3.5.5 -^----.^^-- Fac i 1 i ties
The emission rates listed in Tables 3.5-1 and 3.5-2 are quite low
except for tritium released from the low-level waste disposal site,
However, the whole body dose due to tritium released at this site is
estimated to be less than 10 mrem/y for nearby individuals. Dose
estimates for other radionuclides released from laboratories and the
low-level waste disposal site are less than 1 mrem/y to nearby
individuals. The limited data from NRC licensed ore processing
facilities indicate that off-site radon levels are within the range of
radon background concentrations,
3.5.6 Exist ing Bmiss ion Standards
Laboratories, low-level waste disposal sites, and uranium and
thorium ore processing facilities licensed by NRC or by Agreement
States are subject to the requirements of 10 CFR 20, Appendix B,
Table II.
3.5-5
-------
REFERENCES
Co83 Corbit C. D., Herringtoti W, N., Higby D. P., Stout L. A., and
Corley J. P., Background Information on Sources of Low-level
Radionuclide Emissions to Air, PNL-4670, Prepared for EPA under
U.S. DOE Contract by Battelle Memorial Institute, September 1983,
3.5-6
-------
3.6
3.6A Armed_Forc_e s _Rgj^obio1ogy Re s e a rch In s ti_tute ( AFRSI)
3.6A.1 Gener a1 De s c rigtion
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.
AFRRI's 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.
3.6A.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 investiga-
tions, studies of transient radiation effects on electronics, and
artifact analysis (ShSl).
The reactor cores 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 ra 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.
3.6A.3 Control Technology
Emissions from the AFRRI reactor and accelerator are released
to the atmosphere through a common stack atop the AFRRI building.
3.6A-1
-------
Partlculate emissions are controlled by a roughing filter, prefilters
and HEPA filter.
3.6A.4 Radionuclide Emissions^ Mea_sureingrits_
Annual airborne radionuclide emissions for AFRRI are shown in
Table 3.6A-1. These figures represent average annual emissions for
1981 and 1982 taken from the annual report to the NRC.
Table 3.6A-1. Radionuclide emissions from the Armed Forces
Radiobiology Research Institute
„,.,., Emissions'3'
Source Radionuclide (r'I )
AFRRI stack Argon-41 1.3
AFRRI stack Nitrogen-13, and
Oxygen~15 3.5E-2
'a'Average annual emissions for 1981 and 1982.
3.6A.5 HealthImpact Assessment of AFRRI
The estimated annual radiation doses resulting from radionuclide
emissions from AFRRI are listed in Table 3.6A-2. The distance from the
AFRRI facility to the nearest residence is approximately 200 meters.
These estimates are for a suburban site with a regional population of
2.5E+6 (Reference Site B). The nearby individuals are located 500
meters from the AFRRI facility for purposes of dose estimation.
Table 3.6A-3 lists the estimated lifetime risks to the nearby
individuals and the number of fatal cancers per year to the regional
population from these doses. The risk estimates include estimates
which use a dose rate effectiveness factor of 2,5, as described in
Chapter 8, Volume I.
3.6A.6 Existing Emission Standards and Air PollutionControls
The AFRRI reactor is licensed by NRC and is therefore subject to
the emission requirements of 10 CFR 20, Appendix B, Table II, which
limits air emissions to unrestricted areas. For argon—41, the isotope
responsible for all of the dose3 this limit is 4 x 10~8 microcuries
per milliliter above background.
3.6A-2
-------
Table 3.6A-2. Radiation dose rates from radionuclide emissions
from the Armed Forces Radiobiology Research Institute
, Nearby individuals Regional population
Organ / / \ i i \
(.mrem/y,) Iperson-rem/y)
Average of all organs 4.8E-3 1.7E-3
Table 3.6A-3, Fatal cancer risks due to radionuclide emissions from
the Armed Forces Radiobiology Research Institute^3)
„ Lifetime risk Regional population
Source ..,..,, , _ , , r . *
to nearby individuals (.Fatal cancers/y or operation;
AFRRI Stack 9E-8 (4E-8) 5E-7 (2E-7)
risk estimates in parentheses include a dose rate reduction
factor of 2,5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
3.6A.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.
3.6A-3
-------
REFERENCES
Sh8l Sholtis J. A. and Moore M. L.s Reactor Facility, Armed Forces
Radiobiology Research Institute, AFRRI Technical Report
T&.81-2, APRRI, Bethesda, Md., 1981.
3.6A-4
-------
3.6B.1
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 (FUR) 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 .
3.6B.2 P roc e s s De s c r i pjt J£ n
Both Array reactors are bare, unreflected, unraoderated , 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 (Aab82, AMT81). Operating information for
the APRF and FBR for 1981 is summarized in Table 3.6B-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 (Aab82).
Table 3.6B-1. Number and modes of operations at Army Reactor
Facilities, 1981 (Aab82, AMT81)
_ ,. . . Number of operations
Type of operation ""™^ FBR
Pulse 211 252
Steady State 233 159
Unscheduled Terminations - 8
Total 444 419
3.6B-1
-------
3.6B.3
Techncalogj
Air exhausted from U.S. Army reactor facilities is passed through
HEPA filters before release Co the atmosphere.
3.6B.4 Radionuclide Emission Measurements
Radioactive emissions from Army reactors during 1976S 1978; and
1981 are listed in Table 3.6B-2. For the APRF, participate releases
are reported as gross beta concentrations only. All gaseous releases
from the APRF were below the minimum detectable concentration of 3.GE-3
pCi/m3,
Table 3.6B-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, Aaa773 Aab82, AMT81).
3.6B.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 3.6B-3, The
distances to the nearest offsite 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).
Table 3.6B-4 lists the estimated lifetime risks to nearby
individuals and the number of fatal cancers per year to the regional
population from these doses.
3.6B-2
-------
Table 3.6B-3. Radiation dose rates from radionuclide emissions
from the White Sands Fast Burst Reactor
_ Nearby individuals Regional population
(mrem/y) (person-rem/y)
Endosteum
Spleen
Red Marrow
Muscle
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
Table 3.6B-4. Fatal cancer risks due to radionuclide emissions from
the White Sands Fast Burst Reactor^3)
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
FBR 4E-7 (2E-7) 2E-5 (9E-6)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
This assessment was made only for the White Sands FBR because nearly
all measured radionuclide emissions from Army reactors originate at the
FBR. The risk estimates include estimates which use a dose rate effec-
tiveness factor of 2.5, as described in Chapter 8, Volume I.
3.6B.6 Existing Emission Standards
Airborne emissions from Army facilities are limited by the require-
ments of Army Regulation 385-11, Chapter 5, and Army Technical Manual
3-261. These requirements establish airborne concentration limits
equivalent to NEC 10 CFR 20 concentrations, although the allowable
averaging periods are more restrictive.
3.6B.7 Supplemental Control Technology
Emissions from Army pulse reactors consist mainly of argon-41, for
which no demonstrated treatment technology exists. Reduction of argon-41
3.6B-3
-------
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.
3.6B-4
-------
REFERENCES
Aaa77 Aaserude R.A., Dickinson R. W., Dubyoski H. G. , and Kazi A. H.,
APRFj 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 Division Reactor, Annual Operating Report,
Materiel Testing Directorate, Aberdeen Proving Ground, Md.,
1982.
AMT81 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.
3.6B-5
-------
Page Intentionally Blank
-------
3 . SC U-
3 . 6C . 1 General. Description
Airborne emissions of radionuclides £rora 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
3.6C.2 Process Descr ipt ion
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 seaB beyond 12 miles from shore (Ri82) .
3 . 6C . 3 Cont ro 1 Techpology
All air exhausted from radiological support facilities at naval
shipyards is passed through HEPR filters and monitored during discharge.
h comparison of airborne activity measurements in shipyards with
radioactivity concentrations in ambient air indicates that air exhausted
from these facilities actually contains less activity than the intake air
(Ri82).
3.6C-1
-------
3. SC. 4 Rad lonuc I Ide., Emiss..JQrr ftgasu reitien t s
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 (EPA77).
Results of emission measurements taken at Puget Sound Naval Shipyard
in 1974 are shown in Table 3.6C-1, These measurements showed that the
tritium concentration was below the minimum detectable level of 1.0 pci/l,
and that the level of krypton-85 was within average background levels
(EPA77).
Table 3.6C-1. Radionuclide emissions at Puget Sound Naval Shipyard,
1974
„ ..,.,, Emissions
Source Radionuclide
West of Radiological
Support Building Krypton-85 17.4 + 10%
Radiological support
Building Tritium 0.4 + 50%
Radiological support
Building Tritium 0.3 + 66%
Radionuclide emissions from all naval shipyards were 0.41 Ci/y for
argon-41, 0.21 Ci/y for xenon-133, and 0.25 Ci/y for xenon-135; all other
radionuclide emissions were equal to or less than 0.1 Ci/y (Co83).
3.6C.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, Mr 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 3.6C-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).
3.6C-2
-------
3.6C.6 Health IftaAsessmettt oF
The estimated annual radiation doses resulting from radionuclide
emissions from the reference shipyard are listed in Table 3.&C-3. The
distance to the nearest offsite individual is approximately one km. The
predominant exposure pathway is that of ground shine. These estimates
are for a suburban site with a regional population of 2.5E+6 (Reference
Site B).
Table 3.6C-2. Radionuclide emissions from
the reference facility (Ri82)
Radionuclide
Argon-41
Cobalt-60
Tritium
Carbon- 14
Krypton-83m
Krypton- 85m
Krypton-85
Krypton-87
Krypton-88
Xenon- 13 1m
Xenon- 13 3m
Xenon- 13 3
Xenon- 13 5
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
Table 3.6C-4 presents estimates of the lifetime risks to nearby
individuals and the number of fatal cancers per year to the regional
population from these doses. The risk estimates include estimates which
use a dose rate effectiveness factor of 2.5, as described in Chapter 8,
Volume I,
3.6C.7 Total Health Impact of U.S. Nuclear Naval Shipyards
The total health impact caused by all naval shipyards may be
estimated from Table 3.6C-4 and the ratio of the capacity of the
reference shipyard to the capacity of all nuclear naval shipyards.
3.6C-3
-------
Table 3.6C-3, Radiation dose rates from radionuclide
emissions from the reference facility
_ Nearby individuals Regional population
(mrem/y) (person-rem/y)
Average of all organs 1.6E-2 8.7E-2
Table 3.6C-4. Fatal cancer risks due to radionuclide emissions from
the reference facility^'
„ Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Nuclear naval
shipyard 3E-7 (1E-7) 2E-5 (1E-5)
'a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
3.6C.8 Existing Emission Standards
Because Navy facilities are not licensed by NRC, they are not
subject to radionuclide emission standards.
3.6C.9 Supplemental Control Technology
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.
3.6C-4
-------
REFERENCES
Co83 Corbit C. D., Herrington W. N., Higby D. P., Stout L. A., and
Corley J. P., Background Information on Sources of Low-level
Radionuclide Emissions to Air, PNL-4670, Prepared for EPA under
U.S. DOE Contract by Battelle Memorial Institute, September
1983.
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.
3.605
-------
Page Intentionally Blank
-------
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 and 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 to generate
process steam/hot water, space heat, or electricity for in-house use are
designated industrial boilers.
From 1974 to 1977, about 18 percent of the energy needs in the
United States were met by burning coal; 66 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
(EPA80).
Coal contains mineral matter including trace quantities of
naturally-occurring radionuclides. Uranium-238 and thorium-232 and their
decay products are the radionuclides of interest with respect to air
emissions and potential health effects. Data showing typical uranium and
thorium concentrations in coal are presented in Table 4,0-1, The data
for "All Coals," given at the end of Table 4.0-1, represent more than
5,000 coal samples from the major coal producing regions of the United
States. DOB has analyzed uranium concentrations in more than 3,700 coal
samples and reports concentrations ranging from less than 2 to 130 parts
per million (ppm). These data (see Table 4.0-2) show about 71 percent of
all coals have uranium-238 concentrations less than or equal to 2.0 ppm,
and that 98 percent of all coals have uranium-238 concentrations of 10
ppm or less. Coal also contains the decay products of uranium-238 and
thorium-232 (see Tables 4.0-3 and 4.0-4) in secular equilibrium (Wa82).
Thus, the specific activity of each decay product is equal to the
specific activity of its uranium or thorium parent,
As coal is burned, the minerals in the coal melt and then condense
into a glass-like ash; the quantity depending on the mineral content of
the coal. A portion of the ash settles to the bottom of the boiler
{bottom ash) and a portion enters the flue gas stream (fly ash). The
partitioning of ash between bottom and fly ash depends on the type of
4.0-1
-------
fable 4.0-1. Typical uraniun and thorium concentrations in coal
Uranium
Region/
Coal Rank
Pennsylvania
Anthracite
Appalachian
Bituminous
NR
Bituminous
Bituminous
Illinois Basin
NR
Bituminous
Bituminous
Range Geometric
mean
(ppm) (ppm)
0.3
<0.2
0.4
0.1
0.3
0.2
0.2
- 25
- 11
- 3
NR
— 19
- 5
- 43
- 59
1.2
1.0
1.3
1.1
1.2
1.3
1.4
1.7
Thorium
Geometric
Range
mean
(ppm) (ppm)
2
2
1
0
<3
<0
.8 - 14
- 48
.8-9
NR
NR
.7-5
- 79
.1-79
4.7
2.8
4.0
2.0
3.1
1.9
1.6
3
Refer-
ence
Sw76
Sw76
IGS77
SRI77
Zu79
IGS77
SW76
Zu79
Northern Great Plains
Bituminous-
Subbituminous
Subbltuminous
Lignite
Western
NR
Rocky Mountain
Biturainous-
Subbituminous
Subbituminous
Bituminous
All Coals
<0.2
<0.1
0.2
0.3
0.2
0.1
0.1
<0.1
- '3
- 16
- 13
- 3
- 24
- 76
- 42
~ 76
0.7
1.0
1.2
1.0
0.8
1.9
1.4
1.3
<2
0
0
0
<3
0
<0
<0
"™ ft
.1-42
.3 - 14
.6-6
- 35
.1-54
.2 - 18
.1 - 79
2.4
3.2
2.3
2.3
2.0
4.4
3.0
3.2
Sw76
Zu79
Zu79
IGS77
SW76
Zu79
ZU79
Zu79
Note: 1 ppm uranium-238 is equivalent to 0.33 pci/g of coal.
1 ppm thorium-232 is equivalent to 0.11 pci/g of coal.
MR Not reported.
4.0-2
-------
boiler (see Section 4,1, Utility Boilers). The ash contains the
radionuclides originally present in the coal. Measurements of
radionuclides in bottom and fly ash show that certain radionuclides are
enriched in the fly ash relative to the bottom ash, particularly in the
respirable (less than or equal to 10 micrometers) fraction of the fly
ash (SmSO). The fraction of fly ash that is not captured by the
emission control equipment is released to the atmosphere. Thus, the
quantitiy of radionuclides released depends on the uranium and thorium
content of the coal, furnace design, enrichment factors for fly ash,
and the efficiency of the effluent control system for participates.
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.
Table 4.0-2. Uranium concentrations and distributions in coal
Uranium
concentration
(ppffl)
Less
2
4
6
8
10
12
14
16
18
20
30
60
than 2
- 4
- 6
- 8
- 10
- 12
- 14
- 16
- 18
- 20
- 30
- 60
-130
Number
of coals
analyzed
2669
666
207
67
39
26
17
12
7
5
9
5
2
Percent of coals
within uranium
concentration
range
71.5
17,9
5.5
1.8
1.0
0.7
0,5
0.3
0,2
0.1
0.2
0.1
0.05
Cumulative percent
of coals equal or
less than uranium
concentration range
71.5
89.4
94.9
96.7
97.8
98.5
98.9
99.2
99.4
99.6
99.8
99.9
100.0
Source: Fa79,
4.0-3
-------
Table 4.0-3. Major decay products of uranium-238
Radionuclide
Uranium-238
Thorium-234
Protactinium- 234m
Uranium-234
Thorium- 230
Radium-226
Radon- 222
Polonium-218
Lead-214
Bistnuth-214
Polonium- 214
Lead-210
Bismuth-210
Polonium-210
Source: Le67.
y = years d =
Table
Radionuclide
Thorium- 232
Radium- 2 28
Rctinium-228
Thorium-228
Radium-224
Radon- 220
Polonium- 216
Lead- 212
Bismuth-212
Polonium- 2 12
Thallium- 20 8
Ha If- life
4.5xl09 y
24 d
1.2 m
2.5xl05 y
S.OxlO4 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
days h
4.0-4. Major
Half- life
1.4xlOl° y
6.7 y
6.1 h
1.9 y
3.6 d
55 s
0.15 s
10 h
60 m
3.1xl0^7 s
3.1 m
Rlpha
4.20
4.77
4.68
4.78
5.49
6.00
7.69
5.31
= hours
Principal radiation
Beta
(max)
0.191
2.29
0.65
1.51
0.015
1.160
m = minutes
(Mev)
Gamma
0.093
1.001
0.186
0.352
0.609
0.047
s = seconds
decay products of thorium- 232
Alpha
4.01
5.43
5.68
6.29
6.78
8.78
Principal radiation
Beta
0.055
1.11
0.589
2.25
1.80
(Mev)
Gamma
0.908
0.084
0.241
0.239
0.727
2.614
Source: Le67.
y = years d = days
h = hours
m = minutes
seconds
4.0-4
-------
REFERENCES
EPA80 Environmental Protection Agency, Fossil Fuel-Fired Industrial
Boilers--Background Information for Proposed Standards,
Chapters 3-5, Research Triangle Park, N.C., June i960.
Fa79 Facer J.F., Jr., 1979, Uranium in Coal, U.S. Department of
Energy Report, GJBX-56(79), Washington, D,C.
IGS77 Illinois State Geological Survey, Trace Elements in Coal:
Occurrence and Distribution, MTIS Report No. PB-270-922, June
1977.
Le67 Lederer C.M., Hollander J.M. and I, Perlman, fable of isotopes,
Sixth Edition, John Wiley and Sons, New York, 1967.
SraSO 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.
SRI77 Stanford Research Institute, Potential Radioactive Pollutants
Resulting from Expanded Energy Programs, NT1S Report No.
PB-272-519, August 1977.
Sw76 Swanson V.E., et al., collection, Chemical Analysis, and
Evaluation of Coal samples in 1975, Department of the Interior,
Geological Survey, Open File Report 76-468, 1976.
Wa82 Wagner P. and N.R. Greiner, 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,
Zu79 Zubovic P., et al.r Assessment of the Chemical Composition of
Coal Resources, USGS Expert Paper Presented at the United
National Symposium on World Coal Prospects, Katowice, Poland,
April 15-23, 1979.
4.0-5
-------
Page Intentionally Blank
-------
4.1 u 1 1 1 1 ty Boi 1 e r s
4.1.1 Genera^J)escription
At the end of 1979, the total capacity of U.S. electric utility
generating units was 593 gigawatts (GW) (TRl79a). 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 (DOE79).
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. Host of the units with
capacities of 26 to 100 KW 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 (TR!79a>. 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
Geothermal
(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
Total
2.0
(.3)
593.1 (100.1)*
7.9
766.0
(1.0)
(100,0)
Source: (TRl79a).
*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
(TRI79b, DQE74).
4.1.2 Process Descri^t_i.on
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
-------
[.1-2. Distribution of U.S. coal-fired units by age
and capacity, 1979
Capacity of coal-fired units
Age
0.03-0.1 GW
(units) (GW)
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
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^
(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: (TRI79a) .
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, and 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).
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
(MW)
2,015
24,449
16,440
161,092
203,996
Percent
of total
(1.0)
(12.0)
(8.0)
(79.0)
Source: (DOE76).
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 Furnaces. 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 (Coc75).
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 Ope-cation
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(i) for coal-fired units
operating in the base load mode was 65 percent; for units operating in
a cycling mode, 42 percent (TRI79a). The availability(2) Of a
coal-fired unit generally declines with increasing generating
capacity. Generating units with capacities of less than 400 J$W have
average availabilities of more than 85 percent; those with capacities
of more than 500 MW, only 74 to 76 percent (An77), 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 (Dea77, Deb79, St76, Cob?7).
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.
^•2/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 o£ a wet
scrubber systew or downstream of a spray dryer system,
BSPs 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.
fable 4.1-4 shows the distribution of control equipment in use In
1976 on coal-fired steam electric boilers (DOE76).
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
(SrnSO), 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.
Particulate emission control equipment
by type of boiler, 1916
Control
equipment
No control
Mechanical'3^
Stoker
Capacity
(GW>
0.7
0.8
Units
76
63
Pulverized cyclone
Capacity . ,
I In ire
ni S
4.5 66
0.5 11
Wet scrubbers
Fabric filters
ESP
0.1
19.0
ibination 0.4 24 0.4
62
7
9.5
2.0
44
14
Control
equipment
No control
Mechanical ^
Wet scrubbers
Fabric filters
ESP
Comb ina t ion ' b )
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.
^Combination refers to mechanical-electrostatic precipitators.
Source: (DOE76).
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 participate release is
directly related to coal
consumption, which in turn depends
on heat rate.
Capacity
Total participate 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.
1.1-8
-------
Table 4.1-6. Radionuclide emission rates (mCi/y) measured at
selected coal-fired steam electric generating stations
Sampling location^3)
Radionuclide M-l M-2 M-3 M-4
Uranium- 2 38
Uranium-234
Thorium-230
Radium-226
Lead-210
Polonium-210
Thorium-232
Thorium-228
24
24
1.5
5.3
28
68
0.81
0,72
5.7
7.2
4.1
4.1
15
14
1.5
1.7
0.76
0.81
0.29
0.21
1.4
1.1
0.02
0.30
0.10
0.10
0.08
0.02
0.18
0.16
0.05
0.05
Source: EPA80
'a'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).
Table 4.1-7. Summary of radionuclide emission rates (mCi/y)
measured at five coal-fired steam electric generating units
Sampling location'3-'
Radionuclide
Uranium-238
Thorium-230
Radium-226
Polonxum-210
Lead-210
Thorium-232
M-l
120
17
63
1000
340
8.5
M-33
14
8.8
12
22
30
8
M-15
2
0.3
0.5
<0,6
(b)
0.1
M-34
1
1
3
3
24
1
M-99
0.06
0.06
0.1
< 3.0
(b)
0.06
Source: (Ro83).
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.
4.1-9
-------
Table 4.1-10,
Radiation dose rates from radionuclide emissions
from the reference utility boiler
Urban site
Organ
Lung
Red marrow
Kidney
Endosteum
Liver
Nearby
individuals
(mrem/y)
9.9E-1
l.OE-1
l.OE-1
1.2
5.0E-2
Regional
population
(person-rem/y)
4.2E+3
3.3E+2
1.3E+2
4.6E+3
7.1E+1
Suburban site
Nearby
individuals
(mrem/y)
1.3
2.1E-1
2.3E-1
1.8
1.4E-1
Regional
population
(person-rem/y )
3.9E+2
4.3E+1
8.0E+1
5.01+2
2.7E + 1
Rural site
Organ
Lung
Red marrow
Kidney
Endosteum
Liver
Nearby
individuals
(mrem/y)
1.7
2.1
2.4
4.7
1.9
Regional
population
(person-rem/y)
1.1E+2
1.9E+1
3.0E+1
1.4E+2
1.4E+1
Remote site
Nearby
individuals
(mrem/y)
1.2
1.4E-1
8.6E-2
1.4
6.4E-2
Regional
population
(person-rem/y)
1.1
1.5E-1
2.6E-2
1.7
1.1E-1
Table 4.1-11 presents estimates of lifetime risks to nearby
individuals 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 risk estimates include
estimates which use a dose rate effectiveness factor of 2.5, as
described in Chapter 8, Volume I.
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 b"y EPA, estimated annual
radiation doses were calculated for nearby individuals and regional
population (Tables 4.1-12 and 4.1-13).
4.1-12
-------
Table 4.1-11. Fatal cancer risks from the reference
Site
Lifetime risk
to nearby individuals
Regional population
(Fatal cancers/y of operation)
Urban
Suburban
Rural
Remote
2E-6
4E-6
3E-5
3E-6
(3E-6)
(1E-5)
(2E-6)
1E-1
1E-2
5E-3
3E-5
(3E-3)
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume 1, of this report.
Organ
Table 4.1-12. Radiation dose rates to nearby individuals
from radionuclide emissions from five utility boilers
Nearby individuals (mrem/y)
M-l
M-33
M-15
M-34
M-99
Lung
Red marrow
Kidney
Endosteum
Liver
8.8E-1
1.3E-1
4.2
3.0
1,7
5.2E-1
3.9E-1
3.5E-1
1.1
3.2
1.5E-1
8.6E-3
5.5E-3
6.6E-2
4.5E-3
5.5E-2
7.1E-2
6.7E-2
1.9E-1
7.2E-2
1.1E-2
4.5E-3
1.3E-2
2.0E-2
6.5E-3
Table 4,1-13. Radiation dose rates to the regional population
from radionuclide emissions from five utility boilers
Regional population (person-rem/y)
Lung
Red marrow
Kidney
Endosteum
Liver
M-l
8.8E+1
1.3E + 1
9.1E+1
5.2E + 1
2 . 2E +1
M-33
2.6E+1
4.3
3.6
4.0E-H
2.2
M-15
3.7
1.5E-1
1.1E-1
1.5
6.1E-2
M-34
2.4
5.8E-1
6 . 4E -1
4.2
5.4E-1
M-99
3.6E-1
6.0E-2
3.1E-1
4.5E-1
9.3E-2
4.1-13
-------
Table 4.1-14 presents estimates of lifetime risks to nearby
individuals 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. The risk estimates include estimates
which use a dose rate effectiveness factor of 2.5, as described in
Chapter 8, Volume I.
Table 4.1-14. Fatal cancer risks from radionuclide emissions
from five utility boilers^3'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
M-l
M~33
M-15
M-34
M-99
2E-5
6E-6
3E-7
1E-6
7E-8
(9E-6)
(3E-6)
(5E-7)
(4E-8)
4E-3
IE -3
1E~4
IE -4
1E-5
(3E-3)
(SE-4)
(9E-5)
(9E-5)
'a'The risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low-LET radiations, as described in Chapter 8,
Volume I, of this report.
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. Thus, the health impact from the industry may be
estimated using this information and Table 4.1-11.
4.1.9 Existing Emission Standards and__Air Pollution ....Controls
There are no radionuclide emission standards for utility boilers.
However, particulate emission rates are regulated by EPA and the States.
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
4.1-14
-------
limited to 43 ng/J. The 1979 revised New Source Performance Standards
CRNSPS), which apply to all 73 Mtf(t) or 25 MW(e) electric utility steam
generating units on which construction began after 19 September 1918,
require that participate emissions be limited to 13 ng/J (TRI81).
States regulate participate 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). 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 through 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 4.1-15 shows how these reductions can be related to
population density. The reduction in uranium-238 emissions may be
estimated by assuming a uranium-238 concentration of 9 pci/g in fly ash
emitted to the air. These reductions are listed in Table 4,1-16.
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.
4.1-15
-------
fable 4.1-15, Relationship of particulate emissions reduction to
population density
Population
density^3)
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 mi 11 Ion- 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
rig/J
Total 268,100 81.8
(a>Population within 80 km of a coal-fired utility boiler.
Source: (RC83).
Table 4.1-16. Reduction in uranium-238 emissions caused by
reducing particulate emissions
Population Reduction in uranium-238^
density emissions (Ci/y)
0-50,000 0,2
50,000-100,000 0.2
100,000-250,000 0,2
250,000-500,000 1.2
500,000-1 million 2.3
1 milllon-2.5 million 2
2.5 mllllon-5 million 0.9
5 million-ID million 0.4
Total 7
(aj-jhese 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 - 10^ pci).
4.1-16
-------
REFERENCES
An77 Anson D», 1977,'Availability of Fossil-Fired Steam Power
Plants, Electric Power Research Institute, EPRI-FP-422 SR,
Palo Alto, California, January 1977.
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 HI, Houston, Texas,
April 23-28, 1978.
Coa78 Coles D.G., Ragaini R.C., and Ondov J.M., Behavior of Natural
Radionuclides in Western Coal-Fired Power Plants, Environmental
Science & Technology, Volume 12, 14_, April 1978,
Cob77 Considine D.M. (Editor-in-Chief), 1977, Energy Technology
Handbook, McGraw-Hill Book Company, New York.
Coc75 Cowherd C., et al.s 1975, Hazardous Emission Characterization
of Utility Boilers, NTIS Report No. PB-245-915, July 1975.
Dea77 Dennis R., et al., 1977, Filtration Model for Coal Fly Ash
with Glass Fabrics, EPA 600/7-77-084, U.S. Environmental
Protection Agency, Research Triangle Park, North Carolina.
Deb79 Dennis R. and K.A. Klemm, 1979, Fabric Filter Model Format
Change, Vol. 1, Detailed Technical Report, EPA 600/7-79-0432,
U.S. Environmental Protection Agency, Research Triangle Park,
North Carolina.
DOE74 Department of Energy, 1974S Cost and Quality of Fuels for
Electric Utility Plants, Energy Data Report, FPC Form 423,
DOE/EIA-0075/9C79), September 1974.
DOE76 Department of Energy, Computer Tape Containing Federal Power
Commission Form 67 Data for 1976, Energy Information
Administration, Washington, D.C.
DOE77 Department of Energy, 1974, Cost and Quality of Fuels for
Electric Utility Plants, Energy Data Report, FPC form 423,
DOE/EIA-0075/9(79), September 1974.
DOE79 Department of Energy, 1979, Trends in the Capacity Utilization
and Fuel Consumption of Electric Utility Power Plants,
1970-1978, Energy Information Administration, Washington, D.C.
4.1-17
-------
REFERENCES (Continued)
EPA8Q 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, B.C., 1980.
EPA81 Environmental Protection Agency, The Radiological Impact of
Coal-fired Industrial Boilers, EPA, Office of Radiation
Programs, Washington, D.C., (Draft Report), 1981.
1GS77 Illinois State Geological Survey, 1977, Trace Elements in
Coal: Occurrence and Distribution, NTIS Report No. PB-27G-
922, June 1977.
