United States Office of Radiation Programs EPA/520/1-89-006-1
Environmental Protection (ANR-459) September 1989
Agency
4>EPA Risk Assessments 520139006-1
Environmental Impact
Statement
NESHAPS for Radionuclides
Background Information
Document — Volume 2
SM-XJL
Printed on Recycled Paper
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40 CFR Part 61 EPA 520/1-89-006-1
National Emission Standards
for Hazardous Air Pollutants
Risk Assessments
Environmental Impact Statement
for NESHAPS Radionuclides
VOLUME 2
BACKGROUND INFORMATION DOCUMENT
September 1989
U.S. Environmental Protection Agency
Office of Radiation Programs
Washington, D.C. 20460
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Preface
The Environmental Protection Agency is promulgating National
Emission Standards for Hazardous Air Pollutants (NESHAPs) for
Radionuclides. An Environmental Impact Statement (EIS) has been
prepared in support of the rulemaking. The EIS consists of the
following three volumes:
VOLUME I - Risk Assessment Methodology
This document contains chapters on hazard
identification, movement of radionuclides through
environmental pathways, radiation dosimetry,
estimating the risk of health effects resulting from
expose to low levels of ionizing radiation, and a
summary'Of the uncertainties in calculations of dose
and risks.
VOLUME II - Risk Assessments
This document contains a chapter on each radionuclide
source category studied. The chapters include an
introduction, category description, process
description, control technology, health impact
assessment, supplemental control technology, and cost.
It has an appendix which contains the inputs to all
the computer runs used to generate the risk
assessment.
VOLUME III - Economic Assessment
This document has chapters on each radionuclide source
category studied. Each chapter includes an
introduction, industry profile, summary of emissions,
risk levels, the benefits and costs of emission
controls, and economic impact evaluations.
Copies of the EIS in whole or in part are available to all
interested persons; an announcement of the availability appears in
*-he Federal Register.
111
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For additional information, contact James Hardin at
(202) 475-9610 or write to:
Director, Criteria and Standards Division
Office of Radiation Programs (ANR-460)
Environmental Protection Agency
401 M Street, SW
Washington, DC 20460
IV
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LIST OF PREPARERS
Various staff members from EPA's Office of Radiation Programs
contributed in the development and preparation of the EIS.
Terrence McLaughlin
James Hardin
Byron Hunger
Fran Cohen
Albert Colli
Larry Gray
W. Daniel Hendricks
Paul Magno
Christopher B. Nelson
Dr. Neal S. Nelson
Barry Parks
Dr. Jerome Pushkin
Jack L. Russell
Dr. James T. Walker
Larry Weinstock
Chief, Environmental
Standards Branch
Health Physicist
Economist
Attorney Advisor
Environmental
Scientist
Environmental
Scientist
Environmental
Scientist
Environmental
Scientist
Environmental
Scientist
Radiobiologist
Health Physicist
Chief Bioeffects
Analysis Branch
Engineer
Radiation
Biophysicist
Attorney Advisor
Project Officer
Author/Reviewer
Reviewer
Author/Reviewer
Author/Reviewer
Reviewer
Author/Reviewer
Author
Author
Reviewer
Author/Reviewer
Author/Reviewer
Author
Reviewer
An EPA contractor,. S. Cohen and Associates, Inc., McLean, VA,
provided significant technical support in the preparation of the
EIS.