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.
Ro83 Roeck D.R., White M.O., Kiddie A.M., and Young C.W., Survey of
Five Utility Boilers for Radionuclide Emissions, Prepared for
EPA by GCA Corporation under Contract No. 68-02-3168, December
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.
St76 Stern A.C., Editor, 1976, Air Pollution, Vol. IV, Third
Edition, Academic Press, Inc., New York.
TRI79a Teknekron Research, Inc., Utility Simulation Model
Documentation, Vol. 1, R-001-EPA-79, Prepared for the
Environmental Protection Agency, Washington, B.C., July 1979.
TRI79b Teknekron Research, Inc., Utility Simulation Model, Summary
Resource and Residuals Report, Berkeley, California, 1979.
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.
4.1-18
-------
4.2 irtdustrial Bedlers
4.2.1 General PescrigttQii
Coal-fired industrial boilers (CPlBs) 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 1914, more than 8?
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
Coal consumption
IndUStry (Percent of total)
Chemicals 33
Paper 26
Steel and aluminum 28
Food 10
Other manufacturing 3
Source: (KPR80).
4.2.2 Process Des.cription
Types of .Boil_ers
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 Mtf 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 Mtf 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 MM,
4.2-1
-------
Table 4.2-2 lists the number of boilers and their total installed
capacity (BPR81). 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 ^^ 15_3Q ^^ ?
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
Cga1~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 size 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.
Pu1verized coa1-fired boi 1e r s
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.
[.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:
Boilerfiring method. The type of firing is the most important
factor in determining ash distribution. Stoker^fired units emit
less fly ash than pulverized coal-fired boilers,
Wet ordry 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
-------
C2) the charged particles migrate to a collecting electrode of opposite
polarity where they are collected, and (3) the collected particulates
dislodged from collecting electrodes. Energy is needed to operate
the preclpltator 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 efficiencies.
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.
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
its capacity, characteristics of coal consumed, and level of
partlculate natter 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 BSPs 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 particles.
4.2.4 Hadionuc1ide Bmissions
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.
4.2-5
-------
fable 4.2-5 lists additional estimates of uranium-238 emission
rates from representative coal-fired industrial boilers. The estimates
of particulate emissions reflect the range of emissions that
characterize the entire population of industrial boilers, and the
uranium-238 emissions are calculated using average ¥alues for coal and
boiler characteristics (e.g., heat value, capacity factors) that
influence emission rates (TRI81).
Table 4.2-5. Estimated uranium-238 emission rates
for representative coal-fired industrial boilers
Boiler capacity
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)
IE- 4
4E~4
3E-4
IE- 3
4E-4
6B-3
7E-4
7E-3
9E-4
4E-4
We estimate the uranium-238 emission rate for the entire
population of large (15 MH 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 4.2-6}
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 combina-
tions 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.
4.2-6
-------
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
4.2-1 lists other characteristics of the reference boiler used in the
health impact assessment.
Table 4.2-6. Radionuclide emissions from the reference boiler
Radionuclide
Emissions
(Ci/y)
Uranium series:
Uranium-238
Uranium-234
Thorium-230
Radiura-226
Radon-222
Lead-210
Po-212
l.QB-2
l.OE-2
5.0E-3
1.5E-3
2.5E-1
2.5E-2
2.5E-2
thorium series;
Thorium-232
Radium-228
Actinium- 228
Thorium-228
Radium-224
Radon- 220
Lead-232
Bismuth-212
Thallium-208
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
Table 4.2-7. Reference coal-fired industrial boiler
Parameter
Value
Site
Population
Midwest location (St. Louis)
2.5 million people within
80 km of the site
Stack
Effective height
Diameter
150 meters
1.5 meters
4.2-7
-------
4.2.6 He_altly_. impact. g.§§g.g§E§ilL g£ I§£gI§Bg,t iMugirJ-Aj- Boiler,
The estimated annual radiation doses from the reference industrial
boiler are listed in Table 4.2-8. Table 4,2-9 presents estimates of
the lifetime risk of fatal cancer from these doses.
4.2.7 TotaJ._H_eaIth Impact of Coal-Fired industrial Boilers
The total number of fatal cancers caused by all coal-fired
industrial boilers may be estimated by multiplying the health effects
for the reference boiler by the ratio of the total (estimated)
urariium-238 emissions of the entire CF1B industry and the reference
boiler.
Table 4.2-8. Radiation dose rates from radionuclide emissions
from the reference industrial boiler
Nearby individuals Regional population
rga (mrem/y) (person-rem/y)
Lung
Red marrow
Kidney
Bone
Liver
3.4E-1
4.0E-2
4.0E-2
4.3E-1
2.QE-2
7.6E-H
6.6
9.0
9.0E-H
3.2
Table 4.2-9, Fatal cancer risks due to radionuclide emissions
from the reference industrial boiler^
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Industrial boiler 6E--7 (5E-7) 3E-3
risk estimates in parentheses include a dose rate reduction
factor of 2.5 for low- LET radiations, as described in chapter 8,
Volume I, of this report.
4.2-8
-------
No Federal or state regulations currently exist that limit
emissions of radionuclldes from coal-fired Industrial boilers.
However, the states, through State Implementation Plans (SIPs), and the
Federal government, through Mew Source Performance Standards (NSPS) ,
regulate particulate matter emissions and thus effectively limit
radionucllde emissions.
existing coal-fired industrial boilers are subject to SIPs.
Since the individual SIPs reflect local conditions and needs,
particulate matter emissions vary frora state to state,
All new coal-fired, industrial boilers with capacities greater than
73,3 HW (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 13 MW are subject to limits prescribed by the SIPs.
4.2,9 SuK|?lemental..,_Cont,rol Tgctoojogy
Currently, large coal-fired industrial boilers {15 MH and
greater), which are subject to SIP particulate matter limits, emit
about 0.37 million tons o£ particulate matter per year. Table 4,2-10
lists the costs, particulate matter emission levels, and
cost-effectiveness to retrofit large _ boilers to meet specific uniform
emission levels (RC83) .
Table 4.2-11 lists estimated uranium- 238 emissions for existing
and retrofitted large boilers (15 KW and larger) subject to SIP
particulate matter control.
Table 4,2-12 lists estimated current risks and risk reductions for
particulate matter limits for large (15 MH and greater) coal-fired
industrial boilers,
4.2-9
-------
Table 4.2-10. 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
MW and greater boilers.
Table 4.2-11. Estimated uranium-238 emission rates for existing
and retrofitted large coal-fired industrial
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
5 MW and greater.
4.2-10
-------
Table 4.2-12. Risks associated wlth>large coal-ftred
industrial
Particulate matter Risk reduction
control level factor
Various under SIPs 1
0.1 0.3
0.05 0,4
MW and greater.
4.2-11
-------
REFERENCES
EPA80 Environmental Protection Agency, 1980, Fossil Fuel-Fired
Industrial Boilers—Background Information for Proposed
Standards, Chapters 3-5, Research Triangle Park, M.c.
EPA81 Environmental Protection Agency, The Radiological Impact of
Coal-Fired Industrial Boilers (Draft), EPA Office of Radiation
Programs, Washington, D.C., October 1981.
TRI81 feknekron Research, Inc., Draft Background Information
Document for Coal-Fired Industrial Boilers, Unpublished, May
1981. (Available in EPA Docket R-79-11.)
RC83 Radian Corporation, Boiler Radionuclide Emissions Control:
The Feasibility and Costs of Controlling Coal-Fired"Boiler
Particulate Emissions, Prepared for EPA, January 1983.
4.2-12
-------
Chapter 5: URANIUM MINES
5.1 Gene r a I Pgsc rijat ion
In uranium mining operations, ore is removed from the ground in
concentrations of 0.1 to 0.2 percent l^jOg 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 radionuclide bearing dust and radon- 222 gas.
Uranium is mined in both open pit and underground mines. In 1982
there were 139 underground and 24 open pit uranium mines in operation
in the United States (Table 5-1). These mines accounted for about 75
percent of the uranium produced (DOE83).
Table 5-1. Distribution of 1982 U3Og production
by mining method (DOE83)
Source
Underground mines
Open pit mines
Solution Mining
(In-Situ)
Others:
heap- leach,
mine water,
byproduct , and
low-grade stockpiles
Total
Number
139
24
18
15
196
Tons U_OQ
O O
6,200
3,900
1,500
1,800
13,400
Percent
of total
46
29
11
14
100
-------
In recent years in-situ solution mining has been more widely used;
this method Is expected to increase in future years. During 1982 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 mldwestern parts of the United States.
Table 5-2. Distribution of 1982 U308 production
by State (DOE83)
State 3 8 Percent of
(Short tons) total
New Mexico 3,800 28
Wyoming 2,700 20
Texas 2,200 17
Arizona, Colorado, Florida,
Idaho, Utah, & Washington 4,700 35
Total 13,400 100
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 (EPA83a).
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, ft 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 gra¥itf 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 200,000 cfm.
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 its decay products) emissions
result from surface operations at the underground mine.
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 raining operation. As the
mining progresses, mining and reclamation operations take place
simultaneously — pits are rained In sequence, and the tnined-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 its decay products.
IrirS_itu Hining
In this method, a leaching solution is injected through wells into
the uranium- bear ing ore body to dissolve 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
impoundments. With solution raining, less than 5 percent of the radium
from an ore body is brought to the surface (NRC80). Consequently, the
amount of radon released is considerably less than that from conven-
tional raining. The major sources of radon are the surge ponds, enclosed
surge tanks, inplant surge tanks, and absorption columns (Br81). It is
estimated that the radon released is about 19 percent of the amount
released from a conventional uranium mill (Bra81).
5-3
-------
GENERALIZED UNDERGROUND URANIUM MINE
MODfFIED ROOM AND PIUAR MiTHOD OF MINING
Figure 5-1. An underground uranium mining operation.
-------
Figure 5-2. An open pit uranium mining operation.
-------
A small amount of radon is released from the waste impoundments
used to store contaminated liquids from the operation. examples
of solid wastes
-------
3) The amount of sealants used varied considerably for different
mines. Kown and his associates (K0S0) 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
has a limited life of about eight months because the drift area is
mined after pillars are extracted in a room-and-piliar 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 $348,100 ($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 (Fr81) which is comparable to the cost estimates ($0.66 per
square foot) of the sealants used in this study. Because of its high
cost, the Bureau of Mines feels that sealants may only be used
economically in shops, lunchrooms, and possibly high-emanating areas in
intake airways (Frc83).
A recent study by Battelle (B184) of 13 mines shows an average
cost of $5.80 per ton of ore mined ($0.34 per square foot) if 80 per-
cent of the surface is sealed. This EPA-sponsored 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 (Fra81) have shown that 50 to
75 percent of the radon can be retained in the rock by sealants. The
study by Battelle (B184) shows that a 56 percent reduction in radon
emissions can be achieved by applying sealant to 80 percent of the mine
surfaces.
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
5-7
-------
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
ventilation pipe is used as a bleeder pipe to provide negative pressure
in the extracted stope (Fra81) and to allow the contaminated air to be
diverted to the 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. In a Battelle study (B184)
the average cost to bulkhead 80 percent of the mine at 13 sites was
only $0.08 per ton of ore. Up to 10 bulkheads in each mine were used
in making these estimates.
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 (KoSO). A preliminary study conducted by Battelle on an
actual mine indicated that a radon reduction of 35 percent could be
obtained by using bulkheads (DraSO, Th81). Using bulkheads extensively
in a mine can reduce radon emissions up to 60 percent (B184).
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 (KoSO).
An effective radon control system for the bleeder pipes is still
under study. The system chosen by investigators in an EPA study (KoSO)
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.
5-1
-------
, RADON
I MONITOR
PRIMARY CARBON BED
FILTER BLOWER | (1800 lb«>
100 - 300JCFM
BLOWER
<1-5HP> HEATER
(50KW)
| COOLER
BLOWER
(2HP)
- -to CARBON BED |
(ISOIbs)
CONDENSED
WATER TO
DRUMS
SPARE
PRIMARY
ADSORPTION
CARBON
REGENERATION &
SECONDARY
ADSORPTION
Figure 5-3. Radon removal from mine air by carbon adsorption.
5-9
-------
In evaluating control technology in a model mine, EPA (KoSG) 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 bulk-
heads. The capital and operating costs for each unit are as follows:
Capital Cost ofEachUnit
Major equipment $22,000
Auxiliaries & Installation $11,OOP
Total $33,000
annUQlized Cost of Bach_..Unit
Material (carbon, filters, piping) $ 1,000
Utilities (25,000 kwh @ 4s2/kwh) $ 1,000
Labor (0.25 person-year) $ 8,000
amortizing (an avg. 5-year life at 10 percent interest) $ 8,700
Total $18,700
assuming the lifetime of each unit is 5 years and 12.5 units per
year are needed, the cost over five years would be $1,037,500 or $0.86
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 $1.20 per ton of ore mined for
100 bulkheads. In the study by Battelle (B184) the average cost to
bulkhead with a carbon trap at 13 mines was $0.11 per ton of ore with
an efficiency of 80 percent. Up to 10 bulkheads in each mine were used
for their estimate.
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 may not work in wet mines
because of moisture absorption by the carbon.
The system does appear to be technically feasible utilizing
commercial carbons and standard equipment. However, additional
developmental work may be necessary before such a system can be used in
a mine environment. A recent study by Hopke (Ho84) has concluded that
activated carbon can be used for effective cleaning of small volumes of
air such as effluents from a bleeder pipe for a bulkhead.
5-10
-------
Mine,,,, Pr es_sur izat Ion
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 Fra81). In 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 tests by the Bureau of Mines (Fra81)t
the radon levels in the mine were reduced by 20 percent (releases to
the atmosphere were not determined). The surrounding soil needs to be
permeable enough to hold radon and allow for its decay, but not so
permeable so as to allow significant increases in surface emissions
(Ko80). The costs of mine pressurization were not available because
the process was in a development stage, in a recent report, Battelle
(B184) concluded that positive pressure ventilation has been proven
ineffective in reducing atmospheric emissions of radon.
Miscellaneous Radon ControlTechnology
Argonne National Laboratory is experimenting with strong oxidizing
agents, such as bromine triflouride and dioxygenyl hexaflouro-antimonate,
to convert the radon to another form that can be absorbed on a scrubber
or absorption bed (FraSl). However, the corrosive and toxic nature of
the reactants makes their use in mines impracticable and questionable.
Battelle (B184) mentions other methods such as cryogenic methods,
chemical removal, and gas centrifuge, but the costs are prohibitive.
The study by Hopke (Ho84) reviewed methods for the removal of radon
from uranium mine effluents. Methods, including cryogenic condensation,
molecular sieves, gas centifugation, semipermeable membranes, and hybrid
systems, do not offer much promise for a practical removal system. They
do suggest the exploration of the class of perfluorinated hydrocarbon
compounds as possible candidate scrubbing fluids for a radon scrubbing
system.
Backfilling of worked-out areas with classified mill tailings is
practiced by mine operators to provide ground support in the mine
(Fr81). This procedure can also reduce ventilation requirements. R
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 net radon reduction of 84 percent from
the stope (FrbSl). This was done for only one stope in a mine. PNL
(B184) estimated an efficiency up to 80 percent if classified mill
tailings and surface sands are used for backfilling with an average
cost of $12.64 per ton of ore mined.
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
5-11
-------
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 tee miles from the source." An estimate of cost
is $0.493 - $0.881/ton of ore for a 20-meter stack (Brb84). The
average number of vents for a mine is about 5 (Ja80). Thus, the cost
per mine would be about $3.44 per ton of ore produced.
Vent orientation is an important factor in radon concentrations
near a mine (Drc84). Because of plume rise, concentrations are much
lower when vents are in a vertical configuration (rather than hori-
zontal), resulting in a reduction factor of 80 at sites near a mine
with a vertical vent configuration.
Summaryof Costs and Efficiencies
A summary of the costs and efficiencies of the various radon
control technologies discussed previously is shown in Table 5-3.
Table 5-3, Cost and efficiencies of radon control technologies
for underground uranium mines
Radon reduction Cost
Method (Percent) ($/ton of ore)
Sealant coating
Bulkheading
Bulkheading with activated
carbon
Mine pressurization
Stacks
Backfilling
11 - 56
14 - 60
49 - 80
20
60(a)
80 - 84
1.45
0.08
0.11
-
3.44
12.6'
- 5.80
- 0.34
- 1.20
1
'a'Reduction in exposure to nearby individuals.
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 o£
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 14 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 (fable 5-4) (JaSQ). The
average radon-222 emission rate for these 27 mines was 5,600
-------
Table 5-4. 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
M
BB
CC
DD
BE
FF
GG
HH
II
Average
Number
of vents
4
6
4
2
14
13
5
10
11
9
4
8
8
5
3
2
1
3
2
5
3
2
5
3
3
2
2
5
Measurements {Ci/y}
1979
7,400
4,700
5,200
3,600
29,800
9,200
2,200
15,200
1,700
7,800
7,000
1,500
15,000
1,900
900
1,000
17,500
-
2,100
2,100
_
-
6,500
2,500
200
1,000
500
6,100
1978
-
4,300
3,900
-
-
9,500
1,500
-
-
8,100
5,900
1,300
14,600
-
-
-
-
2,600
1,500
1,800
2,100
1,000
-
-
100
-
—
4,200
Average
7,400
4,500
4,600
3,600
29,800
9,400
1,800
15,200
1,700
7,900
6,400
1,400
14,800
1,900
900
1,000
17,500
2,600
1,800
2,000
2,100
1,000
6,500
2,500
200
1,000
500
5,600
curies/year. The emissions from individual vents ranged from 2 to 9,000
Ci/year with an average of 1,000 Ci/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
5-13
-------
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),
Table 5-5. Estimated annual radon-222 emissions from underground
uranium mining sources (BPA83b>
Source Average large
(Ci/y)
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 3,829
grade = 0.1 percent 1)303. Annual production of ore and
sub-ore = 2 x 10^ MT» and waste rock = 2.2 x 104 MT.
The above-ground sources also emit radionuclides to air as particu-
lates. 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 (EPA83b). An assessment of the health 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 (SPA83b) .
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 Tables 5-1 and 5-8. The reference mine has 5 vents in the
configuration as shown in Figure 5-4.
5-14
-------
Table 5-
Reference underground uranium mine
Parameter
Value
Ore grade
Ore production
Days of operation
Number of vents
Vent height(a)
Radon emissions
0.22 percent
112,000 tons/y
250 days/y
5
3 meters
11,000 Ci/y(b)
'a'In estimating radon-222 concentrations in Table 5-9 for releases
with plume rise, the following vent parameters were used: vent
diameter is 1.5 meters, exit velocity is 16.2 meters/sec, and the
exit temperature is 287°K (Drc84).
Ci/y from each vent.
Table 5-7. Summary of radon™222 emissions by age of underground
uranium mine (Ja80)
Mine
A
B
C
D
E
F
G
H
J
K
L
R
U
V
Y
Z
Average
Age
(years)
3
9
9
7
-
-
4
_
_
_
-
_
4
2
6
_
6
New mines
Radon-222 emissions
(Ci/y)
7,400
4,500
4,600
3,600
_
-
1,800
-
-
_
-
-
900
1,000
17,500
—
5,200
Age
(years)
-
_
-
-
21
20
_
21
20
19
29
20
-
_
-
17
21
Old mines
Radon-222 emissions
(Ci/y)
-
_
-
-
29,800
9,400
-
15,200
7,900
6,400
1,400
14,800
-
_
-
2,600
10,900
from measurements made in 1978 and 1979.
5-15
-------
Table 5-8. Estimated ore production of selected
mines, 1982 (Brb84)
Estimated 1982 production
Mine (103 tons/y)
New Mines
(Mines less than 10 years
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kanab North
La Sal
Heel a
Big Eagle
Golden Eagle
Mt. Taylor
Old Church Rock
Church Rock-East
Rerr-McGee
Section 19
Nose Rock
Mariano Lake
average
Old Mines
(10 years or more)
Sunday
Dermo-Snyder
Wilson-Silverbell
Lisbon
Sheep Mtn.
Church RockHNE
Church Rock- 1
Kerr-McGee
Section 30-East
Section 30-West
Section 35
Section 36
Home stake
Section 23
Section 25
Schwartzwalder
Average
old)
38.0
51.6
137.6
63.1
(a)
(a)
81.7
14.8
16.6
(a)
328.5
28.6
72.3
127.2
(a)
36.8
62
41.7
58.5
16.5
73.3
0
171.9
176.8
119.5
132.4
195.1
111.2
208.9
67.9
198 ,,8
112
operational.
5-16
-------
100 200 300 400 500
METERS
Figure 5-4, Reference underground mine,
5-n
-------
Page Intentionally Blank
-------
mine, the ground level concentrations resulting from emissions from the
reference mine were calculated for both ground level (all horizontal
vents, i.e., no plume rise) and an elevated release (all vertical vents
with plume rise). A ground level release with no plume rise represents
a worst case assumption in terms of the computed ground level radon-222
concentrations. A release with plume rise represents a lower bound
case for computed radon~222 concentrations. The radon concentrations
computed with these two assumptions will cover the range of concentra-
tions which can result from various local influences on plume rise.
Table 5-9 shows the estimated radon~222 concentrations and
resulting working levels and lifetime fatal cancer risks at various
distances from the shaft of the reference mine for releases with and
without plume rise. The most likely radon-222 concentrations at these
locations will fall somewhere within the range of values shown. These
concentrations were computed using EPA's Industrial Source Complex Long
Term Model (Drc84).
The estimated concentrations from ground level releases shown in
Table 5-9 for distances at 500 and 1000 meters from the mine shaft are
worst case situations with locations sited between a series of mine
vents or relatively close (within a few hundred meters) to one of the
vents where all of the vents involved are horizontal (i.e., no plume
rise). It is unlikely at the present time that such extremely high
concentrations actually exist near an underground uranium mine or that
any individual is actually exposed to these high levels. However, as
shown in Table 5-13, several hundred people are living within 1000
meters of underground uranium mine shafts, and these people are esti-
mated to be exposed to increased radon-222 concentrations somewhere
within the range of values shown for these locations in Table 5-9.
Table 5-10 shows estimated equilibrium ratios for radon at various
distances assuming a wind speed of 1 m/sec from the uranium mine.
Estimates of the radon-222 concentration at various distances from an
underground uranium mine with five vents emitting 11,000 Ci/y of
radon-222 are shown in Table 5-9 (Drc84). Also shown in this table are
the estimated lifetime risks of fatal cancer to nearby individuals from
the inhalation of radon~222 decay products produced (inside a house) by
radon-222 concentrations. Table 5-11 shows the relationship between
working levels and risk. The basic assumptions used in developing this
table are discussed in Chapter 8, Volume I. This relationship is not
linear because of competing risks of death from other causes. Using
the relationship between equilibrium ratio and radon concentrations,
the working level inside a structure at the specified distance is
calculated as shown in Table 5-9. Table 5-11 is then used to estimate
the lifetime risk for a person living in a structure 75 percent of the
time near these sites.
To evaluate the extent to which emissions from multiple mines
located close together will influence the radon-222 concentrations in
5-19
-------
Table 5~9« Estimates of working levels and risk of fatal
cancer in buildings at selected distances from
the reference underground uranium
Ground level release
Distance*-
(meters)
b) Radon-
222*-c
Lifetime
•j Working risk to
levels nearby
V poi/ L,J ( A\
individuals^ '
500
1,000
2,000
3,000
5,000
7,000
10,000
27.6
10.2
2.2
1.1
0.5
0.3
0.2
.113
,045
.011
.006
.003
.002
.001
1E-1
5E-2
1E-2
7E-3
3E-3
2E-3
IE-3
(5E-2)
(2E-2)
(5E-3)
(3E-3)
(2E-3)
(IE-3)
(5E-4)
Release with
Radon-
222^c
{ T-vf-t /I
VpUI/ L,
0.4
0.4
0.2
0.2
0.1
0.1
0.1
) Working
x levels
plume rise
Lifetime
risk to
nearby
individuals*-07
.0016
.0019
.001
.001
.0006
.0006
.0006
2E-3
2E-3
IE-3
IE-3
7E-4
7E-4
7E-4
(8E-4)
(9E-4)
(5E-4)
(5E-4)
(3E-4)
(3E-4)
(3E-4)
'a'The lifetime risks were estimated depending on the equilibrium ratios
calculated in the structures at various distances (See Table 5-10).
* 'The distance is measured from the shaft of the model mine. This is
different from the distances shown in Tables 5-8 and 5-9 of the Draft
Background Information Document (EPA83d) where the distances listed
were distances from, vent 5.
values in the first column are based on BEIR-3, NRPB, and EPA
models (see Chapter 8, Volume I). The values in parentheses are based
on UNSCEAR and ICRP risk estimates (see Chapter 8, Volume I).
Table 5-10, Outdoor and indoor equilibrium ratios for radon emitted
from an underground uranium mine at selected distances from the mine''37'
Distance
Time for plume
to reach distance (min)
Equilibrium ratio
(Outdoors)
(Indoors)
-------
Table 5-11. Relationship between working level
risk of fatal cancer
Working level Lifetime risk
.0001
,001
.01
.1
1B-4
1E--3
1E-2
1E-1
(5B-5)
(5B-4)
(5B-3)
(5B-2)
values in the first column are based on BEIR-3, MRPB, and
EPA models (see Chapter 8, Volume I). The values in parentheses
are based on UNSCEAR and ICRP risk estimates (see Chapter 8,
Volume I).
air, PNL carried out a modeling study using the Ambrosia Lake District
of New Mexico as a "case study" (DrbBl). 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 concentrations (above background) in air for this
region. Although these computed concentrations are only approximate
values, because of the complexities of this modeling study, the results
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, Another study of
multiple mines done by PNL (Drc84) confirms these conclusions. The PNL
also looked at the effect of plume rise on concentrations from multiple
mines due to vertical vents. If it Is assumed that all the vents in a
multiple mine area are vertical, (pluiae rise), the concentrations are
much lower than if the vents are assured to be horizontal (ground level
release).
Two measurement studies were also conducted in the Ambrosia Lake,
Mew Mexico, area to determine the concentrations of radon around
uranium mines and mills. The EPA conducted the first study in November
1975 (BPA75) at the request of the New Mexico Environmental Improvement
Agency and found that ambient outdoor radon concentrations were in
excess of typical background levels. It was suggested that a better
definition of background levels in the area be determined and a
thorough evaluation of specific source terms be conducted.
In 1978 the New Mexico Environmental Improvement Division
conducted a two-year program (Bu83) to determine (1) sources of high
concentrations of airborne radioactivity in uranium producing areas,
(2) radioactivity levels due to background as well as levels associated
with uranium milling and mining activities, and (3) if New Mexico
5-21
-------
> V ..
1^>
& A * ;Sv •
iwt
U\^> 1
-^~ , ^x~ \c
y/^? - ^*\. \j j
L. ^ .^.~**# ^ i™"T
•^-L f*w—'-^ y
7 V *
- >w>X^
^~v ^
(
f—^-f rt. >>
Stun
10km
Figure 5-5, Detailed map of mining area showing mine vent source.
5-22
-------
5
KILOMETERS
5
Figure 5-6. computed radon concentration map for region isopleths CpCi/L)
5-23
-------
standards are being exceeded. Background radon concentrations were
determined at six representative undisturbed locations within the
Grants Mineral Belt, Uranium mines were found to be the priioary cause
of elevated radon concentrations in Ambrosia Lake. Ambient radon
concentrations near uranium mines exceeded the New Mexico radiation
standard for an individual member (3pCi/l) of the public at three of
the nine locations in the study.
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 5-12, These estimates are for a site near Grants, New
Mexico, with a regional population of 36,000 using AIRDOS-EPA to
calculate the radon exposures (Appendix A). The number of fatal
cancers per year of operation of the reference mine is estimated to be
about 0.04 to the regional population and 0.08 to the national
population.
The inert radon gas emitted from mines can be transported beyond
the 50-mile regional cutoff. A trajectory dispersion model developed
by NOAA (Tr79) has been used to estimate the national impact of radon
emissions from the mine. This model calculates the potential radiation
exposure to the U.S. population for radon released from four typical
uranium processing locations. (Descriptions of these typical mill
sites--Casper, Wyoming; Falls City, Texas; Grants, New Mexico; and
Wellpinitj Washington—are given in (Tr79).) Only exposures taking
place beyond the 50-mile regional limit are considered. Details of the
model are given in He75. The model yields radon concentrations (pCi/L)
in air which were converted to decay product concentrations by assuming
that 100 pCi/L of radon corresponds to a decay product concentration of
0.7 WL.
Table 5-12. 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 Nat ipnal pppulat ion
Source (Person- (Fatal cancers/y (Person- (Fatal cancer s/y
WL-y) of operation}'3' WL-y) of operation)^3'
Underground
uranium mine 2.2 6E-2 (2E-2) 6.2 1E-1 (6E-2)
values in the first column are based on BEIR-3, NEFB, and EPA
models (see Chapter 8, Volume I). The values in parentheses are
based on UNSCEAR and ICRP risk estimates (see Chapter 8, Volume I).
5-24
-------
Uragium^Mi|iTing
An estimate of the total health impact from radon~222 emissions
from all underground uranium mining (using production values for 1982)
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 ratio is about 25. The estimate for the regional
population is about two fatal cancers/year and for the national
population is about three fatal cancers/year.
5.8 Reduction ofExposuresthrough Land Control
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, the Pacific 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 (Brb83). These mines represented about 84 percent
of the uraniura production from underground mines at that time.
Table 5~13 shows the population data gathered from the PNL 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.
Maps showing the distribution of population around these mines are
located at the end of this chapter.
Table 5-14 represents the percent distribution of land ownership
around the 30 surveyed mines. County tax assessors' 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. Land values for the private land were estimated from:
(1) assessed valuations and applying applicable selling price to
assessed valuation ratios, (2) estimates from local real estate agents,
(3) information supplied by state and county assessors, and (4) local
newspapers. The valuations were based on surface usage and rights
only, since the mineral values would remain intact.
Table 5-15 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 5-15 because it is
5-25
-------
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 tor mountain resort homes. The mine is also located in a
mountainous region so that radon emissions may 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 5-13 through 5-15 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:
Total cost Yearly cost
Type of cost (millions) (millions)
Land cost (100 percent contingency
with 10 percent yearly cost) $15.0 $1.5
Structures (100 percent contingency with
amortization over 5 years at 10 percent) 3.8 1.1
Relocation of 420 non-Indian residents
($5,000/person with amortization
over 5 years at 10 percent) 2.1 0.6
Relocation of Indian residences ($18,000/
person- 198 Indians, with amortization
over 5 years at 10 percent) 3.6 1.1
Total yearly cost 4.3
The 10 percent yearly cost assumes that the land value does not
change and thus is a nondepreciated asset. The present worth factor
Cor amortization over a 5 year period using a 10 percent interest rate
is 0.264. This is rounded to 0.3 to account for taxes.
Assuming that the 29 mines produced 84 percent of the underground
mine yearly production of 6,200 tons of 0303 for the industry
(Brb83)» the cost of land control per pound of 0303 can be
estimated as follows:
cost/lb U,0ft = $4,300,000 = $0.41/lb U308
(.84)(6,200X2,000)
If production costs for 0303 are $30/lb, the increased cost to
the industry would be 1 percent of the cost of production. In a
similar manner, it can be calculated that the cost per ton of ore would
be $1.82/ton of ore. This can then be compared to the cost of radon
control technologies in Table 5-3.
5-26
-------
Table 5-13. Population around selected underground
uranium mines (Brb84)
Mine
Sunday
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kanab North
Derrao-snyder
Wil son-
Si Iverbell
Lisbon
Lasal
Hecla
Big Eagle
Golden Eagle
Sheep Mtn.
Mt. Taylor
Old church
Rock
Church
Rock-MB
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
Ar izona
Colo. /Utah
Utah/Colo.
Utah
Utah
Utah
Wyoming
Wyoming
Wyoming
New Mexico
New Mexico
New Mexico
New Mexico
New Mexico
Mew Mexico
Mew 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
9
0
0
0
3
0
0
0
0
0-1
0
0
0
0
1
0
0
5
0
0
0
16
0
0
0
100
9
11
11
0
3
5
0
0
0
0-2
0
0
0
0
1
0
0
21
0
0
53
20
0
0
0
317
70
22
22
9
3
5
0
0
0
mine
0-3
0
0
0
0
1
0
0
49
12
4
101
40
0
6
0
336
139
26
27
57
3
5
4
0
0
(km)
0-4
0
0
0
0
1
0
0
67
20
44
194
73
0
6
0
336
187
31
31
70
3
5
4
0
0
0-5
0
0
0
0
1
0
0
83
23
44
194
73
0
6
12
336
364
31
31
131
3
6
4
0
0
5-27
-------
fable 5-13. Population around selected underground
uranium mines (Brb84) (Continued)
Mint} State
0-1/2
Horaestake
Sec 23 New Mexico 0
Homestake
Sec 25 New Mexico 0
Nose
Rock^9' New Mexico 0
Mar iano
Lake New Mexico 13
Schwartz-
walder^3^ Colorado 3
Totals 42
Distance from mine (km)
0-1 0-2 0-3 0-4 0-5
00334
00000
0 0 0 26 35
44 75 196 274 352
3 63 102 136 147
205 618 1,009 1,375 1,733
population around this mine is not included in the total
because the location is not typical of the industry.
5-28
-------
Table 5-14. Percent distribution of land ownership around
selected underground uranium mines (Brb84)
Mine
Sunday
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kanab North
Dermo-Snyder
Wi Ison-
silverbell
Lisbon
LaSal
Hecla
Big Eagle
Golden Eagle
Sheep Mtn.
Mt . Taylor
Old Church
Rock
Church Rock
NE
Church Rock
tl
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
Distance from mine (km)
0-1/2
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
0/0/10Q
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
58/16/29
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
45/13/42
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
39/10/51
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
39/7/54
3/0/97
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.
5-29
-------
Table 5-14. Percent distribution of land ownership around
selected underground uranium mines (Brb84) (Continued)
Distance from mine (km
(a)
nine
Homestake
Sec 23
Homestake
Sec 25
Nose Rock
Mariano Lake
Schwartz-
Average
0-1/2
74/0/26
100/0/0
0/50/50
0/0/100
100/0/0
20/22/58
0-1
68/0/32
85/0/15
0/50/50
0/0/100
100/0/0
22/20/58
0-2
61/6/33
59/0/41
0/45/55
0/0/100
100/0/0
22/17/61
0-3
50/18/32
58/1/41
0/41/59
0/0/100
100/0/0
23/13/64
0-4
47/11/36
50/2/48
0/38/62
0/0/100
100/0/0
22/12/66
0-5
53/12/35
43/10/47
0/35/65
0/0/100
100/0/0
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.
5-30
-------
fable 5-15. Estimated value of private land around
selected underground uranium mines^ (BrbS4)
(In thousands)
Mine
Sunday
King Solomon
Velvet
Tony M
Hack Canyon
Pidgeon
Kanab North
Dermo-Snyder
Wilson-
silverbell
Lisbon
Lasal
Heel a
Big Eagle
Golden Eagle
Sheep Mtn.
Mt, faylor
Old Church Rock
church Rock ME
Church Rock-1
Church Rock- las
Kerr-McGee
Sec 30 East
Kerr-McGee
Sec 30 Vest
Kerr-McGee
Sec 19
Kerr-McGee
Sec 35
Kerr-McGee
Sec 36
0-1/2
NA
MR
5.5
NR
NA
Nft
NA
79.7
39.1
NR
4.0
36.8
NA
35.4
18.0
39.6
Nft
NR
NA
t NA
35.0
31.1
Nft
Nft
3.4
0-1
NA
NA
16.0
NA
NA
Nft
NA
260.4
186.4
Nft
228.4
147.3
NA
209.0
42.3
391.5
Nil
NA
MA
MA
35.0
132,2
194,4
NA
23.5
Distance 1
0-2
NA
NA
36.0
NA
NA
NA
Nfi
922,6
535,8
50.0
920,9
380.0
NA
796.2
42.3
2,523.7
MA
MA
MA
MR
35.0
147.8
844.8
37,0
124.3
from mine
0-3
48.0
Hft
172,8
NA
Nft
MA
MR
1,852.1
1,667.2
306.0
1 , 427 . 8
691.0
m.
2,121.0
42.3
2,834.2
NA
HA
NA
122.2
53.5
157.9
1,229.4
137.8
336.0
(kml
0-4
208.0
MA
603.2
MA
NA
Nft
NA
3,028.9
2,861.6
810.5
2,484.5
965.9
NA
3,584.0
150.0
3,227.4
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
WA
(b)
4,432,8
3,968.7
810.5
2,534.5
1,000.5
NA
5,231.0
898.0
3,918.8
1,443.1
NA
NR
355.6
240.0
235.1
1,532,8
336.0
977.8
See footnotes at end of table.
5-31
-------
Table 5-15. Estimated value of private land around
selected underground uranium mines^a' (Brb84) (Continued)
(In thousands)
Mine
Homes take
Sec 23
Home stake
Sec 25
Nose 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
3,016.6
Distance
0-2
994.1
987.8
MA
NA
15,200.0
9,378.2
from mine
0-3
1,158.7
1,478.0
MA
NA
33,600.0
15,835.8
(km)
0-4
1,485.2
1,632.2
NA
NA
58,400.0
24,443.8
0-5
2,361.8
1,645.6
NA
NA
89,200.0
33,354.6
cost of land (80 percent) and structures (20 percent).
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.
5-32
-------
112°S6'
K ,„,_,%(,'"•.-
i> ""C, / ; v>
Jk-
UJ
ft
:•• KnbHs
• > ^ ' ' " ' / •>
s >-^? '-/-ft]
' ; -IA£* - -:/ f
POPUtATIOSI DiSTSIlBlfnON
0-1/a km 0-1 km 0-2 km 0-3 Ism 0-4 tow
HACK
. CANYOM
ALL
LOCATION; POPULATION DISTRIBUTION I o
i 1
123
AMI J.l
I MILES
KILOMETERS
LEGEND:
@ SW@Li OCCUPIED
@ OCCUPIED OWEIUNG
(HO, IS STRUCTURES!
3 MiWS SHAFT
-------
112°3S'
112°30'
POPULATION DISTFU8UTIQN
0-J/2Jm fiM_Nm Q-2Jcm O-SJcm ^jjcro O;Sjgg
KANAB
0 09000
PIGEON 0 00000
01234
1
-j = ^ MtlKS
ag^^__aaa=JaJ»t^4^J? KILOMETERS
@ SIW6LE OCCUPIED I
(2) OCCUPIED DWELL1NS
IS STRUCTURES!
B MINE SHAFT
-------
0-1/2km 0-1 SCOT 0*2 km 0-3 km 0-4 km 9-SScm
. MILES
KILOMETERS
•
(2) OCCUPIED
{SO. ""
a
-------
ALL OCCUPIED
LOCATION; POPULATION
POPULATION DISTRIBUTION
SUNDAY
MILES
JmummA^mJtii^mjg KILOMETERS
« SINGLE OCCUPIED
0 OCCUPIED f
(NO, IS STRUCTURES)
B mim SHAFT
-------
107"S5'
107°50'
107045'
107°40'
*-J
HOWESTAKE SEC Z3OT
x f~^ f \ \ ^
J -x./ I-' _V _ L ^N _».,-^
. (J J ^
POPt'LATION DiSlRIBUTiON
0-1/2km 0-1 km 0-2 fern 0-3 km 0-4 km 0-5 km
LEGEND:
SINGLE OCCUPIED DWELLING
(7) OCCUPIED DWELLING CLUSTER
KEBS-McGEE SEC. 306 3 3 3 333
KERR'-MoGEE SEC. SOW 0 S 6 5 5
KERR-McQEE SgC. 19 0 0 0 4 4
KEBR.-MOGEE SEC. 36 o o o o o o
.38 0 0 0 0 0 0
MOMESTfMCE SEC. 23 0 0 0 334
HOMcSTAKE SEC. 25000 000
{NO. IS STRUCTURES)
B MiNE SHAFT
-------
OLD CHURCN RUf t
CHURCH ROCK HF
CHURCH ROCK K
CHURCH POL US-/*-!
0 (
eTT"
-------
POPULATION DISTRIBUTION
MINE_ _______ 0-1 /2km OilAl1 Qi3-t03 tJli™
44
13
7S 198 274 352
MILES
LEGEND;
• SINGLE OCCUPIED DWELLING
@ OCCUPIED DWELLING CLUSTER
(NO, IS STRUCTURES}
H SHAFT
5-39
-------
107035'
MT.
CIBOLA
ALL OCCUPIED DWELLINGS; DISTRIBUTION
POPULATION DISTRIBUTION
MINE
05jjm
MT.TAYLQR 0 100 317 330 33i
LEGEND:
» SINGLE OCCUPIED DWELLING
(z) OCCUPIED DWELLING CLUSTER
1 2
: MiLES
{NO. IS STRUCTURES)
3 MINE SHAFT
«JI
5-40
-------
MIME, McKINLEY COUNTY, NEW MEXICO
ALL SHAFT LOCATION; POPULATION DISTRIBUTION
POPULATION DISTRIBUTION
MINE 0-1/2km 0-1 km 0-2km Q-Sfwn 0-4km 0-5km
NOSE
ROCK
26
35
« MitES
KILOMETERS
LEGEND:
© SINGLE OCCUPIED DWELLING
@ OCCUPIED DWEtLISMG CLUSTER
(NO. IS STRUCTURES)
H MINE SHAFT
5-41
-------
SAN JUAN
ALL OCCUPIED SHAFT LOCATIONS; POPULATION DISTRIBUTION
MINE
DiRMO-
SNYDEK
WILSON-
SILVERBELL
0 1
POPULATION DISTRIBUTION
(Mkm CK2ton O^Uffi Oj4km 0;6km
S 21 4S 67 83
20 23
i MILES
KflQMITERS
LEGEND:
• SINGtE OCCUPIED DWELLING
(z) OCCUPiED DWELLING CLUSTER
{NO. IS STRUCTURES!
QS MINE SHAFT
5-42
-------
, LjSiQN/taSAL/HECLA
SAN JUAN COUNTY, UTAH
ALL OCCUPIED DWELLINGS; MINE SHAFT
LOCATIONS; POPULATION DISTRIBUTION
POPUlATt
0-1/2fcHi 0-1 tun 0-2ton
U8BON !' 0! S"0
DISTRIBUTION
04fcm 0-SJan
LaSAi
KECLA
0 4 44 44
53 101 194 194
lie! | JTOi .40—73 73
2 3 4 .
6 1 2 a m
iMilES
KILOMETERS
LEGEND:
® SINQLE OCCUPIED DWELLING
(J) OCCUPIED DWELLING CLUSTER
(NO. )S STRUCTURES)
B MINE SHAFT
-------
106°60'
I
r06°46'
tOB°«P'
, :-:-r....
106°36'
TONY M MINE
I3ARFIELD COUNTY, UTAH
ALL OCCUPIED DWELLINGS; MINE SHAFT
LOCATIONS; POPULATION DISTRIBUTION
POPULATION DISTRIBUTION
0-1/2km 0-lfcm 0-2km 0-3km Qj^ftei 0-Bkm
TONYM 0 0 0000
MINE
.MILES
a i * KILOMETERS
LEGEND:
9 SiNGLE OCCUPIED DWSLLiNS.
0 OCCUPIED DWELLING CLUSTER
CNO, IS STROCTURISI
U MINE SHAFT
-------
(SI
\
a OCCUWED i
ALL OCCUPIED DWELLINGS; MINE
LOCATION; POPULATION DISTRIBUTION
-------
BIG EAGLE/SHEEP MOUNTAIN
ALL OCCUPIED DWELLINGS; MINE
MINE
BIG EAGLE
SHEEP
MOUNTAIN
POPULATION DISTRIBUTION
0-172km
-------
1O6"46'
•=06° 36
Ut
BILL SMITH/GOLDEN EAGLE
CONVERSE COUNTY, WYOfyilNG
ALL OCCUPIED DWELLINGS; MINE SHAFT
LOCATIONS; POPULATION DISTRIBUTJOW
MINE
BILL SMITH
GOI.DEW
EAGLE
0 1
POPULATION DISTRIBUTION
0-1/2km Chljcm 0-2 km 0-3 km O£fcm 0-5 km
0
0
0
0
0
0
8
fiSILiS
&
s
8
I KILOMETERS
LEGEND.
® SINGLE OCCUPIED DWiUJNS
© OCCUPIED DWELLING CLUSTER
(NO. IS STRUCTURES!
Q| MINE SHAFT
-------
REFERENCES
B184 Bloomster C. H., Jackson P. 0., Dirks J. A., and Reis J. W.,
Radon Emissions from Underground Uranium Mines, Draft Report,
Pacific Northwest Laboratory, 1984.
BraSl Brown S. H. and Smith R. C., A Model for Determining the
Overall Radon Release Rate and Annual Source Terra for a
Commercial 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.
Brb84 Bruno G. A., Dirks J. A., Jackson P. 0., and Young J. K.,
U.S. Uranium Mining Industry: Background Information on
Economics and Emissions, Pacific Northwest Laboratory,
PNL-5035 (UC-2, 11, 51) March 1984.
Bu83 Buhl T., Millard J., Baggett D., Brough T., and Trevathan S.,
Radon and Radon Progeny Concentration in New Mexico's Uranium
Mining and Milling District, New Mexico Health and
Environment Department, 1983.
DOE83 Department of Energy, Statistical Data of the Uranium
Industry, GJQ-1QO(83), Grand Junction, Colorado, January 1983.
Dra80 Droppo J. G., Jackson P. 0., Nickola P. W., Perkins R. W.,
Sehmel G. A., Thomas C. W., Thomas V. W., and Wogman N. A.,
An Environmental Study of Active and Inactive Uranium Mines
and Their Effluents, Part I, Task 3, SPA Contract Report
80-2, EPA, Office of Radiation Programs, Washington, B.C.,
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,
Drc84 Droppo J. G., Modeled Atmospheric Radon Concentrations from
Uranium Mines, Draft Report, Pacific Northwest Laboratory,
PNL-52-39, September 1984.
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.
5-48
-------
REFERENCES—continued
EPA83a Environmental Protection Agency, Eegulatory Impact Analysis
of Final Environmental Standards for Uranium Mill Tailings at
Active Sites, EPA 520/1-83-010, EPA, Office of Radiation
Programs, Washington, B.C., September 1983.
EPA83b Environmental Protection Agency, Potential Health and
Environmental Hazards of Uranium Mines Wastes, EPA
520/1-83-007, EPA, Office of Radiation Programs, Washington,
B.C., June 1983.
EPA83c Environmental Protection Agency, Final Environmental Impact
Statement for Standards for the Control of Byproduct
Materials from Uranium Ore Processing (40 CFR 192), Volume II,
page A.2-33, EPA 520/1-83-008-2, EPA, Office of Radiation
Programs, Washington, B.C., September 1983.
EPA83d Environmental Protection Agency, Background Information
Document—Proposed Standards for Radionuclide Draft Report,
EPA 520/1-83-001, EPA Office of Radiation Programs,
Washington, D.C., March 1983.
FraSl 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.
FrbSl Franklin J. C. and Weverstad K. B., Radiation Hazards in
Backfilling with Classified Uranium Mill Tailings,
Proceedings of the Fifth Annual Uranium Seminar, Albuquerque,
New Mexico, September 20-23, 1981.
Frc83 Written communication between J« C. Franklin of the Bureau of
Mines and W. J. Shelley of the Kerr-McGee Corporation, May
1983.
He75 Heffter J. L., Taylor A. D., and Ferber G. J., A Regional-
Continental Scale Transport, Diffusion, and Deposition Model,
NOAA Tech. Memo, ERL/ARL-50, 1975.
Ho84 Hopke P. K., Leong K, H., and Stukel J. J., Mechanisms for
the Removal of Radon from Waste Gas Streams, EPA Cooperative
Agreement CR 806819, UILU-ENG 84-0106, Advanced Environmental
Control Technology Research Center, 3230 Newmsrk Civil
Engineering Laboratory, 208 Horth Romine Street, Urbana,
Illinois 61801, March 1984,
5-4
-------
REFERENCES—continued
Ja80 Jackson P. O.s Glissmeyer J. A., Enderlin W. I., Sehwendimam
L. C., Wogman N. A., and Perkins R. W., An Investigation of
Radon-222 Emissions from Underground Uranium Mines—Progress
Report 2, Pacific Northwest Laboratory, Richland., Washington,
February 1980.
Ko80 Kown B. T., VanderMast V. C., and Ludwig K. L., Technical
Assessment of Radon-222 Control Technology for Underground
Uranium Mines, ORP/TAD-80-7, Contract No. 68-02-2616, EPA,
Office of Radiation Programs, Washington, D.C., April 1980.
N8C80 Nuclear Regulatory Commission, Final Generic Environmental
Impact Statement on Uranium Milling, NUREG-Q7Q6, Office of
Nuclear Material Safety and Safeguards, NRC, Washington,
D.C., September 1980.
Th81 Thomas W. V., Musulin C. S., and Franklin J. C., Bulkheading
Effects on Radon Release from the Twilight Uranium Mine,
PNL-3693 (UC-11). Prepared for EPA under a Related Services
Agreement with DOE, Pacific Northwest Laboratory, Richland,
Washington, June 1981.
Tr79 Travis C. C., Watson A. P., Me Dowel1-Boyer L, M., Cotter
S. J., Randolph M. L., and Fields D. E., A Radiological
Assessment of Radon-222 Released from Uranium Mills and Other
Natural and Technologically Enhanced Sources, QRNL/NUREG-55,
ORNL, Oak Ridge, Tennessee, 1979,
5-50
-------
Chapter 6: PHOSPHATE INDUSTRY FACILITIES
6. 1
6.1.1 General Descript ion
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 (DM68). Phosphate rock mines of significant commercial
importance are located in Florida, North Carolina, Tennessee, Idaho,
Wyoming, Utah3 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 „ -
,,. , , , Percent of
„ , , (Metric tons) . . ,
Company and location CinS'i total
International Minerals and Chemicals 11,340 20,5
Bonnie, Florida
Kings ford, 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
Brewster Phosphate 3,175 5.7
Brews ter, Florida
Bradley, Florida
U.S. Steel-Agri-Chera, 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 0.8
Polk City, Florida
6.1-3
-------
6.1-1. Phosphate rock producers and capabilities (EPA79)
(Continued)
Company and location
Beker Industries
Dry Valley, Idaho
J. R. Simplot
Ft. Hall, Idaho
Coininco-- American
Garrison s Montana
George Relyea
Garrison, Montana
Texasgulf
Auroras 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
SurajoarybyRegion
Florida
North Carolina
Tennessee
Western States
78.3
7.8
4.1
9.8
6.1-4
-------
North Carolina possesses a rich phosphate rock deposit in Beaufort
County along the Patalico 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(PO/j.)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 typical uranium
concentration in rocks and soils (2 ppm). The radionuclides of
significance which are present in phosphate rock are: uranium-238,
uraniunr-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
After phosphate rock ha.s 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 organies, it usually is not calcined. Most
calcining is done in fluidized-bed units, but rotary calciners are also
used. The rock is heated to 14GG°-16QO° 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 ControlTechnology(TRW82)
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 finess
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.
Prjer and J3a_lciner 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. Because 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 arid
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
6.1-4
-------
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 Ra dio nuc
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
Total particulates (g/y)
Operating time (hr/y)
Stack emissions (Ci/y)
Uranium-234
Uranium-235
Uranium-238
fhorium-227
Thorium-228
Thoriuni-230
Thoritmr-232
Radium-226
Dryer 1
2.2E+7
4114
7.0E-4
3.0E-5
6. 6E ~4
S.QE-5
1.4E-4
9.7E-5
3.0E-5
9.3E-4
Dryers 3 and 4
5.0E+7
4338
2.6E-3
2»4E-4
2.7E-3
2.0E-4
2.3E-4
2.5E-3
8.0E-5
2.9E-3
More recently, in 1983 and 1984, EPA measured the radionuclide
emissions from phosphate-rock calciners. Because calciners operate at
a higher temperature than dryers, they have the potential for
volatilizing lead-210 and polonium-210. Information on the
measurements made at calciners at elemental phosphorus plants is
presented in Section 6*3, (Note: phosphate rock processing at
6.1-9
-------
elemental phosphorus plants has been analyzed separately from other
phosphate rock processing facilities,} An analysis of the results of
measurements at calciners at wet process phosphoric acid plants has not
yet been completed and the following sections do not include an
assessment of the health impact of radlonuclide emissions from these
calciners.
6.1.5 Reference_Pl_an_t
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 grama of partieulates 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)
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)
3
2.7E+6
6570
40
20
2
10
60°
Low energy
scrubber
130 (0.26)(c)
4
1.2E+6
6460
40
20
2
10
60°
Medium energy
scrubber
25
units process 145 MT/hr; grinder units process 45 MT/hr.
Uranium-238 is assumed to be in equilibrium with its daughter
products.
in Ib/ton.
6.1-10
-------
Table 6.1-4. Radlonuclide emissions from the reference
phosphate rock drying and grinding plant
n ,. ... Emissions (Ci/y)
Had lonuc 1 ide — - •^~
Dryers Grinders
Uranium-238
Uranium-234
Thoriuni-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 nearby individuals
are located 750 meters from the plant.
Table 6.1-6 presents estimates of the lifetime risk to nearby
individuals and the number of fatal cancers per year of operation from
these doses.
The lifetime risk to nearby individuals is estimated to be about
1E-5 and the number of fatal cancers per year of operation is estimated
to be 1E-3. These risks result primarily from doses to the lung from
inhalation of radioactive particulates released from drying operations.
6.1.7 Existing EmissionStandards and Air Pollution Controls
No Federal or State regulations currently exist that limit
radionuclide emissions from phosphate drying, calcining, and grinding
operations. Particulate emissions from these sources are limited by
New Source Performance Standards (NSPS) which apply to facilities
constructed after September 1979, or State Implementation Plans (SIPs)
which cover sources operating prior to September 1979.
NSPS limits for phosphate rock processing are 30 g/MT for dryers,
115 g/MT for calciners handling unbeneficiated rock or a blend of
beneficiated and unbeneficiated rock, 55 g/MT for calciners handling
beneficiated rock, and 6 g/MT for grinders.
SIP limits for phosphate rock operations are less stringent than
NSPS limits. Florida, where approximately 80 percent of the industry
6.1-11
-------
is located, has established the most stringent SIP requirements,
limiting emissions from 30, 100} and 500 Cons/hour processing sources
to 30s 36 s and 47 Ib/hour, respectively. SIP limits in the other six
States where commercial facilities are located are 40, 51, and 79
ib/hour for processing rates of 30, 100, and 500 tons/hour-
6.1.8 Alter native^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.
Table 6.1-5. Annual radiation dose from radioactive particulate
emissions from the reference phosphate rock drying and grinding plant
Nearby individuals Regional population
Organ / / x , / \
vmrem/yj Iperson-rem/yJ
Lung
Endosteum
Red marrow
Kidney
7.2
1.5EH
1.3
1.0
6.0E+1
1.1E+2
9.2
6.8
Table 6.1-6. Fatal cancer risks due to radioactive emissions
from the reference phosphate rock drying and grinding plant
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation),
Dryers 1E-5 1E-3
Grinders IE-6 IE-4
Total 1E-5 1E-3
6.1-12
-------
Table 6.1-7. Annualized cost and risk reductions of alternative
controls for the reference phosphate rock drying and grinding pla
Process _ ^ ,
Control
option^)
JDryers(k) Existing
B-l
B-2
A-l
A-2
Grinders Existing
A-l
A-2
Mission
rate
(g/MT)
130
50
50
3Q(d)
30
25
6
Total
annua 1
cost (°
($1,000)
861
1770
1000
2320
124
4
Fatal cancer risks
Risk to
' nearby
individuals
1E-5
4E-6
4E-6
2E-6
2E-6
1E-6
2E-7
2E-7
Population
(cancer s/y of
operation)
1E~3
4E-4
4E-4
2E-4
2E-4
IE -4
2E-5
2E-5
Cost/fatal
cancer
avoided
(in millions)
1440
2950
1250
2900
1550
50
dryers:
For grinders:
B-l = venturi scrubber (15" W.G.)
B-2 = ESP
A-l = venturi scrubber (25" W.G.)
A-2 = high energy ESP
A-l = venturi scrubber (16" W.G.)
A-2 = 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.
6.1.9 Total Health Impact ofPhosphate Rock Processing Plants
Phosphate rock processing plants (dryers and grinders) release
about 3700 MI of particulate matter per year with the existing level of
control (TRW82). This particulate matter contains about 150 mCi of
uranium-238 and each of its daughter products. These emissions are
estimated to cause about 1E-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
per year from all plants
3700 MT PM/yr
380 MT PM/yr
X 0.0013 HE/yr (reference
facility)
0.013
6.1-13
-------
6.1.10
The industry incremental annualized costs to retrofit existing
phosphate dryer 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 0.003
fatal cancers/year, or a cost of $1830 million per fatal cancer
avoided. Retrofitting to the NSPS level (Control Option A) would cost
an additional $12 million per year and avoid 0.008 fatal cancers per
years or a cost of $1530 million 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 0.0008
fatal cancers per year, or a cost of $430 million per fatal cancer
avoided.
Table 6.1-8. Industry annualized costs and risk reductions for
retrofitting existing phosphate rock dryers and grinders^3'
Control,, ,. Total cost
Process unit . (b) , .,,. %
option (millions)
(c)
Grinders
Fatal cancers
avoided/j
0.34
8E-4
Cost/fatal
cancer avoided
(in millions)
Dryers
B
A
5.5
12.2
3E-3
8E-3
1830
1530
430
(b)jror dryers Option B is a venturi scrubber (15" W.G. )
and Option A is a venturi scrubber (25" W.G.I. For grinders, Option
A is a fabric filter.
^'Incremental cost for installing and operating alternative control
system (i.e., costs above existing costs).
-------
Page Intentionally Blank
-------
REFERENCES
DM68 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 TRW, Particulate Emissions and Control Costs of Radionuclide
Sources in Phosphate Rock Processing Plants. A report
prepared by Stacy G, Smith (TRW Energy and Environmental
Division, Research Triangle Park, N.C.) for Office of
Radiation Programs, December 1982.
6.1-15
-------
6.2 We_t_Proce«4 * NH3 - NH4H2P04 MAP (3)
2NH3 = (NH4)2HP04 DAP (4)
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 Che wet process (EPA79)
6.2-2
-------
6.2.3 .Coroj.^chnolo^ ..... (,TRW82)
Production processes for dxamrnGnium 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 DA?) and gaseous ammonia emissions (DAP
only) are largely unaffected by electrostatic precipitators or
baghcmses. 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 arid 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 Radj.onujy.ji.de J^isj3_ion_ Measuremeiits
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 refereo.ce 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*
-------
Table 6.2-1, Radionuclide stack emissions at wet process
phosphate fertilizer plants (EPA78)
Parameter
TSP dryer
Plant A
TSP dryer
Plant B
DAP dryer
Plant B
Total participates (g/y) 2.0E+7 1.2E+7 1.5E+7
Operating time (hr/y) 4.6E+3 7.4E+3 7.5E+3
Stack emissions (Ci/y)
Uranium-234
Uratiium-235
Uranium-238
Thorium-227
Thoriurn-228
Thorium-230
Thorium-232
Radium-226
Polonium-210
1.1E-4
ND
9.0E-5
ND
4.0E-5
9.0E-5
ND
3.0E-5
6.3E-4
3.QE-4
2.0E-5
2.7E-4
ND
3.0E-5
2.5E-4
7.0E-5
2.2E-4
NA
2.6E-3
1.9E-4
3.3E-3
ND
8.0E-5
3.0E-3
5.0E-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 nearby individuals are
located 1500 meters south of the reference plant.
Table 6.2-5 presents estimates of the lifetime risk to nearby
individuals and the number of fatal cancers per year of operation from
these doses.
The lifetime risk to nearby individuals is estimated to be about
2E-6 and the number of fatal cancers per year of operation is estimated
to be 6E-4. These risks result primarily from doses to the lung from
inhalation of radioactive particulates released from fertilizer
production.
6.2.7 Existing Emission Standards and Air Pollution Controls
No Federal or State regulations currently exist that limit
radionuclide emissions from wet process phosphate fertilizer plants.
Particulate emissions from these facilities are limited to the
quantities established by the States in their State Implementation
Plans (SIPs) for meeting Ambient Air Quality Standards.
6.2-5
-------
Florida,, where almost 80 percent of the industry is located;, has
the most stringent SIP limits. Phosphate processing operations are
limited to 0.3 Ib/ton of product (150 g/MT of product). The other
States with wet process phosphate fertilizer plants have riot
established specific emission limits for phosphate processing, but
restrict emissions to the levels established in their SIPs for general
processing sources. For sources greater than 30 tons/hour, allowable
emissions are determined by the formula:
where
E = (55.0 x P0.11)-4o,
E = emissions, and
P = the processing rate in tons/hour.
6.2.8 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.
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-238s 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
(a'Data from EPA78. DAP Diammonium phosphate.
GTSP Granular triple superphosphate.
6.2-6
-------
Table 6.2-3. Radionuclide emissions from the
reference wet process phosphate fertilizer plant
,., Hnissions (Ci/y)
Radionuclxde J—~
DAP GTSP Total
Uranium-238
Uranium-234
Thorium-230
Radium-226
Lead-210
Poloniuin-210
5. IE -3
5. IE -3
5.1E-3
4.3E-4
2.6E-3
2.6E-3
1.6E-3
1.6E-3
1.6E-3
5.4E-4
8.1E-4
8. IE -4
6.7E-3
6.7E-3
6.7E-3
9.7E-4
3.4E-3
3.4E-3
DAP Diananonium phosphate.
GTSP Granular triple superphosphate.
Table 6.2-4. Radiation dose rates from radionuclide emissions
from the reference wet process phosphate fertilizer plant
Nearby individuals Regional population
(mrem/y) (Person-rem/y)
Lung
Endosteum
Red marrow
Kidney
1.2
2.2
1.5E-1
6.7E-2
2.4E+1
4.1E+1
2.8
1.3
Table 6.2-5. Fatal cancer risks due to radioactive emissions
from reference wet process phosphate fertilizer plant
Lifetime risk Regional population
Source to nearby individuals fatal cancers/y of operation
DAP and GTSP
process emissions 2E-6 6E-4
DAP Diamttionium phosphate.
GTSP Granular triple superphosphate.
6.2-7
-------
6.2.9
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
0.01 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.
6.2.10 Co s t s and Risk Re due t ion s f or Retrofitti ng Ex i s t i ng 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
scrubbers (28 percent of the existing production capacity) would cost
an additional $3 million per year and would avoid 0.001 fatal cancers
per year, or a cost of $3 billion per fatal cancer avoided.
Retrofitting GSTP process exhausts with packed bed scrubbers (19
percent of existing production capacity) would cost an additional
1500,000 per year and would avoid 0.0004 fatal cancers per year, or a
cost of $13 billion per fatal cancer avoided.
fable 6.2-6. Annualized costs and risk reductions of alternative controls
for the reference wet process phosphate fertilizer
Process _ . ,
Control
DAP
GTSP
optionvb)
Existing
Alternative
Existing
Alternative
Total
Fatal cancer risks
Emission annual Individual
rate(c) cost lifetime
(g/MT) ($1,000)
164
100 500
100
79 300
risk
2E-6
1E-6
5E~7
4E-7
Population
(cancers/y of
operation)
5E-4
3E-4
IE -4
8E-5
Cost/fatal
cancer
avoided
(in millions)
2.5E+3
1.5E+4
DAP Diammonium phosphate GTSP Granular triple superphosphate.
Source: TRW82.
isting controls are venturi scrubbers. Alternative controls are
packed bed scrubbers in series with venturi scrubbers.
rt iculate material emission rate.
incremental cost for installing and operating alternative control
systems, i.e., additional costs for installing and operating
packed bed scrubbers.
6.2-6
-------
Table 6.2-7. Industry annualized costs and risk reductions for
adding secondary scrubbers to existing wet process phosphate
fertilizer plants^a)
„ , _, 1 Cost/fatal
Total cost/, % Fatal cancers . , „
Process ,...,, ^(.o) •_,,/ cancer avoided
(.millions) avoided/y /, .,,. ^
(in millions;
DAP
GTSP
3
0.5
1E-3
4E-5
1.3E+4
^^Incremental cost of installing and operating packed bed scrubbers
in series with existing venturi scrubbers. IWenty-eight percent of
DAP production capacity and 19 percent of GTSP production capacity
require retrofit.
DAP Diammonium phosphate.
GfSP Granular triple superphosphate.
6.2-9
-------
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,
D.C., 1979.
TRW82 TRW, Industry and Particulate Matter Control Technology
Information for Dianunonium Phosphate and Granular triple
Superphosphate manufacture. A report prepared by TRW
Environmental Division for the Environmental Protection
Agency, Dec 15, 1982.
6.2-10
-------
6 . 3 El -it Psptwrs 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
1983 approximately 366 thousand tons of elemental phosphorus were
produced .
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 typical rocks and soil (2 ppm). The
radionuclides of significance which are present in phosphate rock are:
uranium-238, uraniunr-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
furnaces can volatilize lead-210 and polonium-210, resulting in the
release of significant quantities of these radionuclides into the air.
There are 6 elemental phosphorus plants in the United States —
located in 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 De s cr i p t ion
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 + 6CaSi03 (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-gases pass through dust collectors and then through
water spray condensers. Pnosphorus is cooled to the molten state in
the condensors. 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 condensors 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
Location
Company
(tons/y of phosphorus)
Idaho
Pocatello
Soda Springs
Montana
Silver Bow
FMC Corporation
Monsanto Chemical Co.
Stauffer Chemical Co.
1.3E+5
9.QE+4
4.QE+4
Tennessee
Columbia
Columbia
Mt, Pleasant
Occidental Chemical Co. 5.7E+4
Monsanto Chemical Co. 7.5E+4
Stauffer Chemical Co. 5.GE+4
^Estimated capacity in 1984 (EPA84d).
6*3.3 ControlTechnology
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
/PHOSPHATE\_
ROCK V~
CALCINER
STACK VENT EXHAUST
CALCINED
NODULES
ELECTRIC
FURNACE
PRECIPITATOR
\
CABBON \
MONOXIDE V
Recycled
CONDENSERS
ELEMENTAL
PHOSPHORUS SALES
CARBON MONOXIDE
FLARE STACK
Figure 6.3-1. Flow diagram of the thermal process for
production of elemental phosphorus.
6.3-3
-------
6.3,4 Radionuclide Emission Measurements
In the period 1975-1980, EPA measured the radionuclide emission
rates from three elemental phosphorus plants. These plants were:
FMC in Poeatello, Idaho (EPA77), Stauffer in Silver Bow, Montana
(AnSla), and Monsanto in Columbia, Tennessee (AnSlb). These tests
included measurements from release points representative of all of the
major process operations in the production of elemental phosphorus.
Measurements were made of the emission rates from: calciners, calciner
coolers, material handling and transfer operations, screening plants,
furnace preparation areas, and furnace tap holes. 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. 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. The high
temperature of the calciners volatilizes 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. Analyses of doses and risks from these emissions show
the emissions of polonium-210 and, to a lesser degree, emissions of
lead-210 to be the major contributors to risk from radionuclide
emissions from elemental phosphorus plants (see Section 6.3.5).
In late 1983, EPA conducted extensive additional radionuclide
emission testing at the FMC plant in Pocatello, Idaho (EPA84a, RC84a),
and the Stauffer plant in Silver Bow, Montana (EPA84b, RC84b). Also in
early 1984, EPA conducted some limited emission testing at the Monsanto
plant in Soda Springs, Idaho (EPA84c, RC84c). This testing was limited
to calciner off-gas streams (based on results of previous emission
testing) and focused primarily on lead-210 and polonium~210 emissions,
The principal objectives of these tests were: (1) to obtain additional
information on the lead-210 and polonium-210 emissions in calciner
off-gas streams, (2) to determine the distribution of lead-210 and
polooium-210 by particle size in calciner off-gas streams, and (3) to
obtain a suitable sample for determining the lung-clearance classifi~
cation of lead-210 and polonium-210 in particulates collected from the
calciner off-gas streams.
Reports on this testing have been prepared for each plant as cited
in the above noted references. These reports contain the following
data and information: (l) radionuclide concentrations in the calciner
feed material and the calcined product (nodules), (2) radionuclide and
particulate concentrations and emission rates in calciner off-gas
streams including both inlet and outlet streams of emission control
devices, (3) particle size distribution of both radionuclides and
particulates in calciner off-gas streams including the distribution
for both inlet and outlet streams of emission control devices,
(4) estimates of the annual emission rates for both radionuclides and
>.3-4
-------
Table 6.3-2.
Radionuclide stack emissions measured at elemental
phosphorus plants (1975-1980) ^a)
Parameter
FMC
Idaho
Stauffer
Montana
Monsanto
Tennessee
Rock processing rate (MT/y)">) 1.6E+6 5.3E+5 1.7E+6
Uranium-238 concentration
of rock (PCi/g)(c) 22.0 27.0 <
Calciner stacks emission rate (Ci/y):^e)
Uranium-238 1.2E-3 2.4E-4 2.2E-3
Uranium-234 1.3E-3 2.0E-4 3.2E-3
Thorium-230 2.2E-3 1.2E-4 1.4E-3
Radium-226 1.3E-3 3.5E-4 2.1E-3
Radon-222 - 8.0 9.6
Lead-210 3.0E-3 2.8E-1 4.8E-1
Polonium-210 6.9 2.0E-1 7.5E-1
Other stacks emission rate (Ci/y):
Uranium-238 4.0E-2 6.2E-4 l.OE-2
Uranium-234 4.6E-2 7.0E-4 l.OE-2
Thorium-230 5.3E-3 1.2E-3 1.2E-2
Radium-226 5.9E-3 1.1E-3 9.0E-3
Radon-222 - ND ND
Lead-210 1.5E-2 2.5E-3 ND
Polonium-210 4.0E-1 5.9E-3 2.7E-3
Fraction of input radionuclides
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-222
Lead-210
Polonium-210
emitted:
1.2E-3
1.4E-3
2.1E-4
2.0E-4
_
5.1E-4
2.1E-1
6.0E-5
6.2E-5
9.0E-5
9.8E-5
5.7E-1
2.0E-2
1.4E-2
1.4E-3
1.5E-3
1.5E-3
1.7E-3
1.1
5.6E-2
8.8E-2
'a'Emissions are in particulate form except for radon-222 which is
released in gaseous form.
'•"'These processing rates were those estimated for these plants at
time of emission testing.
^c'Uraniuin-238 and its daughter products are assumed to be present in
equilibrium in the rock.
'"'Calciner feed material was a blend of Tennessee and Florida
phosphate rock.
^e'Based on 8760 hours of plant operation.
6.3-5
-------
particuiates, (5) estimates of the efficiency of existing control
systems in removing radionuclides and particuiates, (6) descriptions of
the sampling methods and procedures used during the testing, and
(7) test parameters, such as sample volumes and flow rates used in
testing.
A brief description of the major results obtained during this
testing is presented in the following sections.
The limited sampling at the Monsanto, Soda Springs} Idaho, plant
was due to the unavailability of suitable sampling locations for more
detailed testing. The Monsanto plant releases its calciner off-gas
stream through a large diameter demister. Significant modifications to
the demister and installation of a stack extension are necessary before
emission testing equivalent to that conducted at FMC and Stauffer can
be made at the Monsanto plant, (For more details on sampling problems
at the Monsanto plant see RC84c.)
Resultsof1983-1984 Emission Testing
Process Samp> 1_es_
Table 6.3-3 presents the measured radiomiclide concentrations in
the calciner feed material and product samples for the three plants
studied. For the Stauffer and Monsanto plants, both the lead-210 and
polonium-210 concentrations in the calciner product samples were
significantly lower than the concentrations in the feed material,
reflecting the volatilization of these radionuclides during the
calcining operation. For the PMC plant, only the polonium-210 concen-
tration was significantly lower in the product samples than in the feed
material. This indicates that large quantities of lead™210 are not
volatilized during the calcining operation at the FMC plant,
RadionuclideEmission Rates
Table 6.3-4 presents the measured radionuclide emission rates for
the three plants studied in pCi/hr/calciner and the estimated annual
calciner emission rates. The estimated annual polonium-210 emission
rates are: Monsanto, Soda Springs, Idaho = 21 Ci/yr; FMC, Pocatello,
Idaho = 8,6 Ci/yr; and Stauffer, Silver Bow, Montana = 0.74 Ci/yr. The
estimated annual lead~210 emission rates are: Monsanto, Soda Springs,
Idaho = 5.6 Ci/yr; FMC, Pocatello, Idaho = 0.12 Ci/yr; and Stauffer,
Silver Bow, Montana =0.11 Ci/yr.
ParticleSize Distribution
Table 6.3-5 presents the measured distribution of lead-210 and
polonium-210 by particle size in the calciner off-gas streams at the
FMC and Stauffer plants. These samples were collected using Andersen
cascade irapactors. Similar samples could not be collected at the
6.3-6
-------
Monsanto plant because suitable sampling ports and locations were not
available (RC84c). These data show that for both the FMC and Stauffer
plants, most of the poloni.um-210 was associated with submicron parti-
cles. For the FMC plant, an average of 73 percent of the poloniunt-210
was in a particle size range less than 0.5 microns and 86 percent was
in a range less than 1.5 microns. For the Stauffer plant, an average
of 53 percent of the polonium~21G was in a particle size range less
than 0.5 microns, and about 90 percent was in a range less than 1.5
microns,
Table 6.3-3. Measured radionuclide concentrations
in process samples at elemental phosphorus plants
(1983-1984 emission test results)
R a d i on uc1i d e cone en trations (jgCjl/g_j_
. Fe e d s t oc k Calcined product
Uranium- Lead- Polonium- Uranium- Lead- Polonium-
238 210 210 238 210 210
FMC
Pocatello, 21 26 21 22 27 8
Idaho
Stauffer
Silver Bow, 42 46 40 42
Montana
Monsanto
Soda Springs, 32 150 91 37
Idaho (a)
'a'Blended feed material. This plant recycles both dropout chamber
dust and underflow solids from wet scrubber clarifier.
6.3-7
-------
Table 6.3-4. Radionuclide emissions from caLciners at elemental
phosphorus plant (1983-1984 emission test results)
Plant
Average measured
radionuclide emissions
(uCi/h/calciner)(a)
Uranium- Lead- Polonium-
238 210 210
Estimated total
Number calciner emissions
of (Ci/y)(b)(c)
calciners Uranium- Lead- Polonium-
238 210 210
FMC
Pocatello, 0.28 7.5 540 2
Idaho
Stauffer
Silver Bow, 0.04 7.6 50 2
Montana
Monsanto
Soda Springs, 0.78 760 2900 1
Idaho
0.004 0.12 8.6
0.0006 0.11 0.74
0.006 5.6 21
the FMC plant, emission rates were measured from both calciner units,
and the reported values are the average emission rates for these units. For
the Stauffer plant, emissions for only one of the calciner units (kiln-2)
were measured, and the reported values are the average value for this unit.
In estimating total annual emissions, it was assumed that both calciner
units have the same emission rates.
'"'Based on 7400 hours of calciner operation (i.e.
factor).
85 percent operating
'•'-'The conversion of measured emission rates to annual emission estimates
for the FMC plant includes an adjustment for processing rate where appli-
cable (see EPA84a),
= 3-
-------
Table 6.3-5, Measured distribution of lead-210 and polonium-210
by particle size in calciner stack outlet streams at elemental
phosphorus plants (1983 emission test results)^a'
Plant
Approximate
particle size range
(D-50)(microns)
Percent of total
Lead-210
Polonium-210
FMC
Pocatello,
Idaho
Stauffer
Silver Bow,
Montana
>10
3-10
1,5-3
0.9-1.5
0.5-0.9
<0.5
>10
3-10
1.5-3
0.9-1.5
0.5-0.9
<0.5
10
13
9
10
14
44
<1
3
5
14
22
53
7
5
4
6
5
73
2
4
4
17
25
50
^•'Particle size measurement using cascade impactors could not be made
at Monsanto, Soda Springs, Idaho,' because suitable sampling ports and
locations were not available.
Lung-Clearanee Classification Studies
Samples of particulates collected frotn the calciner off-gas
streams at FMC and Stauffer were sent to the Pacific Northwest
Laboratory for testing to determine the lung-clearance classifications
(for use in ICRP lung model)(ICRP66) of lead-210 and polonium-210 in
these particulates. These lung-clearance classifications were
determined by measuring dissolution rates of these radionuclides in
simulated lung fluid. For each plant, testing was conducted on samples
containing particulates in the range of 0 to 3 microns and 3 to 10
microns. A detailed description of the test methods used and results
obtained are presented in PNL-5221 (Ka84). Table 6,3-6 summarizes the
dissolution data for lead~210 and polonium-210 in simulated lung fluid
for these particulate samples.
The results of these tests show that both the lead-210 and the
polonium-210 dissolved only very slowly in the simulated lung fluid.
More than 99 percent of these radionuclides remained undissolved even
after 60 days of testing. Based on these tests, it was concluded that
both lead-210 and polonium-210 in these materials should be considered
Class Y for calculations with the ICRP lung model (i.e., the model used
in EPA in dose calculations).
6.3-9
-------
Table 6.3~6. Dissolution of lead-210 and polonium-210
from parCiculate samples collected from off-gas streams
at FMC and Stauffer elemental phosphorus plants
Sample
„, . particle size
Plant / . %
(micron)
FMC 0-3
Pocatello,
Idaho
3-10
Stauffer 0-3
Silver Bow,
Montana
3-10
Dissolution
t ime ( days )
1.0
3.0
10.0
20.2
37.0
59.0
1.0
3.0
10.0
20.2
37.0
59.0
1.0
2.9
8.9
20.8
40.8
59.0
1.0
2.9
8.9
20.8
40.8
59.0
Fraction of
210pb remaining
undissolved
0.9984
0.9973
0.9968
0,9962
0.9956
0.9950
0.9933
0.9744
0.9682
0.9618
0.9554
0.9490
0.9999
0.9999
0.9994
0.9991
0.9983
0.9978
1.0000
0.9999
0.9990
0.9991
0.9985
0.9979
Fraction of
210po remaining
undissolved
0.9997
0.9990
0.9984
0.9980
0.9979
0.9978
0.9991
0.9988
0.9979
0.9970
0.9943
0.9914
0.9997
0.9996
0.9989
0.9986
0.9981
0.9980
0.9997
0.9993
0.9992
0.9948
0.9942
0.9940
6.3-10
-------
6,3.5 HealthImpact A^£Ji5jE!B!i!l^^ Plants
Tables 6.3-7 and 6.3-8 show the estimated annual calciner emission
rates and stack parameters for each of the six operational elemental
phosphorus plants. These values were used in. estimating the radiation
doses and fatal, cancer risks from these plants.
Table 6.3-9 presents the radiation doses to the lung from radio-
nuclide emissions from calciners at elemental phosphorus plants. Almost
all of the radiation risk from radionuclide emissions from calciners at
these plants results from these lung doses. The lung-clearance
classifications and particle size distributions (AMAD) used in
estimating these doses (ICRP Task Group Lung Model) are shown below:
Clearance Particle Size
RadJ.£micHde_ Class i f ic at ion AMAD
Lead~210, Polonium~210 Y(a) 0.3
Uranium-238, Uranium-234, Y l
Thorium-230
Radium-226 W^b) l^b)
'a'Based on experimental data obtained during emission testing.
^ 'Based on values recommended by ICRP (ICRP66) when experimental
values not available.
Table 6,3-10 presents estimates of the lifetime risk to the nearby
individuals and the number of fatal cancers to the regional population
from radionuclide emissions from calciners at elemental phosphorus
plants. The doses and rislcs to the nearby individuals were calculated
for a location 1500 meters from the plant in the predominant wind
direction. The doses and risks to the regional population were
calculated using the population distribution of the actual plant site.
Table 6.3-11 shows the number of people living within 80 km of these
sites and the source of the meteorological data used in these
calculations.
The fatal cancer risks from radionuclide emissions from calciners
at elemental phosphorus plants result primarily front inhalation of
polonium-210. To illustrate this point, Tables 6,3-12 through 6.3-15
show the doses to the various organs and the relative significance of
various pathways, organs, and radionuclides to the fatal cancer risks
from radionuclide emissions from calciners at both the FMC and
Monsanto, Idaho, plants.
6.3-11
-------
Table 6.3-7. Estimated annual radionuclide emissions
from elemental phosphorus plants^a^
Emissions (Ci/y)
Plant
Uranium-238^b-) Lead-210
Poloniura~210
FMC(c>
Pocatello, Idaho
Monsanto^c '
Soda Springs, Idaho
Monsanto'-0-'
Columbia, Tennessee
Stauffer^
Silver Bow, Montana
Stauffer^)
Mt. Pleasant, Tennessee
Occidental^
Columbia, Tennessee
4E-3
6E-3
2E-3
6E-4
2E-4
2E-4
0.1
5.6
0.4
0.1
0.05
0.05
9
21
'0.
0.
0.
0.
6
7
1
1
(b)
'Emission rates based on 7400 hrs per year of calciner operation
(i.e., 85 percent operating factor).
(c)
(d)
In using these data in estimating radiation doses and risks for
these plants, equal quantities of uranium-234, thorium-230, and
radium-226 were assumed to be emitted along with the uranium-238.
This assumption is supported by data in Table 6.3-2 which shows that
uranium~238 is in equilibrium (within about a factor of 2) with
uranium-234, thorium-230, and radium-226 in the calciner off-gas
streams. In any case, however, as noted previously, these radio-
miclides do not contribute significantly to the doses and risks from
radionuclide emissions from calciners at elemental phosphorus plants.
Based on measurements during EPA testing.
Estimates based on the following percent releases of radionuclides
entering the calciners: polonium-210 = 10 percent, lead-210 = 5 per-
cent uranium-238 = 0.02 percent (i.e., similar to percent releases
for the reference plant in EPA83).
6.3-12
-------
Table 6*3-8. Calciner stack emission characteristics
Plant
FMC
Pocatello, Idaho
Monsanto
Soda Springs , Idaho
Monsanto
Columbia, Tennessee
Stauffer
Silver Bow, Montana
Stauffer
Ml. Pleasant, Tennessee
Occidental
Columbia, Tennessee
Stack height Heat emission
(meters) (calories/sec)
30 8.8E+5
31 2.0E+6
35 l.OE+6
27 3.0E+4
35 6.0E+5
31 L.2E+6
Table 6.3-9. Radiation dose to lung from radion.ucLi.de
emissions from elemental phosphorus plants
Plant
FMC
Pocatello, Idaho
Monsanto
Soda Springs, Idaho
Monsanto
Columbia, Tennessee
Stauffer
Silver Bow, Montana
Stauffer
Mt. Pleasant, Tennessee
Occidental
Columbia, Tennessee
Nearby individuals Regional population
Lung (mrem/y) Lung (person-rem/y }
290 1170
610 750
30 310
60 122
6 33
5 65
6.3-13
-------
Table 6.3-10. Fatal cancer risks from radionuclide
emissions from elemental phosphorus plants
Plant
Lifetime risk to
nearby individuals
Regional population
{Fatal cancers/y
of operation)
FMC 5E-4
Pocatello, Idaho
Monsanto 1E~3
Soda Springs , Idaho
Monsanto 6E-5
Columbia, Tennessee
Stauffer 1E-4
Silver Bow, Montana
Stauffer 1E-5
Mt. Pleasant, Tennessee
Occidental 9E-6
Columbia, Tennessee
0.027
0.018
0.007
0.003
0.001
0.002
Table 6.3-11, Population within 80 km of elemental phosphorus plants
and source of meteorological data used in dose and risk calculations
Plant
Number of people
within 80 km^
Source of
meteorological
FMC
Pocatello, Idaho
Monsanto
Soda Springs, Idaho
Stauffer
Silver Bow, Wyoming
Monsanto
Columbia, Tennessee
Stauffer
Mt. Pleasant, Tennessee
Occidental
Columbia, Tennessee
1.4E+5
8.0E*4
7.7E+4
7.7E+5
6.0E+5
8.0E+5
Pocatello, Idaho
Pocatello, Idaho
Butte, Montana
Nashville, Tennessee
Nashville, Tennessee
Nashville, Tennessee
(fl>Based on 1970 Census.
a from National Climatic Center, Asheville, North Carolina.
6.3-14
-------
Table 6.3-12. Radiation dose rates to various organs
from radionuclide emissions from calciners
at elemental phosphorus plants^3'
IMC Monsanto
Organ Pocatello, Idaho Soda Springs, Idaho
(mrem/y) (mrem/y)
Lung
Kidney
Liver
Endosteutn
Red Marrow
290
10
2
1
0.3
610
18
4
6
0
.9
'Doses to individuals located 1500 meters from the plant in
predominant wind direction.
Table 6.3-13. Fatal cancer risks to nearby individuals
from radionuclide emissions from calciners at
elemental phosphorus plants by cancer type
Lifetime risk to nearby individuals
Cancer FMC Monsanto
Pocatello, Idaho Soda Springs, Idaho
Lung 5E-4 1E~3
Urinary 2E-6 3E~6
Liver 1E-6 2E-6
Lukemia 3E-7 9E~7
Bone 3E-8 1E-7
6.3-15
-------
Table 6.3-14, Fatal cancer risks to nearby individuals
from radionuclide emissions from calciners at
elemental phosphorus plants by radionuclide^8'
Percent of total risk
Radionuclide FMC Monsanto
Pocatello, Idaho Soda Springs, Idaho
Uraniura-234, -238 0.2 0.2
Thorium-230 0.1 0.1
Radium-226 0.01 0.01
Lead-210 1 26
Polonium-210 98 74
^a'These estimates do not include contributions from radon-222
emissions from the calciner. However, previous estimates (EPA83)
showed that radon~222 emissions from calciners at elemental phos-
phorus plants cause only small additional fatal cancer risks, i.e.,
about one percent of the total risk.
Table 6.3-15. Fatal cancer risks to nearby individuals from
radionuclide emissions from calciners at elemental
phosphorus plants by pathway
Pathway
Inhalation
Ingestion^3'
Other
Percent of
FMC
Pocatello, Idaho
99.3
0.7
<0.1
total risk
Monsanto
Soda Springs, Idaho
99.5
0.5
<0.1
intakes used were those for an urban/low productivity site
(see Appendix A).
6.3-16
-------
6.3.6 Alternative Control Technology
An analysis of the cost and polonium~210 removal efficiency for
alternative control systems for reducing polonium-210 emissions from
calciner off-gas streams at the FMC and Monsanto Idaho plants was
carried out for EPA by the Midwest Research Institute (MRl84a, MRl84b).
A summary of these analyses is shown in Table 6.3-16. These plants
were analyzed because they have the highest polonium-210 emissions.
Reducing the polonium-210 emissions will also reduce the lead-210
emissions.
Tables 6.3-17 and 6.3-18 show the risk reduction and cost of con-
trol at various selected polonium-210 emission rates for the FMC and
Monsanto (Idaho) plants, respectively. A more detailed analysis of the
costs and risk reductions, as well as the economic impacts, of
alternative polonium-210 emission rates for these plants is presented
in a regulatory impact analysis of emission standards prepared for EPA
by Jack Faucett Associates (EPA84d).
Table 6.3-16. Cost of alternative control systems for
reducing polonium-210 emissions at FMC and
Monsanto elemental phosphorus plants'3'
Control
system
Scrubber
15-in AP
30-in &P
45-in &P
ESP
200 SCA(b
300 SCA
400 SCA
Fabric
filter
210-Po
removal
(%)
65
77
83
> 72
83
90
98
FMC
Capital
cost
( $ millions)
2.1
2.8
3,7
5.2
5.9
6.7
7.3
Plant
Annualized
cost
( $ millions)
1.6
2.5
3.5
1.4
1.5
1.7
1.9
Monsanto Plant
Capital
cost
(& millions)
1.1
1.5
2.0
2.9
3.2
4.3
4.2
Annualized
cost
( $ millions)
0.9
1.4
2.0
0.8
0.9
1.1
1.3
(a)From Midwest Research Institute Reports (MRl84a and MRl84b) and based
on January 1984 dollars.
^b^SCA-Specific Collection Area in ft2/1000 acfm.
6.3-17
-------
Table 6.3-17. Cost of added controls and risk reduction at selected
poloniura-210 emission rates from calciners at FMC plant
Fatal cancer risks
Polonium-
210
emission
rate (Ci/y)
Current
emissions
2.5
Lifetime risk
to nearby
individuals
5E-4
1E-4
Regional
population
(Fatal
cancers/y
of operation)
0.027
0.008
Risk reduction
Regional
population Control
(Fatal system
cancers/y
of operation)
0.019 Medium
Cost
( $ millions)
capital-
annualized
5.9 1.5
5E-5
0.003
energy
ESP
0.024 High 6.7 1.7
energy
ESP
Table 6.3-18. Cost of added controls and risk reductions at selected
polonium-210 emission rates from calciners at Monsanto (Idaho) Plant
Fatal cancer risks
Polonium-
210
emission
rate (Ci/y)
Current
emissions
10
2,5
1.0
Lifetime risk
to nearby
individuals
1E-3
5E~4
IE-4
5E-5
Regional
population
(Fatal
cancers/y
of operation)
0.018
0.009
0.002
0.001
Risk reduction
Regional
population
(Fatal
cancers/y
of operation)
0.009
0.016
0.017
Control
system
-._-.
15 in AP
scrubber
High
energy
ESP
Fabric
filter
Cost
($ millions)
capital-
annual ized
1.1 0.9
4.3 1.1
4.2 1.3
6.3-18
-------
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~~Ihe Thermal Plant,
ORP/LV-77-3, EPA, Office of Radiation Programs, Las Vegas,
Nevada, 1977.
EPA83 Environmental Protection Agency, Office of Radiation
Programs, Draft Background Information Document, Proposed
Standards for Radionuclides, EPA 520/1-83-001, EPA, Office of
Radiation Programs, Washington, D.C., March 1983.
EPA84a Environmental Protection Agency, Office of Radiation
Programs, Emissions of Lead~210 and Polonium-210 from
Calciners at Elemental Phosphorus Plants: FMC Plant,
Pocatello, Idaho, EPA, Office of Radiation Programs,
Washington, B.C., June 1984.
EPA84b Environmental Protection Agency, Office of Radiation
Programs, Emissions of Lead-210 and Polonium-210 from
Calciners at Elemental Phosphorus Plants: Stauffer Plant,
Silver Bow, Montana, EPA, Office of Radiation Programs,
Washington, B.C., August 1984.
EPA84c Environmental Protection Agency, Office of Radiation
Programs, Emissions of Lead-210 and Polonium-210 from
Calciners at Elemental Phosphorus Plants: Monsanto Plant,
Soda Springs, Idaho, EPA, Office of Radiation Programs,
Washington, B.C., August 1984.
EPA84d Environmental Protection Agency, Regulatory Impact Analysis
of Emission Standards for Elemental Phosphorus Plants, Office
of Radiation Programs, Washington, D.C., EPA 520/1-84-025,
October 1984.
Ka84 KaIkwarf D.R. and Jackson P.O., Lung-Clearance Classification
of Radionuclides in Calcined Phosphate Rock Dust, PNL-5221,
Pacific Northwest Laboratories, Richland, Washington, August
1984.
6.3-19
-------
REFERENCES (Continued)
ICRP66 International Radiological Protection Commission Task Group
on Lung Dynamics, Deposition and Retention Models for
Internal Dosirnetry of Human Respiratory Track, Health Phys.
12:173-207, 1966.
MRl84a Midwest Research Institute, Analysis of Achievable Po~-210
Emissions and Associated Costs for PMC's Pocatello, Idahos
Plant, Midwest Research Institute, Raleigh, North Carolina,
August 1984.
MRl84b Midwest Research Institute, Analysis of Achievable Po-210
Emissions and Associated Costs for Monsanto's Soda Springs,
Idaho, Plant, Midwest Research Institute, Raleigh, North
Carolina, September 1984.
RC84a Radian Corporation, Emission Testing of Calciner Off-gases at
FMC Elemental Phosphorus Plant, Pocatello, Idaho, Volumes I
and II, Prepared for U.S. Environmental Protection Agency
under Contract No. 68-02-3174, Work Assignment No. 131,
Radian Corporation, P.O. Box 13000, Research Triangle Park,
NC, 1984.
RC84b Radian Corporation, Emission Testing of Calciner Off-gases at
Stauffer Elemental Phosphorus Plant, Silver Bow, Montana,
Volumes I and II, Prepared for U.S. Environmental Protection
Agency under Contract No. 68-02-3174, Work Assignment No.
132, Radian Corporation, P.O. Box 13000, Research Triangle
Park, NC, 1984.
RC84c Radian Corporation, Emission Testing of Calciner Off-gases at
Monsanto Elemental Phosphorus Plant, Soda Springs, Idaho,
Volumes I and II, Prepared for U.S. Environmental Protection
Agency under Contract No. 68-02-3174, Work Assignment No,
133, Radian Corporation, P.O. Box 13000, Research Triangle
Park, NC, 1984.
6.3-20
-------
Chapter 7: MINERAL EXTRACTION INDUSTRY FACILITIES
Metal Mines,Hills, and Smelters
Almost all industrial operations involving the removal and
processing of ores to recover metals release some radionuclldes 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.
For the aluminum industry the assessment includes emissions from
an alumina plant and aluminum reduction plants. The assessments of the
copper and zinc industries include assessments of mine, mill, and
smelter emissions. Finally, smelter emissions for the lead industry
are assessed.
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 (R^OgK 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.
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 5 of the 12 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
7.1-1
-------
Table 7.1-1. Location and size of primary aluminum production plants
(DOI80)
Location
Alabama
Arkadelphia
Jones Mills
Listerhill
scottsboro
Indiana
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
Mew Madrid
Kaiser Aluminum & Chemical Corp, 215
Consolidated Aluminum Corporation 30
Eastalco Aluminum Company
Noranda
145
115
Montana
Columbia Falls
North Carolina
Badin
Anaconda Aluminum company
Aluminum Company of America
148
103
Massena
Massena
Ohio
Hannibal
Oregon
The Dalles
Troutdale
Tennessee
Alcoa
New Johnsville
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
7.1-2
-------
fable
1-1. Location and size of primary aluminum production plants
(Continued)
Location
Texas
Point Comfort
Palestine
Rockdale
San Patrlcio
Washington
Ferndale
Goldendale
Longview
Mead
Ravenswood
Tacoma
Vancouver
Wenatchee
Total
Company
Capacity
(1000 MT/y)
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
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).
7.1-2 Pjrocess_Descri£t ion
Bauxite ore is processed at the alumina plant to produce alumina
using a modified "American Bayer" process. EPA measurements indicate
that the ore is elevated in both uranium-238 and thorium-232 with
concentrations of 6.8 and 5.5 pci/g (EPA82). The data in Table 7.1-2
show that most of the radioactivity in the ore is associated with the
impurities rather than the alumina product.
Of the 32 aluminum reduction plants in the United States, ail but
one produce aluminum in electric furnaces (cells) by the Hall-Hiroult
process. In the Hall-Hiroult process, alumina (Ai^Og) 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.
7.1-3
-------
Table 7,1-2. Radionuclide concentrations in alumina
plant process samples (EPA82)
Samp
Uranium- 238 Thorium- 232
Bauxite ore 6.8 5.5
Alumina kiln feed 0.05 0.05
Alumina product 0,28 0.2
Red mud 7.5 5.0
Brown mud 5.5 12.5
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 temperatures 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 MT of aluminum. The breakdown
of raw materials is shown in Table 7.1-3.
Table 7.1-3. Raw materials used in producing aluminum (EPA77)
Raw material MT Feed/MT ftl produced
Alumina (A1203) 1.9
Cryolite (NaAlF6) 0.03-0.05
Aluminum Fluoride (Al?3) 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
7.1-4
-------
aluminum produced, with 1.5 kg/MT being a typical value (EPA76).
Participate emissions generated by the cells vary from 5.95 to 88.5
kg/MT, with 40.65 kg/MT being typical (EPA76).
Materia 1 s
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 7.1-4.
Table 7.1-4. Radionuclide concentrations of
feed materials to aluminum plants (KPA82)
Radionuclide concentration (pCi/g)
Feed material __. ...
Uranium-238 ~ ~~ Thorium 232"
Alumina 0.10 <0.2
Aluminum Fluoride O.U <0.2
Cryolite 0.11 <0.2
7.1.3 Con t r o 1 Techno logy for P r _ima ry_ Aluminumgejduc_t lonJPl arit_s
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 to 99-f 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).
7.1.4 Radiqnuc1ide Emissions
Emissions from the alumina kilns and red mud kilns at an alumina
plant are given in Table 7.1-5. The low radioactivity of alumina is
reflected in the low radionuclide emissions from the alumina kilns.
Emissions of radionuclides from the red mud sinter kiln were below
measurable concentrations except for lead-210, polonium 210, and
7.1-5
-------
radon-222. The high temperature of the kiln causes a large fraction of
lead-210 and polonium 210 to be volatilized.
Table 7.1-5. Radlonucllde emissions from the surveyed alumina plant
(RPA82)
Etnissions (Ci/y)
Radlonuc1ide "
Alumina kilns Red mud kilns
Uranium-238
Uranium 234
Radium 226
Radon- 222
Lead-210
Polonium-210
8 . 7E--3
5.7E-3
2.71-3
2,75
4.8E-2
4.0E-2
Particulate material emitted from an aluminum reduction plant
contains radionuclide concentrations (pCi/g) similar to or greater than
the concentrations in the alumina processed. Because of the high
temperatures of the reduction cells, lead-210 and polonium-210 are
volatilized and released in greater quantities than the other
radioriuclides in the alumina. EPA has measured the radionuclide
emissions from an aluminum reduction plant. The emission estimates for
the reference aluminum reduction plant are based on data from these
measurements.
7.1.5 Reference Fac 1.1 it.les
Measured emissions from a single alumina plant were used to
estimate health impacts for alumina production.
Table 7.1-6 describes the parameters of a reference aluminum
reduction plant which are used to calculate the radionuclide emissions
to air and the resulting health impacts and to give a general idea o£
plant parameters.
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).
7.1-
-------
Table 7.1-6. Reference aluminum reduction plant
Parameter Value
Capacity 136,000 MT/y aluminum
Capacity factor 0.94
Type of equipment Center-worked prebake cells
Stack Parameters
Main stack
Height 36 m (4 stacks)
Diameter 3 m
Exit gas velocity 80 m/s
Exit gas temperature 160° C
Roof monitor
Height 10 m
Diameter 1.2 m
Exit gas velocity 0.01 m/s
Exit gas temperature 37° C
Anode bake plant
Height 30 m
Diameter 1.8 m
Exit gas velocity 4.5 m/s
Exit gas temperature 96° C
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 reference 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.
Sadionuclide emissions for the reference plant were based on
actual measurements of radionuclide concentrations in the particulate
emissions from an existing plant. The resulting releases are listed in
Table 7.1-7.
7.1-7
-------
Table 7.1-7. Radionuclide emissions from the
reference aluminum reduction plant
Emissions (Ci/y)
Uranium-238
Uranium-234
Thorium-230
Ra d i urn- 226
Lead-210
Polonium~210
Thorium-232
Radium-228
Main stack
6.8E-5
6.8E-5
2.4E-4
5.5E-5
3.2E-4
2.7E-4
—
' Roof monitor
8.1E-9
8.1E-9
3.8E-8
7.4E-9
2.0E-7
2.0E-7
2.9E-8
2.9E-8
Anode bake plant
8.0E-5
8.0E-5
4.0E-5
6.0E-5
2.0E-4
2.0E-4
3.2E-5
3.2E-5
main stack emissions were used to calculate doses from
aluminum reduction plant.
7.1.6 Health Impact Assessment
The estimated annual radiation doses and health risks from the
emissions from the alumina plant are given in Tables 7.1-8 through
7.1-10. These estimates are for a rural site with a regional population
of 6E+5.
Table 7.1-8. Radiation dose rates from radioactive participate
emissions from the surveyed alumina plant
Organ
Nearby individuals
(mrem/y)
Regional population
(person-rem/y)
Lung
Red marrow
Endos teum
Breast
Liver
1.7E-2
0.2
2.0
7.5E-2
4.6E-1
15.8
2.63
38.0
0.26
1.35
7.1-8
-------
Table 7.1-9. Annual radon decay product exposures from radort-222
emissions from the surveyed alumina plant
Nearby individuals Regional population
Source "%r.T , 6, ^ 5TT ,
(.WL-y; (.person-WL-y;
Stack 8.5E-10 8E-4
Table 7.1-10. Fatal cancer risks from radionuclide
emissions from the surveyed alumina plant
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Particulates
Radon-222
Total
1E-6
9E-10^a) 4E-lo(b)
1E-6
4E-4
]_E-§(a)
4E-4(a)
SE-S^b)
4E-4(b)
on BEIR-3, NRPB, and EPA models (see Chapter 8, Volume I).
(b)]Jased on USCEAR and ICRP risk estimates (see Chapter 8, Volume I).
The estimated annual radiation doses from radionuclide emissions
from the reference aluminum reduction plant are listed in Table 7.1—11,
These estimates are for a rural site with a regional population of
2.7E+5.
Table 7,1-12 presents estimates of the lifetime risk to nearby
individuals and number of fatal cancers per year of operation from
these doses.
Table 7.1-11. Radiation dose rates from radionuclide
emissions from the reference aluminum reduction plant
-. Nearby individuals Regional population
(mrem/y) (person-rem/y)
Lung
Red marrow
Endosteum
Breast
Liver
Kidney
7.2E-4
0.1
5.3E-1
6.9E-2
3 . 7E-1
1.2
1.29
.35
1.63
,23
1.10
4.06
7.1-9
-------
Table 7,1-12. Fatal cancer risks due to radionuclide emissions
from the reference aluminum reduction plant
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Aluminum
reduction plant 8E-7 7E-5
(particulates)
7.1.7 Existing Emission Standards andAir Pollution Controls
No Federal or state regulations currently exist that limit
radionuclide emissions from alumina plants or aluminum reduction
plants. Particulate emissions from these sources are limited to the
quantities established by the states in their State Implementation
Plans (SIPs) for meeting Ambient Air Quality Standards.
Several states have established specific SIP limits for aluminum
reduction plants, ranging from 15 to 20 Ibs/ton of aluminum produced.
In states where no specific limits have been established for aluminum,
emissions from these sources are regulated according to the limits
established in the SIPs for general processing sources.
7.1-10
-------
REFERENCES
DQ180 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 Participate Emissions,
EPA-450/3-77-010, March 1977.
EPA79 Environmental Protection Agency, Primary Aluminum: Draft
Guidelines for Control of Fluoride 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.
7.1-11
-------
Page Intentionally Blank
-------
7.2 CongerIndustry
7,2.1 GeneralDescription
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).
ftll 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 particuiate 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 particuiate
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
(Sc79)
Plant
location
Arizona
Hayden
Miami
Hayden
San Manuel
Morenci
Douglas
A jo
Michigan
White Pine
New Mexico
Hurley
New Mexico
Hidalgo
New York
McGill
Tennessee
Copper Hill
Texas
El Paso
Utah
Gar fie id
Washington
Tacoma
Total
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,
ASARCO, Inc.
Capacity
(1000 Mf)
163
136
73
181
161
115
63
82
73
127
45
20
104
254
91
1688
First year
of operation
1890
1958
1950
1942
1910
1950
1905
1939
1976
1907
1845
1905
1907
1890
^Rebuilt as of 1979.
7.2-2
-------
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 as 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 to 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 Cpnt rojl Techno logy
Of the 15 primary copper smelters currently operating, 11 use
reverberatory furnaces and 7 have roasters. Of these 7, 4 use
multi-hearth roasters while the other 3 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 SO2 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
EPA has recently carried out radionuclide measurement studies at
both an underground copper mine and mill and an open pit copper mine and
mill (EPA82). The results of these studies indicate that radon-222 is
the only significant radionuclide emitted from the underground mine. At
the open pit mine and mill, radioactive particulates and radon-222 are
emitted, primarily during truck loading and dumping and crushing
operations.
The measurement studies also included analysis of radioactivity in
various process samples. Table 7.2-2 lists the uranium-238 and
thorium- 232 concentrations in process samples from both the underground
mine and mill and the open pit mine and mill.
7.2-3
-------
Table 7.2-2. 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
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 poloniura-210)
may be volatilized and released in greater quantities than the other
radionuclides in the ore concentrates.
Very little information has been available to date on radionuclide
emissions from copper smelters. EPA has recently surveyed two copper
smelters, and the data from these studies were used in estimating
radionuclide emissions from the reference copper smelter.
7.2.5 Reference Facilities
Actual emissions data from EPR's measurement studies were used to
assess potential health impacts from the underground mine (Table 7.2-3)
and open pit mine and milling complexes (Table 7.2-4).
Table 7.2-3. Radionuclide emissions from the
underground copper mine (EPA82)
_., Emissions
Radionuclide ,„,, .
(Ci/y)
Radon-222 6.5
Table 7.2-5 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 56,000
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 con-
trol equipment applied to the reference facility was chosen to repre-
sent typical equipment on actual copper sraelters.
7.2-4
-------
Table 7.2-4, Radionucllde emissions from copper mill,
open pit mine, and concentrator (EPA82)
„ ,. ... Emissions
Radionuciide ,_. , ,
Uranium-238 3.1E-4
Uranium-234 3.8E-4
Radium-226 1.8E-4
Radon-222 1.9
Lead-210 1.9E-3
Total annual emissions of radionuclides from the reference copper
smelter are given in Table 7.2-6. These values were derived from 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 7.2-5. Reference copper smelter
Parameter Value
Capacity 56,000 MT/y
Capacity factor 0.75
Type of equipment used Reverberatory furnace
Stack Parameters
Main stack
Height 183 m
Diameter 2,6 m
Exhaust gas velocity 28 m/s
Exhaust gas temperature 135° C
Reid plant
Height 30.4 m
Diameter 1.8m
Exhaust gas velocity 16.5 m/s
Exhaust gas temperature 79° C
Particulate Emission Rate
Main stack 247 kg/h
Acid plant 11 kg/h
7.2-5
-------
Table 7.2-6. Radionuclide emissions front the reference
copper smelter (southwestern site)
Radionuclide
Uranium-238
Uranium~234
Thorium-230
Radium-226
Lead-210
Poloalum~210
Thorium-232
Thorium-228
Emissions
(Ci/y)
4.0E-2
4.0E-2
2.1E-3
1.5E-3
6.5E-2
3.0E-2
1 . 2E-3
1.3E-3
7-2.6 Health Impsct_ Ass e ssment
The estimated radiation doses from radionuclide emissions from the
underground mine and mill, the open pit mine and mill, and the
reference copper smelter are listed in Tables 7.2-7 through 7.2-10.
These estimates are for a low population density southwestern site with
a regional population of 3.6E+4.
Table 7.2-11 presents estimates of the lifetime risk to nearby
individuals and number of fatal cancers per year of operation resulting
from these doses.
Table 7.2-7. Annual radon decay product exposures from radon-222
emissions from the underground copper
Nearby individuals Regional population
(WL~y) (person-WL~y)
Mine vent 2.5E-5 4.2E-4
'a/Based on a ground level release.
7.2-6
-------
Table 7.2-8. Radiation dose rates from radioactive particulate
emissions from the open pit copper mine and
» Nearby individuals Regional population
(mrem/y) (person~rem/y)
Lung
Red marrow
Endosteum
Breast
Liver
1.1E+1
1.6
2.5E+1
2.2E-2
9.9E-2
2.1E-1
2.8E-2
4.5E-1
4.2E-4
1.8E-3
^a'Based on a 10-meter stack height.
Table 7.2-9. Annual radon decay product exposures from radon-222
emissions from the open pit copper mine and mill
_ Nearby individuals Regional population
Source J ,„_ % t TIT %
(WL-y) Cperson-WL-y)
Stack 7.2E-6 1.2E-4
Table 7.2-10. Radiation dose rates from radionuclide particulate
emissions from the reference copper smelter (southwestern site)
n Nearby individuals Regional population
(mrem/y) (person-rem/y)
Lung 2.0E-1 0.95
Red marrow 3.2E-3 1.4E-2
Endosteum 4.5E-2 0.21
Breast 2.9E-4 1.2E-3
Liver 2.2E-3 8.7E-3
7.2-7
-------
Table 7.2-11. Fatal cancer risks from radlonuclide
emissions from the underground copper mines the open pit copper mine
and mill 3 and the reference copper smelter
Source
Particulates
Radon-222
Total
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Underground copper mine
IE-7
4E-5U)
4E-5U)
2£-5
4E-8
Particulates
Radon-222
Total
Particulates
Open pit copper mine and mill
2E-5
9E-6(a) 4E-7
-------
REFERENCES
EPA82 Environmental Protection Agency, Emissions of Naturally
Occurring Radioactivity from Aluminum and Copper Facilities,
SPA 520/6-82-018, Las Vegas, Nevada, November 1982.
Sc79 Schroeder H. J., Mineral commodity Proflies--Copper, U.S.
Department of the Interior, Bureau of Mines, Washington, D.C.,
1979,
7.2-9
-------
Page Intentionally Blank
-------
3 g tnc Indus t i_y
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, U.S. demand for zinc metal has grown slowly,
but 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_Descjrigt_ioji
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 an
-------
Table 7.3-1. Location size of primary zinc production plants
CCa78)
, „ First year Capacity
Location Company _ . . ,_. , *• ,,m\
r J of operation (Thousands of MT)
Idaho Bunker Hill 1928 95
Kellogg
-------
Table 7.3-2. Radionuclide emissions from the zinc mine
and mill (EPA82)
Radionuclide
Uranium- 238
Uranium- 234
Thorium- 230
Radium- 226
Radon 222
Lead- 210
Polonium- 2 10
Thorium- 232
Thorium- 228
Emissions
(Ci/y)
1.8E-3
1.8E-3
1.5E-3
8 , 2E- 4
2.3E+2
2.6E-3
2.2E-3
6.QE-4
4.7E-4
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.
7.3,5 ReferenceFacilities
Actual emissions from a mine and mill complex (chosen because of
the high working level measurements reported for the mine, and high
production rates) were used to estimate health impacts from these
sources.
Table 7.3-3 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 (DO176). The flow rate was
derived by adjusting available data for differences in capacity and
capacity factor. The stack height and diameter were estimated from
available data.
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 (302) plant. Off-gases from the
electrolytic reduction step are vented directly to the atmosphere.
7.3-3
-------
The total annual radionuclide emissions for the reference zinc
smelter are listed in Table 7.3-4.
Table 7,3-3. Reference zinc smelter
Parameter
Value
Process
Capacity
Capacity factor
Radionuclide concentration
of input ore'a'
Uranium- 238
Thorium- 232
Stack Parameters
Number
Height
Diameter
Exhaust gas velocity
Exhaust gas temperature
Electrolytic reduction
88 E+3 KT/yr zinc
0.8
0.18 pCi/g
0,08 pCi/g
1
100 meters
2 meters
20 m/s
150° C
measurements by EPA (EPA82) at a zinc smelter.
Table 7.3-4. Radionuclide emissions from the reference zinc smelter
Radionuclide
Uranium- 238
Uranium- 234
Thorium- 230
Radium- 226
Radon- 222
Lead- 210
Polonium- 210
Thorium- 23 2
Thorium- 228
Emissions
(Ci/y)
5.6E-4
3.7E- 4
1 . 4E- 3
4.5E-3
2 . 8E- 1
2.5E- 2
1.5E-3
3 . 4E- 4
3.4E-4
7.3-4
-------
7.3.6. Hgalth jinga.c_t__Assessmeu^
The estimated annual radiation doses from the radiotiuclide emissions
of the zinc mines mill, and smelter are listed in Tables 7.3-5 through
7.3~8« These estimates are for a rural site with a regional population
of 6E+5. The lifetime risk to nearby individuals and number of fatal
cancers per year of operation are shown in Tables 7.3-9 and 7.3-10.
Table 7.3-5. Radiation dose rates from radioactive particulate
emissions froai the zinc mine and
.. Nearby individuals Regional population
Organ / , , 6, / \
(mrem/y) ^person-rem/yj
Lung
Red marrow
Endosteum
Breast
Liver
4.1
3.6E-1
5.8
4.7E-3
2.1E-2
14.1
1.27
20.1
1.8E-2
8.1E-2
(a'Based on Arkansas population.
Table 7.3-6. Annual radon decay product exposures from radon-222
emissions from the zinc mine and mi
_ Nearby individuals Regional population
(WL-y) (person-WL-y)
Mine vent 1E-4 4E-1
'a'Based on Arkansas population.
Table 7.3-7. Annual radon decay product exposures from radon-222
emissions from the zinc smelter
„ Nearby individuals Regional population
(WL-y) (person-WL-y)
Zinc smelter 5.3E-10 6E-5
7.3-5
-------
Table 7.3-8. Radiation dose rates from radionuclide
emissions from the reference zinc smelter^-3'
_ Nearby individuals Regional population
• (mrera/y) (person-rem/y)
Lung
Red marrow
Endosteum
Breast
Liver
l.OE-2
1 . 7E-3
2.2E-2
2.4E-4
1.4E-3
1.12
2.0B-1
2,5
2.7E-2
1.6E-1
'a'Based on Arkansas population.
Table 7.3-9. Fatal cancer risks from radionuclide
emissions from the zinc mine and mill'*^
„ Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Particulates
Radon-222
Total
7E-6
2E-4(b) 8E-5^C)
2E-4(b) 9E-5^C)
3E-4
8E_3(b) 4£_3(c)
9E-3^b) 4E-3(c)
on Arkansas population.
on BEIR-3, NRPB, and EPA models (see Chapter 8, Volume l).
on USCEAR and ICRP risk estimates (see Chapter 8, Volume I).
Table 7.3-10. Fatal cancer risks from radionuclide
emissions from the reference zinc smelter^3'
Lifetime risk Regional population
to nearby individuals (Fatal cancers/y of operation)
Particulates 2E-8 4E-5
Radon-222 7E-10(b) 3E-10(C) lE-6^b) 6E-7^C)
Total 2E~8(b»c) 4E-5(b,c)
(a)Based on Arkansas population.
(k)Based on BEIR-3, NRPB, and EPA models (see Chapter 8, Volume I).
(c)Based on USCEAR and ICRP risk estimates (see Chapter 8, Volume I).
7.3-6
-------
7.3.7 ExistjLng_Emtss.ion Standards and a.ir__polj.utlon__Con.trQ|.s
No Federal or state regulations currently exist that limit
radionuclide emissions from zinc smelting. Participate emissions £rom
zinc smelting are regulated by New Source Performance standards (*)SPS)
for plants built after October 1974, or by the limits established in
State Implementation Plans (Sips) for meeting Ambient Air Quality
Standards. The NSPS for zinc smelting is less than 50 mg/Dry Standard
Cubic Meters.
7.3-1
-------
REFERENCES
Ca78 Cammarota V. A., Jr., Mineral Commodity Profiles-Zinc,
MCP-12, U.S. Department of the Interior, Bureau of Mines, Hay
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.
7,3-8
-------
7,4 Le ad Indus t ry
7.4.1 General JDes£ri£jtlon
Galena (PbS), frequently containing cerussite (PbCO^) and
anglesite (PbSO^), 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 (RSARCO'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_ ..Dejscjripj: ion
Lead smelting involves three distinct processes: sintering, to
convert the ore from a r.ulfide to an oxide or sulfate form and prepare
the feed materials for furnaeing; furnacing, to reduce the oxide feed
to lead metal; and dross ing, 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 Contrp i_Techng;logjy_
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.
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
7.4-1
-------
Table 7.4-1. Location size of primary lead production plants
(D0177)
Location
Idaho
Missouri
Boss
Glover
Herculanium
Company
Bunker Hill
Amax- Homes take
RSRRCO
St. Joe Minerals
First year
of operation
capacity
(Thousands of
tons of Pb)
1911 117
1968 127
1968 100
1892 220
rebuilt 1970's
Montana
East Helena
ASARCO
1888
82
Texas
El Paso
Now shut down.
RSARCO
1887
82
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 participate
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
of f-gases enter the ESP or fabric filter.
Sinter off-gas is typically fed to an acid plant for recovery of
sulfur dioxide after particulate cleaning, as described above. Effi-
cient operation of the acid plant requires gases containing 5 percent
or more SC>2. The circuit of gases through the sinter machine may be
quite complex with weak (in sc^) gases being recirculated through an
upstream section of the machine to enrich the SC>2 content before
going to the acid plant.
B1ast 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.
7.4-2
-------
7.4.4 IiiOTiulde^ ^Emisjjigris
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. EPA has recently measured
the radionuclide emissions at a lead smelter, and results of these
measurements are used in this report. Radionuclide emissions are
presented in Table 7.4-3.
7.4.5 Re f erence Fac il i t y
Table 7.4-2 describes the parameters of the reference facility
which were used to describe the radioactive emissions to the atmosphere
and the resulting health impacts.
The reference lead smelter has a capacity of 220,000 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. For calculational purposes, however, emissions were treated as
coming from one stack.
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.
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.
7.4.7 Existing Emission Standards and Air Pollution Controls
No Federal or state regulations currently exist that limit
radionuclide emissions from lead smelting. Particulate emissions from
lead smelters are regulated by New Source Performance Standards (NSPS)
for plants built after October 1974, or by State Implementation Plans
(SIPs). The NSPS for lead sintering machines, blast furnaces, and
dross furnaces is less than 50 mg/Dry Standard Cubic Meters.
7.4-3
-------
Table 7.4-2. Reference lead smelter
Parameter Value
Capacity 2.2E+5 MT/yr lead
Capacity factor 0,92
Radionuclide concentration
of input ore:
Uranium-238 0.9 pCi/g
Thorium-232 0.5 pCi/g
stack Parameters
Number 1
Main stack
Height 30 meters
Diameter 1 meter
Exit gas velocity 9 m/s
Exhaust gas temperature 90°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
Radionuclide Emissions^8)
(Cl/y)
Uranium-238 8.6E~3
Uranium™234 8.6E~3
Thorium-230 7.3E-4
Radium 226 5.9E-4
Lead-210 2.6E-2
Polonium-210 2.IE-2
Thorium-232 7.0E-4
Thorium-228 7.OB-4
stack only.
7.4-4
-------
Table 7,4-4, Radiation dose rates from radionuclide
emissions from the reference lead smelter
Organ
Nearby individuals
(mrem/y)
Regional population
(person-rem/y)
Lung
Red marrow
Endosteum
Breast
Liver
4.8
1.2E-1
1.8
9.9E-3
5.9E-2
6.9E+1
1.8
2.6E+1
0.17
1. 01
Table 7.4-5. Fatal cancer risks due to radionuclide emissions
from the reference lead smelter
Source
Lifetime risk
to nearby individuals
Regional population
(Fatal cancers/y of operation)
Particulates
8E-6
1.6E-3
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.
DOI77 U.S. Department of the Interior, Lead Mineral Commodity
Profiles--Lead, Washington, D.C., December 1977.
RPA75 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.
I8RL79 Industrial Environmental Research Laboratory, Control of
Particulate Emissions in the Primary Nonferrous Metals
Industries, NTIS Report No. PB-80-151822, Cincinnati, Ohio,
December 1979.
7.4-6
-------
APPENDIX A
ASSESSMENT METHODOLOGY
-------
Page Intentionally Blank
-------
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-10
A, 7 Collective Assessment A-ll
A.8 AIRDQS-EPA Parameters and Input Data A-12
A.9 DARTAB—Dose and Risk Tables A-14
References A-21
TABLES
A-l Characteristics of the generic sites A-7
A-2 Sources of food for the maximum individual A-ll
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-15
A-5 Site independent parameters used for AIRDOS-EPA
generic site assessments A-17
A-6 Element dependent factors used in AIRDOS-EPA assessments A-19
A-3
-------
Page Intentionally Blank
-------
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, pj.ume 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, Cs 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 kin
(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 Billings3 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, as discussed in Volume I, Chapter 6, to
calculate the individual and collective radionuclide concentrations for
these assessments.
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
12.10 c
1000 in
9.23E+5 persons
1.72E+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
3.80S+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 Becord:
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=12S15)
A-E
74/01-74/12
142 cm
22.0° c
1000 m
1.55E+3 persons
1.51E+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
5.89E+5 persons
1.19E+4 head
2.57E+5 head
2.94E+3 ha
A-8
-------
Table A-l. Characteristics of the generic sites~-continu«
Site E —New Mexico
Meteorological data:
Stability Categories:
Period of Record:
Annual Rainfall:
Average Temperature:
Average Mixing Height:
Population: (0-8 kin):
(0-80 km):
Dairy Cattle (0-80 km):
Beef Cattle (Q-8Q km):
Vegetable Crop Area:
(0-80 km)
Grants/Gnt-MiIan (WBAN=93057)
A-F
54/01-54/12
20 cm
13.20 c
800 m
0 persons
3.6QE+4 persons
2.30E+3 head
8.31E+4 head
2.78EH-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)
Bill ings/Logan
A-F
67/01-71/12
20 cm
8.10 c
700 m
0 persons
1.30E+4 persons
1.86E+3 head
1.47E+5 head
1.77E+4 ha
(WBAN=24033)
A-9
-------
Table
Characteristics of the generic si.tes~~conti.nued
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 ra
4.17E+4 persons
1.40E+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.
Ground surface and soil concentrations are calculated for those
nuclides subject to deposition due to dry deposition and precipitation
scavenging.
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 Volume I, Chapter 8.
A.6 Individual Assessment
The nearby individuals were assessed on the following basis:
A-10
-------
1) The nearby individuals for each source category are 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
nearby individuals.
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.
3) 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 nearby individuals' diet comes from the assessment area or
from outside the assessment area.
Table A~2. Sources of food for the maximum individual
Food Urban/Low productivity Rural
(Sites A, Bs E-G) (Sites C & D)
Fl F2 F3 Fl F2 F3
Vegetables .076 0. .924 .700 0. .300
Meat .008 0. .992 .442 0. .558
Milk 0. 0. 1. .399 0. .601
Fl and F2 are the home-produced fractions at the individuals'
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).
A. 7
The collective assessment to the population within an 80 km radius
of the facility under consideration was performed as follows:
A-ll
-------
1) The population distribution around the generic site was based on
the 1970 census. The population was assumed to remain stationary in time,
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 b.e 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 AIRPOS-EPA Parameters and Input Data
Site independent parameter values'used for AIRD0S-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 deposition
velocity of 0.0018 m/s was used for participates and 0.035 m/s for
reactive vapors (e.g., elemental iodine) unless otherwise indicated.
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 JiTability Mray) meteorological data summaries
were obtained from the National Climatic Center, Ashevilles North
Carolina. Data for the station considered most representative for each
generic site were used. Generally, these data are from a nearby
A-12
-------
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,
Table A-3. Some site parameters used with A1RDOS-EPA
Average mixing Rainfall Scavenging
Generic height rate coefficient
site (m) (cm/y) (s~^)
Site A
Site B
Site C
Site D
Site E
Site F
Site G
1000
600
1000
600
800
700
615
102
102
142
127
20
20
27
l.OE-5
l.OE-5
1.4E-5
1.3E-5
2.0E-6
2.0E-6
2.7E-6
Dairy andBeefCattle
Dairy and beef cattle distributions are part of the AIRDOS-EPA
input. A constant cattle density is assumed except for the area closest
to the source or stack in the case of a point source, i.e., no cattle
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 NRG
(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 equiva-
lent 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.
Ve ge tab 1 e C r o pAre a
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
NRG 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).
A-13
-------
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 (CE0) and the geographic coordinates
of the population eentroid 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 250S000 CEDs in the United States with an
average population of about 300 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.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 (BeSO) using dose and risk conversion factors (see
Volume I, Chapters 7 and 3) to produce the dose and risk assessments for
this report.
-------
Table A~4. Cattle densities and vegetable crop
distributions for use with AIRBOS-EPA
State
Alabama
Arizona
Arkansas
California
Colorado
Connecticut
Delaware
Florida
Georgia
Idaho
11 lino is
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
#/kn»2
7.02E-1
2.80E-1
5.9QE-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.G7E-1
6.11
3.13
3.51
4.88
8.7GE-1
1,89
9,27E~2
8.78S-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-H
3.60
6.48
1.28E+1
1.43E+1
7.19
3.33E+1
3.34E + 1
7.40E+1
2.90E + 1
2 . 6 5E +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
km^/km^
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
S.97E-2
I.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~15
-------
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 Vlrgina
Wisconsin
Wyoming
Dairy cattle
density
#/km2
1.26
6.25E-1
4,56
7.13E-1
4.53E-1
6.46
2.30
7.02E-1
8.85E-1
2.00E-1
5.30E-1
4.46E-1
8.88
1.84
1.50
6.00E-1
1.43E+1
5.79E-2
Beef cattle
density
#/km2
1.02E+1
1.18E + 1
2.03E+I
2.68E + 1
4.56
9 = 63
2.50
8.87
2.32E+1
2.11E + 1
1.90E4-1
2.84
4.71
1.31E+1
5.62
6,23
1.81E+1
5.12
Vegetable
crop fraction
km^ /km^
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
LOSE -3
8.70E-3
5.20E-2
1.16E-3
1.78E-2
1.59E-3
A-16
-------
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
PS UBS
De script ion
Breathing Rate (cm^/h)
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~l)
Exposure period-pasture (h)
Exposure period-crops or leafy
vegetables (h)
Productivity-pasture (dry
weight) (kg/m2)
Productivity-crops and leafy
vegetables (kg/ra--)
Time fraction-pasture grazing
Pasture feed fraction-while
Value
9.17E+5
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
UK
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-17
-------
Table A~5. Site independent parameters used for AIRDOS-EPA
generic site assessments (Continued)
Symbolic
variable
Description
Value
FSUBG
FSUBL
TSUBB
P
TAUBEF
MSUBB
VS UBM
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~l) 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-18
-------
Table A~6« Element, dependent factors used in AIRDQS-EPA assessments
Element
Ac
Ac
Ag
Am
As
Ba
Be
Bi
Ce
Cm
Co
Co
Gr
Cr
Cs
Eu
Fe
Ga
La
Mn
Mo
Nb
Pa
Pb
Po
Po
Pr
Pu
P
Ra
learance
c lass
Y
W
Y
Y
W
D
W
W
Y
Y
W
Y
D
Y
D
Y
W
W
W
Y
D
W
Y
W
I)
D
W
Y
W
W
I)
Y
Y
D
W
Fl
0.10E-2
Q.1QE-2
G.5QE-1
0.10E-2
0.30E-1
0,10
0.20E-2
0.5QE-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
O.LOE-3
0.10E-3
0 . 10
0,95
0.95
0.10E-1
0.10E-2
G.30E-1
0 , 10
0.10
0.10E-3
0.30E-4
0,80
0.20
*m
(d/L)
2.0E-5
2.0E-5
3 . OE -2
3.6E-5
6 . 2E -5
3.5E-4
9.1E-7
5.QE-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.QE-5
9 . 7E -6
2.0E-6
9 . 9E -3
2,OE-5
2.0E-5
8.4E-5
1.4E-3
3.5E-2
2.0E-2
5.0E-6
8 , 7£ -5
1 . 2E-4
1.2E-4
2.QE-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
1,3.6-2
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 . OE -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
5.0E-4
Biv.
l.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
Biv,,
2.5E-3
2.5E-3
1 , 5E -1
1.5E-3
1.7E-2
2 , OE™ 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
2.0E-2
A~19
-------
fable 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
c lass
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
Q.20E-2
J?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, OS -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
1 . 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-20
-------
REFERENCES
At74 Athey T. W.s 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.
Ba81 Baes C. F, III and R. D. Sharp, A Directory 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.
Be81 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.
EPA80 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.
Ho72 Bolzworth 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.
Mo79 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.
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-21
-------
-------
APPENDIX B
R&DIONUCLIDB EMISSIONS TO AIR FROM FORMER
MANHATTAN ENGINKERING DISTRICT AND
ATOMIC ENERGY COMMISSION SITES
(FU5RAP)
-------
Page Intentionally Blank
-------
B: RADIONUCL1DE EMISSIONS TO AIR
MANHATTAN ENGINEERING DISTRICT AMD
ATOMIC ENERGY COMMISSION SITES
CFUSRAP)
CONTENTS
Page
B.I Background B-5
B.2 Current Status B-5
B.3 Potential for Airborne Releases B-13
B.4 Site Summaries B-13
B.5 Discussion B-16
TABLES
B-l Formerly Utilized Sites Remedial Action Program (FUSRAP)
Sites, March 1983 B-6
B-2 FUSRAP sites with legislative authority for remedial action B-ll
B-3 FUSRAP sites without legislative authoirty for remedial
action B-13
FIGURE
B-l Radon concentration near the tailings pile B-18
B-3
-------
Page Intentionally Blank
-------
Appendix B: RADIONUCLIDE EMISSIONS TO AIR
MANHATTAN ENGINEERING DISTRICT AMD
ATOMIC ENERGY COMMISSION SITKS
(FUSRAP)
The original program for the development and use of atomic energy,
established by the Army Corps of Engineers' Manhattan Engineering
District (MED) and continued by the Atomic Energy Commission (ABC), was
conducted under contract at Federally-, privately-, and institutionally-
owned sites. When the contract terminated, the sites were decontami-
nated according to the health and safety criteria then in effect and
were released for unrestricted use. Changing radiological criteria
prompted the ABC to re-examine the radiological status of these sites
in 1974, to determine if further remedial actions were required,
This re-examination was continued under the Energy Research and
Development Administration (ERDR) and the Department of Energy (DOE)
and was expanded to include radiological surveys of former HED/ASC
sites. When the results of several site surveys showed that remedial
actions would be necessary, the DOE initiated the Formerly Utilized
Sites Remedial Action Program (FUSRAP) to identify all former MED/ABC
sites and to resolve any site radiological problems. As of March 1983,
36 sites had been designated as FUSRAP sites. These sites, their
locations, and present owners are listed in Table B-l.
B.2 current_Status
Of the 36 FUSRAP sites, determinations that remedial actions are
required have been made for 22 sites and are pending for the remaining
14 sites. As shown in Table B-l, DOB has legal authority for carrying
out remedial actions under the provisions of the Atomic Energy Act of
1954, as amended, at only 14 of the 36 FUSRAP sites. The status of
remedial actions at these 14 sites is summarized in Table B-2. At 3 of
the 14 sites, no determination has been made that remedial actions are
required. At six sites,, some actions have been initiated. Remedial
actions have been completed at five sites,
At the 22 remaining sites (see Table B-3), the DOE's authority
extends only to characterizing the radiological status of the site,
determining the need for remedial action (completed for 11 sites), and
-------
fable B-l. Formerly Utilized Sites Remedial Action Program
(FUSRAP) Sites, March 1983
CO
Site/Location
* I. Acid/Pueblo Canyon
Los Alamos, NM
2, Albany Metallurgical Research Center
Albany, OR
* 3. Ashland Oil Co, (No. 1}
Tonawanda, NY
4, Ashland Oil Co.
-------
Table B-l. Formerly Utilized sites Remedial Action Program
(PUSRAP) Sites, March 1983 (Continued)
Site/Location
Ownership
Designated
for reaiedial
action
Authority for
remedial action
S3
*1Q. Gardiner, Inc.
Tampa, PL
11, w. R. Grace & Co.
Curtis Bay, MD
*12. Guterl Steel Corp.
Lockport, NY
13. Harshaw Chemical Co.
Cleveland, OH
14. Iowa State University
Ames, IA
*15, Kellex/Pierpoint Research Facility
Jersey City, NJ
*16, Linde Rir Products
Tonawanda, NY
17. Niagara Falls Storage Site
(Vicinity Properties)
Lewiston, NY
Gardiner, Inc. Yes
¥. R. Grace & Co. Pending
Guterl Special Steel Co., Yes
Simmons Steel Division
Harshaw Chemical Co. Pending
Iowa State University Pending
Municipality of ftmes
Delco-Levco, Yes
Pierpolnt Associates
Union carbide Corp., Yes
Linde Air Products Division
Town of Lewiston, Fort Yes
Conti Corp., S. Washuta,
Niagara Mohawk Power Co.,
the Somerset Group, Inc.,
U.S. Air Force, Services
Corporation of America
Mo
No
NO
MO
Yes
See footnote at end of table.
-------
Table B-l. Formerly Utilized Sites Remedial Action Program
(PUSRftP) Sites, March 1983 (Continued)
Site/Location
*18. Mallinckrodt, inc.
St. Louis, MO
*19. Middlesex Landfill
Middlesex, NJ
*20, Kiddlesex Sampling Plant,
fSMdlesex and Piscataway, NJ
21. Mont iceIlo (Vicinity Properties)
Mont ice Ho, UT
*22, National Guard Rrmory,
Chicago, IL
23. Olin Chemical Corp.
Joliet, IL
*24« Palos Park Forest Preserve
Cook County, IL
25, Pasadena Chemical Co,
Pasadena, TX
Ownership
Designated
for remedial
action
Mallinckrodt, inc. Yes
Borough of Middlesex, Yes
Middlesex Presbyterian
Church
U.S. Government, Yes
Multiple Private Ownership
Multiple Private Ownership Pending
State of Illinois Pending
Olin Mathieson Chemical Pending
Corporation
Cook County Forest Yes
Preserve District
Pasadena Chemical Co. Yes
Authority for
remedial action
No
Yes
Yes
Mo
NO
NO
No
*26. St. Louis Airport Storage Site
St. Louis, MO
St. Louis JVirport
Authority
Yes
No
See footnote at end of table,
-------
Table B-l. Formerly Utilized Sites Remedial Action Program
(FUSRAP) Sites, March 1983 (Continued)
Site/Location
Ownership
Designated
for remedial
action
Multiple Private Ownership Yes
Seaway Industrial Park Yes
Development Co., Inc.
U.S. Army Yes
Ms. L. Shpack Yes
Unknown Pending
Vulcan Cyclops, Inc. Pending
University of California Yes
University of Chicago Pending
Authority for
remedial action
ro
vO
27, St. Louis Airport Site
(Vicinity Properties)
St. Louis, MO
*28, Seaway Industrial Park
Tonawanda, NY
*29. Seneca Army Depot
RomuIus, NY
*30, Shpack Landfill
Norton, MA
31. Staten Island
Staten Island, NY
*32, Universal Cyclops, Inc.
Aliquippa, PA
33. University of California
Berkeley, CR
*34. University of Chicago
Chicago, IL
Yes
No
No
Yes
No
Yes
Yes
See footnote at end of table.
-------
CEJ
t-*
o
Table B-l. Formerly Utilized Sites Remedial Action program
(FUSRAP) Sites, March 1983 (Continued)
Site/Location
Ownership
Designated
for remedial
action
Authority for
remedial action
35, Ventron Corporation
Beverly, MR.
36. Watertown Arsenal
Watertown, MA
Thiokol Corporation Pending
Watertown Redevelopment Pending
Corporation
No
No
*Sites for which Radiological Survey Reports are publicly available; see References,
-------
Table B-2, FUSRAP sites with legislative authorityW for remedial action
Status of remedial action as of March 1983
Site name
and location
Determine
need and
authority
Prel Jminary
engineering
completed
Select
action
options
Design
engineering
initiated
NEPA
process
completed
Select
remedial
action
Design
engineering
completed
Remedial
action
completed
Acid/Pueblo Canyon,
Los Alamos, NH
Albany Metallurgical
Research Center,
Albany, OR
Bayo Canyon,
Los Alamos, NM
Chupadera Mesa, White Sands
Missile Range, NM
E.I. OuPont DeNemours
00 & Co., Oeepwater, N3
c- Kellex/Pierpoint Research
Facility, Jersey City, NJ
Linde Air Products,
Tonawanda, NY
Niagara Falls Storage Site
(Vicinity Properties)
(Formerly the Late Ontario
Ordnance Works)
(1) 19 acres of disposal
faciIi ty
(2) central and west ditches
(3) remaining 30 properties
See footnotes at end of table.
-------
Table B-2. FUSRAP sites with legislative authority^3' for remedial action (Continued)
OB
?
to
Site name
and location
Status of remedial actjgn_as of terch 1983 :
Determine Preliminary Select Design NEPA Select Design
need and engineering action engineering process remedial engineering
authority cosipletecl options initiated completed action completed
Middlesex Municipal X
Landfill, Middlesex, MJ
Middlesex Sampling Plant,
Hiddlesex and Piscataway, NO
(1) 33 Off-site properties X
(2) On-site X
St. Louis Airport
Storage Site,
(Vicinity Properties)
St. Louis, HO
Shpack Landfill,
Morton, HA
University of California,
Berkeley, CA
University of Chicago,
Chicago, IL
XV V V
A A A
(remedial action suspended pending site of NJ
action on selection of disposal site)
^Authorized by the Atonic Energy Act of 1954 and amendments.
Status Legend: X - Phase completed; P - Partially completed
Remedial
action
-------
Table B-3. FUSRAP sites without legislative authority^3)
for remedial action
Status of remedial action
as of March 1983
Site name and location Designated Preliminary
for remedial engineering
action completed
Ashland Oil Co. (No. 2), Tonawanda, NY
Ashland Oil Co. (No. 1), Tonawanda, NY
Clecon Metals Inc., Cleveland, OH
Conserv, Inc., Nichols, FL
Gardiner, Inc., Tampa, FL
Guterl Steel Corp.s Lockport, NY
Harshaw Chemical Co., Cleveland, OH
Iowa State University, Ames, IA
Mallinckrodt , Inc., St. Louis, MO
Monticello (Vicinity Properties), UT
National Guard Armory, Chicago, IL
01 in Chemical Co., Joliet, IL
Palos Park Forest Preserve s
Cook County, IL
Pasadena Chemical Company, TX
Seaway Industrial Park, Tonawanda, NY
Seneca Army Depot, Romulus, NY
St. Louis Airport Storage Site,
St. Louis, MO
Staten Island, NY
Universal Cyclops Inc., Aliquippa, PA
Ventron Corporation, Beverly, MA
W. R. Grace and Co., Curtis Bay, MD
Watertown Arsenal, Watertown, MA
P
X
X
X
X
X
P
P
X
P
P
P
X
X
X
X
X
P
P
P
P
P
X
P
X
P
X(b)
P
{^Radiological, surveys, determinations of need for remedial actions,
and planning for these sites were conducted under the authority
of the Atomic Energy Act of 1954, as amended. No legislative
authority exists for conducting remedial actions at these sites.
'"'Department of Array is responsible for remedial action. No further
action required under FUSRAP,
Status legend: X - Phase Completed, P- Partially Completed.
B-13
-------
planning. Completion of remedial actions at these sites will require
DOE to obtain additional legislative authority.
8• 3 Potentia 1 for Airborne Re leases
To assess the potential for airborne releases of radioactive
materials from FUSRAP sites, we have reviewed all of the Radiological
Survey Reports which are publicly available (see References). These
reports cover 22 of the 36 sites. For the sites where no Radiological
Survey Reports are available, we reviewed "A Background Report for the
Formerly Utilized Manhattan Engineering District/Atomic Energy
Commission Sites Program" (DOE/BV-0097), to determine the potential for
significant airborne releases, Although the information contained in
this document is mainly descriptive, it does not appear that any of
these sites has a greater potential for airborne release than the sites
for which Radiological Survey Reports are available?
Based on our review, eight representative sites were selected for
further study including the St. Louis storage site which appears to
have the greatest emissions of radionuclides to air. The other seven
sites were selected randomly, and indicate the range of potential
releases from FUSRAP sites. All of these sites have been designated
for remedial action.
B.4 Site Summaries
St. Louis Airport, St. Louis. MO
This 21.7 acre site adjacent to the St. Louis Airport was used to
store residues of contaminated scrap from the Mallinckrodt Chemical
Corporation's uranium-processing operation. Residues were stored in
the open, in steel drums, and in an open concrete pit. All residues
were removed from the site during 1966 and 1967.
The radiological survey of the site identified significant surface
and subsurface contamination both on- and off-site. Measurements of
external gamma radiation 1 meter above the surface ranged from near
background levels (10 yR/hr) to 330 yR/hr. The highest measurement
was off-site, and continuous exposure could result in an integrated
dose equivalent of approximately 2.9 rent/year. Radon flux measurements
averaged 6.3 pCi/m^-sec, equivalent to an annual Rn-222 source term
of approximately 17 Ci. On-site radon-222 measurements ranged from 30
to 130 fCi/L, and airborne concentrations of Ra-226, Th-230, Pb-210,
U-238, and Ac-227 near the west fence (the point of highest
concentrations), 14 fci/m3f 13 fci/in3, 30 fci/m3, 5 fCi/m3, and
1.6 fei/nr', respectively.
*This judgment does not apply to the Niagara Falls storage Site.
We are waiting for DOE Monitoring Reports on this facility, and will
update our findings as necessary when we have reviewed the data.
-------
The 10-acre site was used to dispose of 8,000 tons of residue from
ore processing operations at the Linde Air Products refinery. The
residue, containing approximately 0.54 percent uranium, was spread over
two-thirds of the site. In 1974, 6,000 cubic yards of residue were
removed to the adjacent Seaway Industrial Park (see below), and the
site developed for oil storage.
The radiological survey of the site identified extensive soil
contamination. External gamma radiation 1 meter above the surface on
the site ranges from 17 tiR/hour (or slightly above the 8-14 yR/hour
background level in the area) to 190 jjR/hour, and averaged 33
nR/hour over the entire site. Continuous exposure to the highest and
average measured gamma radiation would result in an integrated dose
equivalent of approximately 1.6 rem/year and 0.3 rem/year,
respectively. In addition, a radon flux of 7 pCi/m-^-sec was
estimated as the average for the entire 10-acre site. This would
result in an annual radon source terra of approximately 9 ci.
Bavo Canyon Area., Los Rlamps. MM
Bayo Canyon was used from 1943 through 1961 as an experimental
area for high explosives. Test assemblies of natural and depleted
uranium using lanthanum-140 as a tracer were exploded in the area,
dispersing approximately 1.3 curies of natural uranium, 1.2 curies of
depleted uranium, and between 30 and 40 curies of strontium-90 (present
as a contaminant of lanthanum-140). An additional 85 and 120 curies of
strontium-90 were deposited in waste handling facilities in the area
and some fraction migrated into the subsurface environment. The area
was decontaminated in 1963, and most of the debris was removed. The
area is currently used as a recreational area, although residential
development has been proposed.
The radiological survey of the site shows no statistically
significant difference between the airborne concentrations of Sr-90 or
uranium in the Bayo Canyon area compared with other northern New Mexico
locat ions.
Clecon Metals, Inc., Cleveland, OH
Two of the three buildings at this 3.5 acre industrial site were
used in the production of granular thorium metal for MED/ABC. The
contamination in these buildings was removed or covered due to
construction modifications after the thorium operations were ended.
The radiological survey of the site indicates that most of the
contamination is within the two buildings used for thorium processing.
However, some surface contamination is present. External gamma expo-
sure rates 1 meter from the surface average approximately 3 pR/hour
higher than the normal 10 yR/hour in the area. Continuous exposure
B-15
-------
would result in an Incremental dose above normal background of
approximately 26 rarera/year. The average concentration of thorium-232
in the soil is 6 pCi/g. Assuming 6 pCi/g fh-232, resulting in
6 PCi/m^-sec Rn~220f the annual source term for Rn-220 is estimated
to be approximately 2.7 ci,
LJ^^
The 700-acre Chambers Works site is adjacent to Deepwater, NJ,
MED/AEC operations involving uranium conversion were conducted at
three buildings at the site and a low-level radioactive burial ground
(licensed by New Jersey), Only one of the three buildings used for
MED/AEC work is still standing.
The radiological survey of the site identified some surface
contamination (primarily next to the remaining processing building),
but the primary cause for remedial action is contamination of the
building itself. The highest external gamma radiation level 1 meter
above the surface was 23 pR/hour, with most measurements in the range
of 3-6 pR/hour. Continuous exposure at the highest exposure rate
would result in an integrated dose equivalent of approximately
0.2 rera/year. Radon daughter concentrations in air ranged from 0.0001
to 0.0006 WL.
!JMl£sex_J^^
MED/ABC operations involving the sampling, weighing, assaying, and
storing of uranium and thorium ores were conducted at six buildings on
this 9.6 acre site.
The radiological survey of the Middlesex Sampling Plant, identified
extensive soil contamination and elevated external gamma exposure rates
and radon and radon daughter concentrations. External gamma readings
1 meter above the surface ranged from 22 pR/hour to 147 pR/hour.
The highest measurements were made both near the center of the site and
along the site boundary. Continuous exposure at the maximum rate could
result in an integrated dose of approximately 1.3 rein/year. The
average radon emanation rate for the site was 3.2 pci./m2-sec, or
approximately 4 ci/year,
Off-site measurements were also made at the facility, and indicate
widespread contamination beyond the site. External gamma levels as
high as 235 yR/hour were measured off-site, and radon daughter
concentrations ranged from 0.004 to 0,014 WL in off-site residences and
commercial buildings,
Seaway _ Industrial Pa^k,-iiiTon.awand_a,__MY
The site covers approximately 100 acres, of which 13 acres
adjacent to the Ashland Oil Company were used to receive approximately
6,000 cubic yards of uranium-processing residue.
B-16
-------
The radiological survey identified significant surface
contamination in the three areas where residues from the Ashland Oil
property were dumped. External gamma levels as high as 80 pR/hour
were measured; continuous exposure could result in an integrated dose
equivalent of approximately 0.7 rem/year. The radon emanation rate
from the site Is estimated at 5 pci/m^-sec, equivalent to an annual
release of about 8 Ci of Rn-222.
Seneca flrmy Depot, Romulus. WY
Eleven munitions bunkers at the facility were used to store
pitchblende ore. When MED/ftEC activities terminated, the bunkers
reverted to use as munitions bunkers.
A radiological survey of the site indicates that significant
contamination of the bunkers occurred. However, this contamination is
limited to the bunkers themselves and the soil immediately surrounding
the entrances. Most of the measurements of soil were at background
levels, and it does not appear that this facility has any significant
potential for airborne contamination or direct gamma irradiation
outside the bunkers.
B.5 Discussion
It is reasonable to assume that the most significant airborne
emission from a typical FUSRAP site during normal conditions (not
during decommissioning operations) is that of radon. To estimate radon
concentrations from the reported emission rates and site areas, we used
Figure B-l. This figure presents radon concentrations in pCi/liter as
a function of distance from the center of a tailings pile, for various
pile sizes and a fixed radon emission rate of 280 pCi/m^-sec. The
figure suggests that the radon concentration at the fencepost is rather
insensitive to pile size.
For the above eight sites, on-site radon concentrations in the
range of 0.1 to 0.2 pCi/liter are estimated, with fencepost concen-
trations in the range of 0.025 to 0.05 pel/liter. These radon con-
centrations translate into radon daughter levels of 0.001 to 0.002 WL
on-slte, and 2.5X10"4 to 5.0X1Q"4 WL at the fencepost. The levels
at the fencepost are in the range of 0.008 to 0.017 of the 10 CFR 20
MFC (3 pci/liter). The estimated lifetime risk of fatal cancer at
the fencepost to the nearby individuals is in the range of 3X10"4 to
6X10"4.
B-17
-------
20
80 ha
10-
<
-------
REFERENCES
1. "A Background Report for the Formerly Utilized Manhattan
Engineering District/Atomic Energy commission sites Program," U.S.
Department of Energy, September 1980.
2, "A Report to Congress to Accompany the Proposed Residual
Radioactive Material Control Act." Draft Report, U.S. Department
of Energy, ORNL/Sub-80/13829/1, June 1980.
3. "Formerly Utilized MED/AEG sites Remedial-Action Program,
Radiological Survey of Kent Chemical Laboratory, The University of
Chicago, Chicago, Illinois, September 7-13, 1977, Final Report,"
U.S. Department of Energy, DOE/EV-0005/25, May 1982.
4. "Formerly Utilized WED/ABC Sites Remedial-Action Program,
Radiological Survey of Ryerson Physical Laboratory, The'University
of Chicago, Chicago, Illinois, September 11-25, 1976, Final
Report," U.S. Department of Energy, DOE/EV-0005/23, May 1982.
5, "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of Site A, Palos Park Forest Preserve,
Chicago, Illinois, Final Report," U.S. Department of Energy,
DOE/EV-0005/7, April 1978.
6. "Formerly Utilized MBD/REC Sites Remedial-Action Program,
Radiological Survey of the fishland Oil Company, (Former Haist
Property), Tonawanda, Mew York, Final Report," U.S. Department of
Energy, DQE/EV-GGQ5/4, May 1978.
7. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Bayo Canyon, Los Alamos, Mew Mexico,
Final Report," U.S. Department of Energy, DOE/BV-0005/15, June
1979.
8. "Formerly Utilized MED/AEC Sites Remedial-fiction Program,
Radiological Survey of the Building Elite 421, United States
tfatertown Arsenal, Watertown, MA, Final Report," U.S. Department
of Energy, DOE/EV-OQ05/19, February 1980.
9. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the E.I. DuPont DeNeroours and Co.,
Deepwater, New Jersey, Final Report," U.S. Department of Energy,
DOE/EV-0005/8, December 1978.
10. "Formerly Utilized MED/REC Sites Remedial-Action Program,
Radiological Survey of the Former GSA 39th Street Warehouse,
1716 Pershing Road, Chicago. Illinois, Final Report," U.S.
Department, of Energy, DOE/EV-0005/9, January 1978,
3-19
-------
REFERENCES
(Continued)
11. "Formerly Utilized MED/AEC Sites Remedial-Retion Program,
Radiological Survey of the Former Horizons Inc., Metal Handling
Facility, Cleveland, Ohio, Final Report," U.S. Department of
Energy, DOE/EV-0005/10, February 1978.
12. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Former Linde Uranium Refinery,
Tonawanda, New York, Final Report," U.S. Department of Energy,
DOE/EV-OOQ5/5, May 1978.
13. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Former Simonds Saw and Steel Co.,
Lockport, Mew York, Final Report," U.S. Department of Energy,
DOE/EV-0005/17, November 1979.
14. "Formerly Utilized MED/ABC Sites Remedial-Action Program,
Radiological Survey of the Former Virginia-Carolina chemical
Corporation Uranium Recovery Pilot Plant, Nichols, Florida, Final
Report," U.S. Department of Energy, DOE/EV-0005/18, January 1980.
15. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the George Herbert Jones Chemical
Laboratory, The University of Chicago, Chicago, Illinois, June
13-17, 1977, Final Report," U.S. Department of Energy,
DQE/EV-Q005/26, May 1982.
16. "Formerly Utilized MED/AEC sites Remedial-Action Program,
Radiological Survey of the Middlesex Municipal Landfill,
Middlesex, New Jersey, Final Report," U.S. Department of Energy,
DOE/EV-0005/20, April 1980.
17. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Middlesex Sampling Plant, Middlesex,
New Jersey, Final Report," U.S. Department of Energy,
DOE/EV-OOQ5/1, November 1977.
18. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Museum of Science and Industry,
57th Street and Lake Shore Drive, Chicago, Illinois, Final
Report," U.S. Department of Energy, DOE/EV-0005/13, February 1979.
19. "Formerly Utilized MSD/AEC Sites Remedial-Action Program,
Radiological Survey of the Seaway Industrial Park, Tonawanda,
New York, Final Report," U.S. Department of Energy, DOE/EV-0005/6,
May 1978.
B-2Q
-------
REFERENCES
(Continued)
20. "Formerly Utilized MED/R.BC Sites Remedial-Act ion Program,
Radiological Survey of the Seneca Army Depot, Romulus, New York,
Final Report," U.S. Department of Energy, DOE/EV-0005/11, February
1979.
21. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Site of a Former Radioactive Liquid
Waste Treatment Plant (TA-45) and the Effluent Receiving Areas of
Acid, Pueblo, and Los Alamos Canyons, Los Alamos, New Mexico,
Final Report," U.S. Department of Energy, DOS/EV-0005/30, Hay 1981,
22. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the St. Louis Airport Storage Site,
St. Louis, Missouri, Final Report," U.S. Department of Energy,
DOE/EV-0005/16, September 1979.
23. "Formerly Utilized MED/AEC Sites Remedial-Action Program,
Radiological Survey of the Universal Cyclops Inc. Titusville
Plant, (Formerly Vulcan Crucible steel Company) Aliquippa,
Pennsylvania, May 2-8, 1978, Final Report," U.S. Department of
Energy, DOE/EV-0005/33, May 1982.
24. "Formerly Utilized KED/AEC Sites Remedial-Action Program,
Radiological Survey of the West Stands, New Chemistry Lab and
Annex, and Ricketts Laboratory, The University of Chicago,
Chicago, Illinois, August 31-September 2, 1977, Final Report,"
U.S. Department of Energy, DOE/EV-0005/34, May 1982.
25. "Radiological Survey of Properties in the Middlesex, New Jersey,
Area, Supplement, Final Report," U.S. Department of Energy,
DOE/EV-0005/1, March 1981.
26. "Radiological Survey of the Former Kellex Research Facility,
Jersey city, New Jersey, Final Report," U.S. Department of Energy,
DOE/EV-0005/29, February 1982.
27. "Radiological Survey of the Former Uranium Recovery Pilot and
Process Sites, Gardiner, Incorporated, Tampa, Florida, Final
Report," U.S. Department of Energy, DOS/EV-0005/21, March 1981,
28. "Radiological Survey of the Mallinckrodt Chemical Works,
St. Louis, Missouri, Final Report," U.S. Department of Energy,
DOE/EV-0005/27, December 1981.
29. "Radiological Survey of the Shpack Landfill, Norton,
Massachusetts, Final Report," U.S. Department of Energy,
DOE/EV-0005/31, December 1981.
B-21
-------
REFERENCES
(Continued)
30. "Draft Environmental Impact Statement for Standards for the Control
of Byproduct Materials from Uranium Ore Processing (40 CFR 192),"
U.S. Environmental Protection Agency, EPA 520/1-82-022, March 1983.
B-22
-------
APPENDIX C
RADON EMISSIONS FROM DEPARTMENT OF ENERGY AND
NUCLEAR REGULATORY COMMISSION LICENSED FACILITIES
-------
Page Intentionally Blank
-------
APPENDIX C: RADON EMISSIONS
OF AND NUCLF.AR REGULATORY
COMMISSION LICENSED FACILITIES
CONTENTS
Page
C.I DOE Facilities C-5
DOE References C~2Q
C.2 Nuclear Regulatory Commission Source Material Licensees C-21
NEC References C-29
TABLES
C-l ^2%adon concentrations (pCi/1) at the Niagara Falls
storage site C-ll
C-2 Surface area, volume, and content of the Weldon Spring
Raffinate Pits C-13
C-3 Radioactive wastes stored in Weldon Spring Quarry C-15
C-4 Radon concentrations at the Weldon Spring site, 1982 c-56
FIGURES
C-l Feed Materials Production Center Environmental Features c-7
C-2 Radioactive waste storage locations and fencepost radon
monitors C-9
C-3 Weldon spring Raffinate Pits area C-12
C-4 Weldon Spring Quarry site C-17
C-5 The Middlesex Sampling Plant site C-18
C-6 Fansteel Metals Plant site C-22
C-7 Molycorp Plant site C-24
C-8 Stepan Chemical plant site c~25
C-9 Diagram of the Kerr-McGee facility C-27
C-3
-------
Page Intentionally Blank
-------
Appendix C: RADON EMISSIONS FROM DEPARTMENT
OF ENERGY- AND NUCLEAR REGULATORY
COMMISSION-LICENSED FACILITIES
This report presents information on radon emissions from
Department of Energy (DOE) and Nuclear Regulatory Commission (NRG)
licensed facilities.
C.1 DOEFacilities
To determine which DOE sites have radon emissions, we reviewed
environmental monitoring, radiological survey, hazard characterization,
engineering evaluation, and environmental assessment reports prepared
for DOE facilities. Our review of these sources identified four sites
where uranium residues and wastes are stored or where previous
operations involving uranium and thorium resulted in significant
contamination of soils.* Releases of "^radon from these sites are
found to be large enough to cause radon concentrations at the site
boundaries that are detectable in the presence of the naturally
occurring radon,**
Identified as having potentially significant radon releases are
the following five sites: (1) Feed Materials Production Center (FMPC),
Fernald, OH; (2) Niagara Falls Storage Site (NFSS), Lewiston, NY;
(3) Weldon Spring Site (WSS), Weldon Spring, MO; (4) Middlesex Sampling
Plant (MSP), Middlesex, NJ; and (5) Monticello Uranium Mill Tailings
Pile (MUMT), Monticello, Utah. Brief descriptions of each of these
sites, the source of the radon emissions3 and the approximate amounts
of radon emissions are presented below.
The Feed MaEerialsProduct ion Center
The FMPC, near Pernald, OH, is a prime contractor site operated by
National Lead of Ohio (NLO) for the DOE, The FMPC produces purified
* Once our literature review was completed, we verified the
comprehensiveness of our findings during conversations with
cognizant DOE personnel. We believe that the sites covered in
this report are the only DOE facilities where radon emissions are
large enough to be of concern.
** The source term at the Weldon Spring Site also includes 220rac|oll
from the thorium wastes at the sites.
C-5
-------
uranium metal and components for use at other DOE facilities. Feed
materials include ore concentrates, recycled uranium from spent reactor
fuel, and various uranium compounds. Thorium can also be processed at
the site. Only minor amounts of radon are released from the production
operations conducted at the site. The primary source of radon
emissions at the FMPC is pitchblende residues stored in two concrete
storage tanks. As shown in Figure C-l, the storage tanks are located
on the western portion of the site, south of the chemical waste pits
and approximately 325 meters from the western site boundary.
The pitchblende residues were received from the Mallinckrodt
Uranium Refinery in St. Louis, MO, during the period that the
Mallinckrodt plant was operated for the Atomic Energy Commission
(AEC). Until June 1983, the residues were owned by AFRIMET (the U.S.
subsidiary of the Belgian firm that originally supplied the ores) and
stored under a lease storage agreement with the DOE. Upon expiration
of the agreement, AFRlMBT paid a reported fee of eight million dollars
and transferred ownership of the residues (and additional residues
stored at the NFSS, see below) to the United States (St83).
The residues are reported to have a radium concentration of
0.2 ppm, equivalent to about 200,000 pCi/g 226Ra. The 8,790 metric
tonnes of residue contain almost 1,760 curies of radium. Residues are
stored in two concrete tanks. Earthen berms have recently been erected
around the tanks to reduce gamma exposure. PMPC is awaiting the result
of an engineering analysis before placing earthen covers on top of the
concrete covers of the tanks (St83). The placement of earthern covers
on the tanks could result in lower radon emissions as well as reduced
gamma exposures .
No measurements of current emission rates of radon- 222 on these
tanks are available. However, data from the 1981 monitoring report
(NL082), show average radon concentrations at the site boundaries
ranging from 0.28 to 0.70 pci/l. These data are presented below. The
locations of the monitors are shown in Figure C-l.
concentrations in Air
at the FMPC Boundary, 1981
Location Number ____ KaQ3§ ___
of of Maximum Minimum Average
..... Monitor __ ___S§SEl§s ______ B£i/l ______ pCi/ 1 ___ pjgi/1
BS1 4 0.94 0.11 0.58
BS2 4 1.35 0.17 0.61
BS3 3 0.60 0.13 0.42
BS4 4 0.66 0.05 0.34
BS5 3 0.40 0.08 0.28
BS6 4 0.80 0.34 0.57
BS7 3 1.07 0.30 0.70
C-6
-------
Figure C-l
MATERIALS PRODUCTION CENTER
ENVIRONMENTAL FEATURES
BOUNDARY AIR SAMPLING STATIONS
SCAtE - !" = 1380'
C-7
-------
Data frost an off-site monitor located approximately 13 kilometers
east-northeast of the site showed an average "2^ concentration of
0.67 pci/l, while a single measurement at a location eight kilometers
west-northwest indicated 0.36 pel/I.
The Niagara Falls Storage site
The NFSS in Lewiston, NY, is a DOE surplus facility, operated by
Bechtel National, Inc. The 77-hectare site is part of the former Lake
Ontario Ordnance Works, and is used solely for the storage of uranium
and pitchblende residues. The residues are stored in six buildings
that were originally part of the facility's water treatment plant and
in a spoils pile north of Building 411 (see Figure c-2). The major
residues stored at the NFSS are summarized below:
Major Pitchblende Residues Stored
at the DDE-Niagara Palls Storage Site
Residue
I.D.
Storage
Location
Weight
Metric
Tonnes
Volume
m3
Surface
Area, m2
226Radiunt
Content
;-65
,-30
,-50
'-32
;ands
Bldg. 434
Bldg. 411
Bldgs. 413-414
Recarb. Pit
Bldg. 410
Spoil Pile, N
of Bldg, 411
3,530
7 , 460
1,700
130
2
7,470(b)
3,080
6,020
1,624
110-336
117
1.860U)
562
unknown
175 unknown
7,084(b) 37,373
200 ppb
-10 ppb
-10 ppb
unknown
unknown
~3 ppb
Source: Ba81
(a) these residues are partially covered by water.
(b) approximate weight and volume at time of emplacement; con-
taminated material includes ~11,340 m^ of overburden and
~35,QQQ m3 of contaminated soil. Unknown quantities of
wastes have been added to the pile during remedial actions at
the NFSS.
As noted, the residue storage buildings at the NFSS were originally
part of the facility's water treatment plant. The K-65 residues are
stored in Building 434 (see Figure C-2), which is the old water header
tank for the system.. The tank is a concrete silo, 50 meters tall. The
top loading port of the silo was capped and sealed during the fall of
-------
DOE-NFSS FENCELINE
o
Figure C~2
RADIOACTIVE WASTE
STORAGE LOCATIONS AND
FENCEPOST RADON MONITORS
,_E—K_s— n__)|— |_._|i—»—»,— «
i 34 i|
434
DOE-NIAGARA FALLS STORAGE SITE
LOCATION MATERIAL
BiDQ.410
411
411
414
434
MIDDLESEX SAND STORAGE
(flECARB I»IT> F-33
L-30 RESiOUi
L-50 •
K-65
If PlalcMt R
-------
1980. The other residue storage buildings are isolated from the 434
silo, on the southwest section of the site. Buildings 413-414 are also
water storage tanks, approximately 19 meters in diameter. Buildings
410, 411, and the recarbonation pit are located adjacent to the 413-414
storage tanks. The R-10 spoils pile is north of Building 411. The
spoils pile originally contained the R-10 wastes, but contaminated soil
and materials from on-slte and off-site cleanup activities have also
been placed on the pile.
Radon monitoring at the NFSS during 1980 and 1981 showed radon
concentration at the site boundary west of the R-10 spoils pile in
excess of the DOE standard of 3,0 pci/l. To reduce exposures, a new
fenceline was established 145 meters west of the former western
exclusion boundary and remedial actions were initiated to reduce radon
emissions from the site. Much of the cleanup, which is scheduled for
completion during 1985, centers on cutting and diking the R-1Q spoils
pile, Additional effort is being placed on sealing buildings and
cleaning up contaminated portions of the site (Ba84). The effective-
ness of these activities can be partially seen by comparing the 1981
and 1982 radon concentrations at the site boundary, Annual average
concentrations reported in the 1981 and 1982 Environmental Monitoring
Reports (Be82 and Be83a) are presented in Table c-1. Figure C-2 shows
the monitoring locations corresponding to the monitor ID'S given in the
table. Radon monitoring results for 1983 should be available by May of
1984, The 1983 data should confirm or deny the effectiveness of the
remedial actions that have been taken at the NFSS.
The We 1 don sjgr InejL _S 1 te
The WSS, near Weldon Spring, MO, is a DOE surplus facility
operated by Bechtel National, Inc. Like the NFSS, it is used for the
storage of uranium and thorium wastes. The site consists of two
separate properties: the 2l-hectare Raffinate Pits site; and the
3.6-hectare Quarry site, located about six kilometers southwest of the
Raffinate Pits area,
The Raffinate Pits area (see Figure c-3) is a remnant of the
tfeldon Spring chemical Plant. During the period that the chemical
plant was operated for the Atomic Energy Commission, the four raffinate
pits received residues and waste streams from the uranium and thorium
processes conducted at the facility. Pits one and two contain
neutralized raffinates from uranium refining operations and washed slag
residues from uranium metal production operations. Pits 3 and 4 contain
uranium wastes similar to those contained in Pits 1 and 2. in addition,
they contain thorium contaminated raffinate solids from processing
thorium recycle materials. During decontamination of the chemical
Plant, drummed wastes and contaminated rubble were disposed of in Pit 4.
The surface areas, volumes, and contents of the pits are summarized in
Table c-2. Surface water (varying in depth with the seasons) always
C-iO
-------
Table C-l, '^Radon concentrations (pci/1)
at the Niagara Palls storage site
Monitor
I.D.
01
02
03
04
05
06
07
08
09
10
11
12
13
14
15
Average
1981(a)
0.91
0.32
0.30
0.31
0.30
0.48
1.33
4.06
4.82
4.75
1.40
1.10
NR
MR
NR
Average
1982(b,c)
1.15
NR
0.60
0.71
0.62
0.65
1.02
2.32
2.97
1.93
0.89
0.83
0.88
0.68
0.76
Sources: Be82 and Be83a
(a) Measurements made by Mound Laboratory.
(b) Measurements made by Bechtel National, inc.
(c) According to Be82, the monitors used by
Bechtel average approximately 25 percent
higher than the monitors used by Mound.
C-ll
-------
ARMY FENCE LtNE
B'igure C-3
Weldon Spring Raffinate Pits Area
-------
Table e~2. surface area, volume, and content of the
Weldon Spring Raffinate Pits
Pit
1
2
3
4
TOTALS
Constructed
1958
1958
1959
1964
Surface
area
(acres)
1.2
1.2
8.4
15.0
25.8
Total pit
volume
(yd3)
18,500
18,500
166,700
444,400
648,100
Total
waste
volume
(yd3)
17,400
17,400
129,600
55,600
220,000
Percent
filled
94
94
78
12
Total
uranium
content
(kg)
9,100
9,100
91,000
27,600
136,800
Total
thorium
content
(kg)
...
—
500
63,600
64,100
Source: 8e83b
C-13
-------
covers the residues in Pits 3 and 4. Pits 1 and 2 are usually covered
by water as well, but evaporation during the months can leave
the residues exposed,
The Quarry site (see Figure C-4), located about six kilometers
southwest of the Raffinate Pits area, was initially used by the U.S.
Army to dispose of TNT-contaminated rubble from the Weldon Spring
Ordnance Works. The quarry was first used to dispose of radioactive
wastes in 1959, when the ABC deposited thorium residues in drums.
During 1963 and 1964, approximately 32,000 re3 of uranium- and
radium-contaminated building rubble, process equipment, and
contaminated soil generated during the demolition of the Destrehan
Street Feed Plant in St. Louis, were dumped in the quarry. In 1966,
additional drummed and uncontained thorium residues were deposited when
process equipment was removed from the Weldon Spring chemical Plant.
Additional TNT-contaminated stone and earth, disposed of later in 1966
by the Army, covers these thorium residues. The final deposits to the
quarry were made in 1968 and 1969, when the Array's decontamination of
the Chemical Plant generated approximately 4,600 ra3 of contaminated
equipment and rubble. Table C-3 summarizes the radioactive wastes
stored in the quarry.
Environmental monitoring in the vicinity of the two disposal areas
Includes a network of 15 radon monitors. Table C-4 summarizes the
results of the WSS radon monitoring network during the period December
1981 - September 1982, The sampling locations of the monitors at the
Raffinate Pits and the Quarry are shown in Figures C-3 and C-4,
respectively. The off-site monitors are located north of the Raffinate
Pits area. The results presented in Table C-4 are for total radon,
including background.
The Middlesex SamplingPlant
The MSP, Middlesex, MJ, was used by the Manhattan Engineering
District and the Atomic Energy Commission between 1S43 and 1967, for
sampling, weighing, assaying, and storing uranium and thorium ores.
After termination of operations at the site in 1967, it was decontami-
nated and released to the U.S. Marine Corp for use as a training
center. Radiological surveys of the site and nearby private properties
discovered widespread contamination from windblown materials and use of
material from the site as fill. The DOE took responsibility for the
site and its cleanup. The cleanup, which was completed in 1982,
consisted of recovering contaminated soils from off-site properties and
removing contaminated soil areas from the site. All materials were
consolidated in a storage pile on the southern portion of the site (see
Figure C~5).
The temporary storage pile at the site is approximately 91 meters
by 121 meters and 1,7 meters high. More than 31,000 metric tonnes of
contaminated soil are contained in the pile. The average radium
C-14
-------
Table C" 3. Radioactive wastes stored in Weldon Spring Quarry
Date
deposited
Volume
Radioactive
materials
(kg)
Comments
3,8 percent 1959
thorium residues
185 4,500 Drummed residues; volume
estimated; most of the
residues below quarry
water; principal source
of radioactivity:
radium- 228.
Destrehan Street 1963 -
Plant demolition 1964
rubble
50,000
Contaminated equipment,
building rubble, estimate
of uranium and thorium
content not available;
principal source of
radioactivity:
radium-226.
3 percent 1966
thorium residues
555 11,800 Drummed residues; volume
estimated; stored above
water level; principal
source of radioactivity:
radium 228.
tfeldon Spring
Chemical Plant
rubble
1968
1969
5,555
Contaminated equipment,
building rubble; uranium
and thorium content and
radioactivity not avail-
able; principal sources
of radioactivity:
radium-226 and
radium- 228.
TOTALS:
56,295
16,300
C-15
-------
Table C~4. Radon concentrations at the
Weldon Spring site, 1982
December 1981- April 1982-
Sampling March 1982 September 1982 Average
location^ (pci/l) (pCi/1) (pCi/1)
R-l
R-2
R- 3
R-4
R-5
R--6
R-7
R-8
R-9
R-10
R-ll
R- 1.2
R-l 3
R-l 4
R- 15
0.58
0.35
0.27
0.30
0.84
0.40
0.48
0.86
1.07
0.30
0,51
0.12
0.25
0,23
0.15
0,53
0.55
0.26
0.16
0.24
0,32
1.03
0.92
1.55
0,52
0.47
0,13
0.30
0.18
0.41
0,56
0.45
0.27
0.23
0.54
0.36
0.76
0.89
1.31
0.41
0.49
0.13
0.28
0.21
0.28
Source: Be83b
locations R--1 through R-5 are at the boundary of the
Raffinate Pits area, R-6 through R-ll are at the boundary of
the Quarry area, and R- 12 through R-15 are off- site, north of
the Raffinate Pits area.
C-16
-------
*"**""? Upper
I Gflte
TLD (T) WD
0 OO
Figure C-4
Weidon Spring Quarry site
-------
o
J-—I
CO
-------
concentration is estimated to be 79 pCi/g, so that there are about 2,5
curies of 226ra(j£um £n ^le pi_ie. xhe pile is covered with a hyplon
cover, which serves both to stabilize the pile and reduce the radon
£lux from the wastes. The radon flux from the pile, with cover
installed, is estimated to be 8«4 pCi/m^-sec (Fo79).
No monitoring data for the MSP were found. Given the proximity of
the waste pile to the site boundary, and the estimated radon flux from
the pile5 it is possible that radon concentrations exceed background at
the site boundaries. Based on calculated concentrations presented in
the environmental statement for inactive mill tailings sites (EPA82),
we estimate that boundary concentrations at the MSP coula be as high as
0.5 - 0.7 pCi/1.
The Monticello Uranium Mill Tailings Pile
The Monticello Uranium Mill Tailings Pile (MUMI) is located at
Monticello, Utah, and has been inactive since 1960. About 900,000 tons
of uranium mill tailings were impounded in four separate areas covering
about 40 acres total. The mill was purchased by the Federal Government
in 1948 and operated by the AEC to recover uranium from 1949 to January
1960 when it was permanently shutdown. The Government owns the
tailings site. In addition, some offsite contaminated properties at
Monticello are included under DOE's FUSRAP program (see Appendix B).
Uranium ore was processed by both acid and carbonate leaching and thus
the tailings exhibit properties of both of these processes (Ab83,
AEC63, AEC66, BFEC76).
The tailings were stabilized in 1961 by grading and leveling the
tailings and the dikes made of tailings. The tailings were then
covered with about one foot of pit run gravel and dirt, folloxved by one
foot of top soil which was seeded with local vegetation (AEC63).
Further demolition and decontamination activities were conducted in
1974 and 1975 to reduce radiation levels at the site and improve the
esthetic quality (BFEC76).
Radiation measurements at the site were primarily for external
gamma radiation. These levels were reduced by stabilization to a range
of 2 to 3 about background levels. Radon emission measurements ranged
from 175 to 675 pCi/m2-see for the 4 areas covered by tailings
(Ab83). EPA estimated the total radon emissions from the pile using
methods described in EPA83. It was assumed the ore processed at the
site averaged 0,2 percent uranium, which was typical for ore processing
during the 1950"s. Thus, the radium content is 560 pCi/g tailings.
For the 40-acre site with 2 feet of cover materials, the annual
emission of radon is about 2800 Ci. The cover material is not
effective in retaining radon, according to an analysis by Rogers (Ro81).
C-19
-------
DOE REFERENCES
Ab83 Abramiuk I. N., et al., Monticello Remedial Action Project
Site Analysis Report, Grand Junction Operations, GJ-10(83),
Draft, November 1983.
AEC63 U.S. Atomic Energy Commission, Erosion Control, Uranium Mill
Tailings Project, Monticello, Utah. Grand Junction Office,
December 20S 1963.
AEC66 U.S. Atomic Energy Commission, Supplement to the Report of the
Monticello Mill Tailings Erosion Control Project, Monticello,
Utah. Grand Junction Office, Supplement to RMO-3005,
April 20, 1966.
Ba81 Battelle Columbus Laboratories, A Comprehensive
Characterization and Hazard Assessment of the DDE-Niagara
Falls Storage Site, BMI-2074 REV., Columbus, OH, June 1981.
Ba84 Telephone conversation with Mr. J. Baublatz, U.S. DOE, Surplus
Facilities Management Office, January 1984.
Be82 Bechtel National, Inc., Niagara Falls Storage Site (NFSS):
Environmental Monitoring Report, Calendar Year 1981,
10-05-202-001, Oak Ridge, TN, May 1982.
Be83a Bechtel National, Inc., Niagara Falls Storage Site (NFSS):
Environmental Monitoring Report, Calendar Year 1982,
10-05-202-002, Oak Ridge, TN, May 1983.
Be83b Bechtel National, Inc., Weldon Spring Site (WSS):
Environmental Monitoring Report, Calendar Year 1982,
10-05-201-002, Oak Ridge, TN, June 1983.
BFEC76 Bendix Field Engineering Corporation, Uranium Ore Stockpile
Site Decontamination and Mill Site Foundation Removal,
Monticello, Utah. BFEC-1976-7, June 1976.
EPA82 U.S. Environmental Protection Agency, Final Environmental
Impact Statement for Remedial Action Standards for Inactive
Uranium Processing Sites (40 CFR 192), EPA 520/4-82-013, Vol.
1, October 1982.
EPA83 U.S. Environmental Protection Agency, Final Environmental
Impact Statement for Standards for the Control of Byproduct
Materials from Uranium Ore Processing (40 CFR 192), EPA
520/1-83-008-1, Vol. 1, September 1983.
C-20
-------
DOE REFERENCES (Continued)
Fo79 Ford, Bacon and Davis Utah, Inc., Environmental Analysis of
the Former Middlesex Sampling Plant and Associated Properties,
Middlesex, New Jersey, FBDU 230-005, Salt Lake City, UT, April
1979.
NL082
National Lead of Ohio, Inc., Feed Materials Production
Center: Environmental Monitoring Annual Report for 1981,
NLCO-1180, Cincinnati, OH, May 1981.
Ro81 Rogers V. C. and G. M. Sandquist, Long- Term Integrity of
Uranium Mill Tailings Covers, Report to the Nuclear Regulatory
Commission, RAE-21-1 (Rev. 1), August 1981,
St83 Telephone Conversation with Mr. S, Stief, Safety Division, Oak
Ridge National Laboratory, December 1983.
C-21
-------
C,2
Commission Source Material Licensees
Facilities that could have potentially significant radon emissions
are those which process material containing greater than 0.05 percent
by weight of uranium or thorium (source material). Such facilities are
required to be licensed by the Nuclear Regulatory Commission. The NRC
has licensed more than five hundred facilities to possess and use
source material. We relied on information provided by personnel in the
NRC's Material Licensing Branch to identify facilities with potentially
significant radon emissions. Listed below are the six facilities so
provided.
Facility
Fanstee1, Inc.
Muskogee, Oklahoma
Molycorp
York, Pennsylvania
Stepan Chemical
Maywood, New Jersey
vistron Corp.
Lima, Ohio
Licensed Amount of
Sour ce Ma t e r ia 1
30 MT 0.1% U
67 MT 0.22% Th
45 MT Natural Th
0.1 MT Natural U
9500 yd3 0.1% Th
8600 yd3 0.1% Th
15 MT UsOe plus
catalysts containing
50 MT U
Kerr-McGee
Rare Earths Facility
West Chicago, Illinois
Mallinckrodt Co.
St. Louis, Missouri
1400 MT Th02
20 MT UaOe
27,1 MT U in Natural and
Synthetic Ores
30 MT Th in Natural and
Synthetic Ores
The dockets for each of these facilities were examined. However,
only a limited amount of information on radon emissions from these
sites was found. Each is described below.
Fansteel, Inc.
The Fansteel Metals Plant is on a 110-acre site near Muskogee,
Oklahoma. Raw materials are processed to extract tantalum and
columbium, and the liquid residues containing uranium and thorium are
pumped to settling ponds where the solids settle out and the liquid is
processed and disposed of. The site layout is shown in Figure C-6.
There is no scale shown on Figure C-6; however, the proposed basic
C-22
-------
AUG 1 5 1883
LEGEND
L: Total Retention Lagoons
(listed in Appendix A)
P: Treatment Ponds
(listed in Appendix B)
W: Ground Water Monitoring Location
Figure C-6
C-23
-------
residue pond toward the bottom is 570 feet long (east to west) by 277
feet wide (north to south) measured from the inside top or the dikes,
The facility is licensed to possess no more than 30 MT of natural
uranium and 6? MT of natural thorium. We could not determine how much
material is actually on hand; however, the licensee requested approval
in March 1983 to construct the proposed pond shown on Figure O6
because pond 8 (to the northwest of the proposed pond) will be full
within two years .
The Molycorp plant at York, Pennsylvania (site layout shown in
P'igure C-7), operates a rare earth extraction process which produces
about 26 MT/month (dry weight) of residues containing 0.65 wt% thorium
and 0.002 wt% uranium. These residues are currently put into 55-galion,
plastic- bag- lined, steel drums pending future disposal. Apparently
there are plans to approximately double production. Current plans for
disposal of the residues call for them to be added to the tailings
being disposed of at a Nuclear Regulatory commission licensed tailings
impoundment at Sweetwater, Wyoming. A measurement at the south plant
boundary, near the vent scrubber, indicates a radon concentration o£
0.002-0.003 working levels,
In addition to these residues, which are apparently going to be
disposed of, there was reference to about 800 cubic yards of contami-
nated earth at the York plant, and a thorium slag waste pile at a
Washington, Pennsylvania, facility. We could not. obtain information
on these potential sources of radon, apart from one statement that the
radiation level at the surface of the contaminated earth at the York
plant was as high as 580 pR per hour (SR per year),
2 . 3 SbejEtart, Chemica 1
Stepan Chemical does not use source material; however, its plant
is on land formerly owned by the Maywood Chemical company, who between
1895 and 1959 operated a process which resulted in thorium waste.
Because there were no restrictions on disposal of such waste during
this period, it was simply put in piles at various places on the
Maywood Chemical Co. property, some of which was later sold. The site
and some of the surrounding property (including some residential
property) have been found to be contaminated with thoriura. The Nuclear
Regulatory commission is presently negotiating with Stepan chemical
regarding the steps to be taken to clean up the area. The plant site,
along with measured radiation levels in yR/hour, is shown in
Figure c-8. Apparently there are (according to the license) about 8600
cubic yards of 0.25% thorium residues buried in the area identified as
"Burial III" on Figure C-8. No information, was available regarding the
amounts in other (off-* site) areas. The Nuclear Regulatory Commission
has stated that the otf-site contamination does not pose any immediate
threat to public health and safety. The Nuclear Regulatory Comiission
C-24
-------
PLANT PLOT PLAN
TTt-zrv
to%»S «.^-
earth
chloride resi3ug 'storage
;r^:r™
WINDSOR 5T
Air samcie iocaiioris are m itctc 'voe
nurr.oe'S in -Iotic 'yes C'S gc^r-c s«."/ew Cd
MOLYBDCNUM
JQji? • _£?
Figure C-7
C-25
-------
o
N,Y. SUSQUEHANNA & WESTERN
I I I II I M I I I I I I 333
54 20 !"»
65
/*•
I
o
42'
/
c-'
HOV3B
-------
has noted that there is a potential for a few persons living in some of
the residences to receive radiation doses in excess of the accepted
limits for members of the public.
Vis t ron Corgora_t ion
Vistron at one time manufactured uranium-bearing catalysts but
does not do so any longer. As of 1976, 420 MT of catalysts containing
about 50 MT of uranium and about 15 MT of UaOa were stored in the
Vistron Plant, This material is stored in sealed drums in an abandoned
warehouse on the plant site. The measured radiation level at one foot
from the surface of these drums was 0.4 mrem/hr for the drums
containing catalysts and 1.1 mrem/hr for the drums containing UaOe,
Kerr-McGee
Liquid waste from the production of thorium and rare earth
elements was generated at the West Chicago site from 1932 to 1973. The
site layout is shown in Figure C-9. Plans for its decommissioning are
currently underway. The Nuclear Regulatory Commission, in its final
environmental statement related to the decommissioning (NUREG-Q9Q4, May
1983), recommended that all radioactive material be stabilized and
stored on-site for an indefinite period, with ultimate disposal to be
determined later. The amounts of ThOa and UaOa are as shown
be low:
Quantity, MT
Location ThOa UaOs
Ore residue pile 210 3
sediment pile (near 470 6
Building 18)
Ponds 1-3 760 12
The Nuclear Regulatory Commission has estimated current releases to
be 70 Ci/yr of radon-222 and 14,000 Ci/yr of radon-220, and doses of
<1 Kirem/yr to the whole body. 4 mrem/yr to the bone, 25 mrem/yr to the
lung, and 260 mrem/yr to the bronchial epithelium of the nearest
resident. With the recommended action, these doses would be reduced to
zero.
MallinckFodt Company
The Mallinckrodt Company's columbium-tantalum processing facility in
St. Louis, Mo., is licensed to possess 27.1 MT of uranium and 30 MT of
thorium in natural and synthetic ores. The docket for this facility
(40-6563) does not contain information on the layout of the facility or
the location of the uranium and thorium ore storage areas at the site.
Nor does the docket contain data on radon emissions or boundary
C-27
-------
STREET
ZOO
j Residue ('B
/ Pile
\Jj>
FIGURE C-9. DIAGRAM OF THE KERR-McGEE FACILITY
Adapted from ER. Figure 2,5,1.
C-28
-------
concentrations. Our attempts to obtain such Information from the
(both from headquarters and the cognizant regional office) were
unsuccessful. Estimates of the actual amounts of material stored at
the site or of the radon emissions from the materials are not available,
029
-------
NRG REFERENCES
The following documents were used in preparing the NRC section of
this report and are available for inspection in the NRC Public Document
Room under the appropriate docket number.
1- Fansteel, Inc. (Docket 40-7580)
USNRC, Draft Safety Evaluation Report Related to New Waste
Treatment Pond No. 9, 1983.
2. Holycorp (Docke. t 40-8794)
Application for source material license, 1981.
Eberline Instrument Corp. report of Radiation survey of Molycorp
Plant at York, Pennsylvania, 1981.
Molycorp Response to NRC Notice of violation, 1981.
3. Stepan ChemicaI (Docket 40-6610)
NRC Report on Thorium Contamination in the Area of Maywood and
Rochelle Park, New Jersey, 1981.
4- VistronCorp. (Docket40-7604)
Letter from R.C. Shower (Vistron) to J.M. Bell {NRC), February 24,
1976.
Letter from O.K. DOSS (Vistron} to K.S. Dragonette (NRC),
January 13, 1976.
5. Kerr McGee (.Docket 40-2061)
USNRC, Final Environmental Statement Related to the Decommissioning
of the Rare Earths Facility, West Chicago, Illinois, NUREG-0904,
May 1983.
6. Mallinckrodt Company (Docket 40-6563)
e-3o
-------
APPENDIX D
-------
Page Intentionally Blank
-------
APPENDIX D
DEPARTMENT OF ENERGY GOCO FACILITIES*
(Governtaent-OwBed, Contractor-Operated Facilities) where contractors
are subject to DOE Procurement Regulation 9-50.704™2(a)
Responsible
Field Office
California
1. a. Lawrence Berkeley Laboratory SAN
University of California
Berkeley, California
b. Dormer Laboratory SAN
University of California
Berkeley, California
c. Chemical Biodynamics Laboratory SAN
University of California
Berkeley, California
d, Dymo Facility (Building 934) SAN
University of California
Berkeley, California
Principal Contractor:
University of California
Berkeley, California 94720
2, a. Lawrence Livermore Laboratory SAN
University of California
End of East Avenue
Livermore, California
b, Lawrence Livermore Laboratory - Site 300 SAN
17 miles east of Livermore on Corral Hollow Road
Livermore, California
Principal Contractor:
University of California
P.O. Box 808
Livermore, California 94550
See key to abbreviations on page D-21.
D-3
-------
Responsible
Field Office
California (continued)
3. Sandia Laboratories, Livemore AL
End of East Avenue
Livemore, California
Principal Contractor:
Western Electric, Inc.
Livermore, California 94550
4, 130 Robin Hill Road NV
Goleta, California
Principal Contractor:
EG&G, Inc.
130 Robin Hill Road
Goleta, California 93017
5. a. Energy Technology Engineering Center SAN
DOE Triangle at Santa Susana
Canoga Park, California
b. Energy Technology Engineering Center SAN
Two DOE-owned buildings, total about
5,000 square feet, outside DOE triangle
Canoga Park, California
Principal Contractor:
Rockwell International
Atomics International Division
P.O. Box 1449
Canoga Park, California 91304
6. Stanford Linear Accelerator Center SAN
2572 San Hill Road
Menlo Park, California
Principal Contractor:
Stanford University
P.O. Box 4349
Stanford, California 94305
7. 2801 Old Crow Canyon Road NV
San Ramon, California
Principal Contractor:
EG&G, Inc.
P.O. Box 204
San Ramon, California 94583
D-4
-------
Responsible
gield^Office
California (continued)
8. Research and Development building, Project No. 37 OR
2525 West 190th Street
Torrance, California
Principal Contractor:
AiResearch Manufacturing Company
A Division of Garrett Corporation
2525 West 190th Street
Torrance, California 90509
Colorado
1. Rocky Plats Plant AL
25 miles northwest of Denver - Highway 93
Between Boulder and Golden, Colorado
Principal Contractor:
Rockwell International
Atomics International Division
P.O. Box 464
Golden, Colorado 80401
2. Solar Energy Research Institute CH
Contract No. EG-77-C-01-4042
Golden, Colorado 80401
Principal Contractor:
Solar Energy Research Institute
1617 Cole Boulevard
Golden, Colorado 80401
3. DOE Compound GJ
Grand Junction, Colorado
Principal Contractor:
Bendix Field Engineering Corporation
P.O. Box 1569
Grand Junction, Colorado
Connecticut
1. Knolls Atomic Power Laboratory SNR
Windsor Site
Windsor, Connecticut
D-5
-------
Responsible
Field Office
Connecticut (continued)
Principal Contractor:
General Electric Company
P.O. Box 545
Witidsors Connecticut 06095
Florida
1. Pinellas Plant AL
5 miles southeast of Largo on Bryan Dairy
and Belcher Roads
St. Petersburg, Florida
Principal Contractor:
General Electric Company
P.O. Box 11508
St. Petersburg, Florida 33733
2. Sandia Laboratories Mobile and Remote Range Facility AL
Building 1690
Cape Canaveral, Florida 32920
Principal Contractor:
Western Electric, Inc.
P.O. Box 5800
Albuquerque, New Mexico 87115
Hawaii
1. Sandia Laboratories AL
Barking Sands, Kauai, Hawaii
Principal Contractor:
Western Electric, Inc.
P.O. Box 478
Waimeas Kauai, Hawaii 96796
2. Communications and Scientific Station AL
Haleakalas Mauis Hawaii
Principal Contractor
Western Electric, Inc.
Pacific Area Support Office
P.O. Box. 9186
Haleakala, Maui, Hawaii
D~6
-------
Responsible
Field Office
Idaho
I. Idaho National Engineering Laboratory ID
40 miles west of Idaho Falls, on U.S. Highway 20
Principal Contractors:
EG&G Idaho, Inc. ID
Argorme National Laboratory CH
Exxon Nuclear Idaho Company, Inc. ID
Westinghouse Electric Corporation PNR
Resident Construction Contractor:
Morrisori-Knudsen Company, Inc. ID
Project Construction Contractors,"
Jones-Boecon (J-B) ID
Catalytic, Inc. IB
2. Idaho Falls, DOE, Office Building ID
550 Second Street
Idaho Falls, Idaho 83401
3. Contractor Operated Facilities
a. Computer Science Center ID
1155 Foote Drive
Idaho Falls, Idaho 83401
b. Computer Science Technical Support building ID
1520 Sawtell
Idaho Falls, Idaho 83401
c. Technical Support Building Addition ID
1580 Sawtell
Idaho Falls, Idaho 83401
d. First Street Building ID
550 First Street
Idaho Falls, Idaho 83401
e. Idaho Falls Warehouse Building ID
3600 Bombardier Boulevard
Idaho Falls, Idaho 83401
f. Idaho Falls Library Building Basement ID
457 Broadway
Idaho Falls, Idaho 83401
D-7
-------
Responsible
Field Office
Idaho (continued)
g. Idaho Geothermal - Raft River Project ID
Cassia County - approximately 50 miles
southeast of Burley off U.S. 30 on
approximately 5,000 acres of National
Resource Land and other lands within
the boundaries of DOE application for
withdrawal filed with the BLM and
assigned Serial Register No. I - 7435
Principal Contractor:
EG&G Idaho, Inc.
1955 Fremont
Idaho Falls, Idaho 83401
4. Willow Creek Office Building ID
1955 Fremont
Idaho Falls, Idaho 83401
Principal Contractors:
EG&G Idaho, Inc.
Exxon Nuclear Idaho Company, Inc.
Morrison-Knudsen Company, Inc.
Catalytic, Inc.
rilinois
1. Argonne National Laboratory CH
9700 South Cass Avenue
Argonne, Illinois 60439
Principal Contractor:
Argonne Universities Association
P.O. Box 307
Argonne, Illinois 60439
2. Fermi National Accelerator Laboratory CH
Off Kirk Road on West Boundary
Batavia, Illinois 60510
Principal Contractor:
University Research Associates, Inc.
2101 Constitution Avenue
Washington, D.C. 20037
D-8
-------
Field Office
Iowa
1. Ames Laboratory CH
a. Reactor Building - Scholl Road
b. Physics Addition Building
c. Laboratory and Office Building
d. Spedding Hall - Spammell Drive
e. Metallurgy Building - Spammell Drive
f. Metals Development - Spammell Drive
g. Warehouse Building - Maintenance Area
h. Mechanical Maintenance - Maintenance Area
i. Painting and Air Conditioning Shop - Maintenance Area
Principal Contractor:
Iowa State University
Ames, Iowa 50011
Kentucky
1. Paducah Gaseous Diffusion Plant OR
Paducah, Kentucky
Principal Contractor:
Union Carbide Corporation
P.O. Box 1410
Paducah, Kentucky 42001
Maryland^
1. WE Headquarters Building HQ
Germantown, Maryland
Principal Contractor;
Calculon Corporation
c/o U.S. Department of Energy
Washington, B.C. 20545
Massachuse 11 s
1. Bates Linear Accelerator CH
Middletott, Massachusetts
Principal Contractor:
Massachusetts Institute of Technology
77 Massachusetts Avenue
Cambridge, Massachusetts 02139
D~9
-------
Field Office
Missouri
1. Kansas City Plant AL
Bannister Road and Troost
Kansas City, Missouri
Principal Contractor:
The Bendix Corporation
P.O. Box 1159
Kansas City, Missouri 64141
2. Weldon Springs Retention Basin and Quarry OR
Off U.S. Highway 70 West
Weldon Springs, Missouri
Principal Contractor:
National Lead Company of Ohio
P.O. Box 39158
Cincinnati, Ohio 45329
Montana
1. Magnetohydrodynatnic, Component Development and ID
Integration Facility
53.16 acres near the Butte Industrial Park,
approximately 5 miles south of Butte, Montana
Principal Contractor:
Kaiser Engineers (Construction)
Montana Energy Research and Development
Institute (Operations)
MHD Site Office, P.O. Box 3562
Butte, Montana 59701
Nevada
1. 2753 South Highland Avenue NV
Las Vegas, Nevada
Principal Contractors:
Holmes & Narver, Inc.
2753 South Highland Avenue
Las Vegas, Nevada 89114
Wackenhut Services, Inc.
2753 South Highland Avenue
Las Vegas, Nevada 89114
D-10
-------
Nevada (continued)
Computer Sciences Corp.
2753 South Highland Avenue
Las Vegas, Nevada 89114
2. Nevada Test Site NV
Mercury, Nevada
Principal Contractors:
Reynolds Electrical & Engineering Co., Inc.
P.O. Box 14400
Las ?egas, Nevada 89114
Westinghouse Electric Corporation/Advanced
Energy Systems Division
P.O. Box 327
Mercury, Nevada 89023
3. Tonopah Test Range AL
47 miles southeast of Tonopah
Tonopah, Nevada
Principal Contractor:
Western Electric, Inc.
P.O. Box 871
Tonopah, Nevada 89049
4. a. 680 East Sunset Road NV
Las Vegas, Nevada
b. 6367 Escondido Road NV
Las ¥egass Nevada
Principal Contractor:
EG&G, Inc.
P.O. Box 1921
Las Vegas, Nevada 89101
5, a. 25 Wyandotte Street NV
b. Las Vegas, Nevada
b. 2300 West Rancho Drive, Suite 216 NV
Las Vegas, Nevada
c. 3084 South Highland Drive NV
Building 65 7, and 8
Las Vegas, Nevada
D-U
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Responsible
Field Office.
Nevada (Continued)
Principal Contractor:
Reynolds Electrical & Engineering Co., Inc.
P.O. Box 14400
Las Vegas, Nevada 89114
6. North Las Vegas Facility NV
316 East Atlas Circle
North Las Vegas, Nevada
Principal Contractor:
EG&G, Inc.
P.O. Box 1921
Las Vegas, Nevada 89101
New Jersey
1. Princeton Plasma Physics Laboratory . CH
"C" Site and "A" Site on the Forrestal Campus
Princeton, New Jersey
Principal Contractor:
Princeton University
P.O. Box 682
Princeton, New Jersey 08540
2. Burns & Roe Services Corporation CH
Contract No. DE-AC02-79ET14850
Oradell, New Jersey 07649
Principal Contractor:
Burns & Roe Services Corporation
496 Kinderkamack Road
Oradell, New Jersey 07649
NewMexico
1. Sandia Laboratories, Albuquerque AL
Kirtland Air Force Base - East
Albuquerque, New Mexico
Principal Contractor:
Western Electric, Inc.
P.O. Box 5800
Albuquerque, New Mexico 87115
D-12
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Responsible
Field Office
New Mexico (Continued)
2, Sandia Laboratories Mobile and Remote AL
Range Facility
Building 1137-1
White Sands Missile Range, New Mexico 88002
Principal Contractor:
Western Electric, Inc.
P.O. Box 5800
Albuquerque, New Mexico 87115
3. Inhalation Toxicology Research Institute AL
Kirtland Air Force base - East
Albuquerque, New Mexico
Principal Contractor:
Lovelace Medical Foundation
Building 9200, Area Y
KAFB - East
Albuquerque, New Mexico 87115
4, EG&G Operations NV
Kirtland Air Force Base - West
NC-135 Area
Albuquerque, New Mexico
Principal Contractor:
EG&G, Inc.
c/o Nevada Site Manager
KAFB - West
P.O. Box 4339
Albuquerque, New Mexico 87106
5. Los Alamos Scientific Laboratory AL
Los Alamos, New Mexico
Principal Contractor:
University of California
P.O. Box 1663
Los Alamos, New Mexico 87544
6. 1100 4th Street NV
Los Alamos, New Mexico
Principal Contractor:
EG&G, Inc.
P.O. Box 809
Los Alamos, New Mexico 87544
D-13
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Responsible
Field Office
7. 901 Trinity Drive AL
Los Alamos, New Mexico
Principal Contractor:
The Zia Company
901 Trinity Drive
Los Alamos, New Mexico 87544
8. Waste Isolation Pilot plant AL
32 miles SE of Carlsbad
Principal Contractor:
Western Electric, Inc.
1502 West Stevens Street
Carlsbad, New Mexico 88220
9. Fenton Hill Geothermal Site - TA-57 AL
45 miles west of Los Alamos
Principal Contractor
University of California
P.O.. Box 1663
Los Alamos, New Mexico 87544
10. Ross Aviation AL
Albuquerque Sun Port
Albuquerque, New Mexico
Principal Contractor:
Ross Aviation, Inc.
P.O. Box 9124
Albuquerque, New Mexico 87119
New York
1. Brookhaven National Laboratory CH
Off William Floyd Parkway
Upton, New York
Principal Contractor:
Associated Universities, Inc.
Upton, New York 11973
2. Knolls Atomic Power Laboratory SNR
River Road
Niskayuna, New York
Principal Contractor:
General Electric Company
P.O. Box 1072
Schenectady, New York 12301
D-L4
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Responsible
Field Office
New York (continued)
3. Knolls Atomic Power Laboratory SNR
Kesselriog Site
West Miltonj New York
Principal Contractor:
General Electric Company
P.O. Box 1072
Schenectady, New York 12301
4, Niagara Falls Boron Plant OR
Model City, New York
Principal Contractor;
National Lead Company of Ohio
P.O. Box 39158
Cincinnati, Ohio 45329
Ohj.p_
1. Portsmouth Gaseous Diffusion Plant OR
Off Highway U.S. 23
Piketon, Ohio
Principal Contractor:
Goodyear Atomic Corporation
P.O. Box 628
Piketon, Ohio 45661
2. Mound Facility AL
Miamisburg, Ohio
Principal Contractor:
Monsanto Research Corporation
P.O. Box 32
Miaraisburg, Ohio 45342
3. Feed Materials Production Center OR
6 miles north of Cincinnati - off Highway
U.S. 50 bypass west
Fernald, Ohio
Principal Contractor:
National Lead Company of Ohio
P.O. box 39158
Cincinnati, Ohio 45239
D-15
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Responsible
Field Office
Pennsylvania
1. Bettis Atomic Power Laboratory PNR
West Mifflin, Pennsylvania
Principal Contractor:
Westinghouse Electric Corporation
P.O. Box 79
West Mifflin, Pennsylvania 15122
2. Pittsburgh Energy Technology Center CH
4800 Forbes Avenue
Pittsburgh, Pennsylvania 15213
Principal Contractor:
General Electric Company
MATSCO
P.O. Box 7507
Philadelphia, Pennsylvania 19101
3. Shippingport Nuclear Power Station PNR
Shippingport, Pennsylvania
Principal Contractor:
Duquesne Light Company
P.O. Box 57
Shippingport, Pennsylvania 15077
South Carolina
1. Savannah River Plant SR
18 miles south of Aiken on State Route 125
Aiken, South Carolina
Principal Contractors:
E.I. du Pont de Nemours and Company
Aiken, South Carolina 29801
University of Georgia
Drawer E
Aiken, South Carolina 29801
Tennessee
1. Oak Ridge National Laboratory OR
Bethel Valley Road - About 12 miles
from Oak Ridge
Oak Ridge> Tennessee
D-16
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Responsible
Field Office
Tennessee (continued)
1. Oak Ridge National Laboratory (continued)
Principal Contractor:
Union Carbide Corporation
P.O. Box X
Oak Ridge, Tennessee 37830
2. Y-12 Plant OR
Bear Creek Road - About 1.5 miles
from Oak Ridge
Oak Ridges Tennessee
Principal Contractor:
Union Carbide Corporation
P.O. Box Y
Oak Ridge, Tennessee 37830
3. Oak Ridge Gaseous Diffusion Plant OR
Oak Ridge Turnpike - About 8 miles
from Oak Ridge
Oak Ridge, Tennessee
Principal Contractor:
Union Carbide Corporation
P.O. Box P
Oak Ridge, Tennessee 37830
4. Comparative Animal Research Laboratory OR
1299 Bethel Valley Road
Oak Ridge, Tennessee
Principal Contractor:
University of Tennessee
P.O. Box 1071
Knoxvilles Tennessee 37901
5. a. New Museum OR
Tulane Avenue
Oak Ridge, Tennessee
b. Medical Division Complex OR
Vance Road
Oak Ridge, Tennessee
c. Atmospheric Turbulence and Diffusion Laboratory OR
South Illinois
Oak Ridge, Tennessee
D-17
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Responsible
Field Jgffice
Tennessee (continued)
Oak Ridge National Laboratory (continued)
d. Warehouse Bays 4, 5 and part of 3 of OR
Building 1918-T2
Warehouse Road
Oak Ridge, Tennessee
e. Special Training Division OR
Building 2714 (F, G, and Annex) and 2715
Laboratory Road
Oak Ridge, Tennessee
Principal Contractor:
Oak Ridge Associated Universities
P.O. Box 117
Oak Ridge, Tennessee 37830
6. a. Water Treatment Facilities OR
Oak Ridge, Tennessee
b. Building 1916-T2
Warehouse Road
Oak Ridge, Tennessee
Principal Contractor:
The Rust Engineering Company
P.O. Box 587
Oak Ridge, Tennessee 37830
7. Charlotte Hall and Cheyenne Hall OR
Oak Ridge Turnnpike
Oak Ridge, Tennessee 37830
Principal Contractor:
Union Carbide Corporation
P.O. Box Y
Oak Ridge, Tennessee 37830
Texas
1. Pantex Plant AL
21 miles northeast of Amarillo, 2 miles north
of U.S. Highway 60
Amarillo, Texas
Principal Contractor:
Mason & Hanger - Silas Mason Co., Inc.
P.O. Box 647
Amarillo, Texas 79177
D-18
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Responsible
Field Office
Washington
1. Solvent Refined Coal Pilot Plant OR
Fort Lewis, Washington 98433
2. Hanford Project RL
5 miles north of Richland Federal Building
Riehland, Washington
Principal Contractors:
Rockwell International
Rockwell Hanford Operations
P.O. Box 250
Richland, Washington 99352
Battelle-Pacific Northwest Laboratory
P.O. Box 999
Richland, Washington 99352
&CS Richland, Inc.
P.O. Box 300
Richland, Washington 99352
Hanford Environmental Health Foundation
P.O. Box 100
Richland, Washington 99352
J.A. Jones Construction Company
801 First Street
Richland, Washington 99352
United Nuclear Industries, Inc.
P.O. Box 490
Richland, Washington 99352
Vitro Engineering Corporation
P.O. Box 296
Richland, Washington 99352
Westinghouse Hanford Company
P.O. box 1970
Richland, Washington 99352
3. 700 Area RL
Richland Federal building
825 Jadwin Avenue
Richland, Washington
D-19
-------
Responsible
Field Office
Washington (Continued)
3. 700 Area (Continued)
Principal Contractors:
Rockwell International
Rockwell Hanford Operations
Battelle-Pacific Northwest Laboratory
BCS Richland, Inc.
Hanford Environmental Health Foundation
United Nuclear Industries, Inc.
Vitro Engineering Corporation
Westinghouse Hanford Company
4. 703 Building RL
Knight Street
Richland, Washington
Principal Contractors:
Rockwell International
Rockwell Hanford Operations
Battelle-Pacific Northwest Laboratory
BCS Richlands Inc.
Hanford Environmental Health Foundation
5. a. 712 Building RL
Northgate Drive
Richland, Washington
b. 1100 Area RL
Stevens Drive
Richlandj Washington
c. Columbia Bank Building RL
1100 Jadwin Avenue
Richland, Washington
D-20
-------
Responsible
Field Office
Washington (continued)
d, Tannadore Building RL
1155 Jadwin Avenue
Richlandj Washington
e. Richland Sky Park RL
Terminal Building
Richland Airport
Richland, Washington
Principal Contractor:
Rockwell International
Rockwell Hanford Operations
P.O. Box 250
Richland, Washington 99352
6. 747 Building RL
Knight Street
Richland, Washington
Principal Contractors:
Hanford Environmental Health Foundation
BattHe-Pacific Northwest Laboratory
7. a. 748 Building RL
Swift Street
Richland, Washington
b. Medical-Dental Building RL
Swift Street
Richland, Washington
Principal Contractor:
Hanford Environmental Health Foundation
P.O. Box 100
Richland, Washington 99352
8. 3000 Area RL
First Street
Richland, Washington
Principal Contractor:
J, A. Jones Construction Company
801 First Street
Richland, Washington 99352
D-21
-------
Responsible
Field Office
Washington (continued)
9. a. Port of Benton Building RL
2592 George Washington Way
Richland, Washington
b. Hanford Square 1 Building RL
3080 George Washington Way
Richland, Washington
c. Group V Building RL
3200 George Washington Way
Richland, Washington
d. GESA Building RL
723 Parkway
Richland, Washington
e. Robert Young Building RL
1933 Jadwin Avenue
Richland, Washington
f. Robert Young Building RL
1955 Jadwin Avenue
Richland, Washington
g. Hanford Square 4 Building RL
3060 George Washington Way
Richland, Washington
h. Sigma III Building RL
316 George Washington Way
Richland, Washington
i. Sigma IV Building RL
3170 George Washington Way
Richland, Washington
Principal Contractor:
Battelle-Pacific Northwest Laboratory
P.O. Box 999
Richland, Washington 99352
10. Robert Young Building RL
1933 Jadwin Avenue
Richland, Washington
D-22
-------
Responsible
Field Office
Washington (continued)
Principal Contractor:
Vitro Engineering Corporation
P.O. Box 296
Richland, Washington 99352
11. a. Jadwin Building
1135 Jadwin Avenue
Richland, Washington
b. 3190 Building
3190 George Washington Way
Richland, Washington
c. 3180 Building
3180 George Washington Way
Richland, Washington
Principal Contractor:
Westinghouse Hanford Company
P.O. Box 1970
Richland, Washington 99352
Puerto Rico
1. a. Nuclear Research and Training Center
Rio Piedras, Puerto Rico
b. Nuclear Research and Training Center
Mayaguez, Puerto Rico
c. El Verde Terrestrial Ecology Station
Loquillo National Forest
Puerto Rico
RL
RL
RL
OR
OR
OR
Abbreviations:
AL - Albuquerque
CH - Chicago
OR - Oak Ridge
RL - Richland
SAN - San Francisco
NV - Nevada
GJ - Grand Junction
SNR - Schenectady Naval Reactor
ID - Idaho
PNR - Pittsburgh Naval Reactor
HQ - Headquarters
D-23
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(Fieast read Instruction! on she reverse be}o,
1. REPORT N<
EPA 520/1-84-022-2
3. RECIPIENT'S ACCESSION
4, TJTL.E AND SUBTITLE
Background Information Document (Volume II)
(Integrated Risk Assessment)
Pittal Rules for Radionuclides
5. REPORT DATE
October 22, 1984
6, PERFORMING ORGANIZATION COD6
7. AUTHORCSt
8. PERFORMING ORGANIZATION REPORT NO,
9. PERFORMING ORGANIZATION NAMi AND ADDRESS
Environmental Protection Agency
Office of Radiation Programs
Washington, B.C. 20460
10, PROGRAM ELEMENT NO.
11, CONTRACT/GRANT NO,
12. SPONSORING AGENCY NAME AMD ADDRESS
13, TYPE OF REPORT AND PERIOD COVERED
14. SPONSORING AGENCY CODE
15. SUPPLEMENTARY NOTES
16. ABSTRACT
This report serves as a background information document in support of
the Environmental Protection Agency's final rules for sources of emissions
of radionuclides pursuant to Section 112 of the Clean Air Act.
This report 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 licensed facilities 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, we present the following information; (1) a
general description of the source category, (2) a brief description of the
processes that lead to the emissions of radionuclides into air, (3) a
summary of emissions data, and (4) estimates of the radiation doses and
health risks to both individuals and populations.
7.
KEY WORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
b.lDENTlPlERS/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
8. DISTRIBUTION STATEMENT
Unlimited
19, SECURITY CLASS (This Report)
Unclassified
21. NO. OF PAGES
474
20, SECURITY CLASS (Thispage)
Unclassified
22. PRICE
EPA Fofui 2220-1 (R*», 4-77) PREVIOUS EDITION is OBSOLETE
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