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TABLE OF CONTENTS
VOLUME II: RISK ASSESSMENT
LIST OF TABLES ...................... ix
LIST OF FIGURES .................... xxxiii
1. INTRODUCTION .................... -L_i
2. DEPARTMENT OF ENERGY (DOE) FACILITIES ........ 2-1
2.1 OVERVIEW AND SUMMARY OF RESULTS. ... ....... 2-1
2.2 RMI COMPANY ................. '.'.'. 2-21
2.3 LOS ALAMOS NATIONAL LABORATORY .!!!!!!!•"" 2-24
2.4 HANFORD RESERVATION ........... '.'.''' 2-33
2.5 OAK RIDGE RESERVATION ......... ...... 2_42
2.6 SAVANNAH RIVER PLANT .......... .*.*.*** 2-51
2.7 FEED MATERIALS PRODUCTION CENTER ........ 2-58
2.8 BROOKHAVEN NATIONAL LABORATORY .... " " ' 2-66
2.9 MOUND FACILITY ............. .".*.*.*" 2-71
2.10 IDAHO NATIONAL ENGINEERING
LABORATORY ................. 2-72
2.11 LAWRENCE BERKELEY LABORATORY ........'"' 2-79
2.12 PADUCAH GASEOUS DIFFUSION PLANT ...... . 2-82
2.13 LAWRENCE LIVERMORE LABORATORY ....... *.*.*. ".2-84
2.14 PORTSMOUTH GASEOUS DIFFUSION
2_86
2.15 ARGONNE NATIONAL LABORATORY ....... 2-89
2.16 PINELLAS PLANT ............... * 2_92
2.17 NEVADA TEST SITE .......... ....... 2-94
2.18 KNOLLS LABORATORY - KESSELRING ..... ! ! ! ! ! 2-96
2.19 BATTELLE COLUMBUS LABORATORY ....... ! ! ! " 2-98
2.20 FERMI NATIONAL LABORATORY ......... . . . ! 2-100
2.21 SANDIA NATIONAL LABORATORY ..... ...... ! 2-103
2.22 BETTIS ATOMIC POWER LABORATORY ....... '. '. ! 2-106
2.23 KNOLLS LAB - WINDSOR ............ • . 2-108
2.24 ROCKY FLATS PLANT .......... '.'.'.'' ' 2-109
2.25 PANTEX PLANT .............. . . . . ! 2-112
2.26 KNOLLS LAB - KNOLLS ........... ! ! ! ! ! 2-114
2.27 AMES LABORATORY ............. ! ! ! * * 2-117
2.28 ROCKETDYNE ROCKWELL .......... .".**'' " 2-118
2.29 REFERENCES ................. ] \ ! 2-122
3. NRC-LI CENSED AND NON-DOE FEDERAL FACILITIES ....... 3-1
3.1 INTRODUCTION AND BACKGROUND .......... " "3-1
3.2 HOSPITALS ................... [ '3-2
3.3 RADIOPHARMACEUTICAL MANUFACTURERS ......'' 3-6
3.4 LABORATORIES .................. . 3_10
3.5 RESEARCH AND TEST REACTORS ........ . . . . . 3-13
3.6 SEALED SOURCE MANUFACTURERS ......... '.'.'. 3-16
3.7 NON-LWR FUEL FABRICATORS ............. [ 3-2Q
3.8 SOURCE MATERIAL LICENSEES ......... ! ! ! ! 3-23
3.9 LOW- LEVEL WASTE INCINERATORS ......... '.'.'. 3-25
3.10 NON-DOE FEDERAL FACILITIES ...... ...... ] 3-28
VI
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3.11 SUMMARY OF THE COLLECTIVE RISKS FROM ALL FACILITIES 3-30
3.12 REFERENCES 3~32
4. URANIUM FUEL CYCLE FACILITIES 4-1
4.1 INTRODUCTION 4-1
4.2 URANIUM MILLS 4-3
4.3 URANIUM CONVERSION FACILITIES 4-25
4.4 FUEL FABRICATION FACILITIES 4-31
4.5 NUCLEAR POWER FACILITIES 4-40
4.6 SUMMARY 4-67
4.7 REFERENCES 4-69
5. HIGH-LEVEL WASTE DISPOSAL FACILITIES 5-1
5.1 DESCRIPTION OF THE HIGH-LEVEL WASTE DISPOSAL
FACILITIES 5-1
5.2 BASIS OF THE EXPOSURE AND RISK EVALUATION 5-4
5.3 RESULTS OF THE DOSE AND RISK ASSESSMENT 5-7
5.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS 5-9
5.5 REFERENCES 5-11
6. ELEMENTAL PHOSPHORUS PLANTS 6-1
6.1 DESCRIPTION OF THE SOURCE CATEGORY 6-1
6.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT 6-3
6.3 RESULTS OF THE EXPOSURE AND RISK ASSESSMENT .... 6-13
6.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS 6-18
6.5 REFERENCES 6~23
7. COAL-FIRED UTILITY AND INDUSTRIAL BOILERS 7-1
7.1 INTRODUCTION 7~1
7.2 UTILITY BOILERS 7~5
7.3 INDUSTRIAL BOILERS 7-19
7.4 REFERENCES 7"25
8. INACTIVE URANIUM MILL TAILINGS 8-1
8.1 DESCRIPTION OF INACTIVE URANIUM MILL TAILINGS
SITES 8~1
8.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT 8-3
8.3 RESULTS OF THE RISK ASSESSMENT FOR INACTIVE MILLS . .8-5
8.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS 8-17
8.5 REFERENCES 8~30
9. LICENSED URANIUM MILL TAILINGS FACILITIES 9-1
9.1 DESCRIPTION OF LICENSED URANIUM MILL TAILINGS . . . .9-1
9.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT 9-5
9.3 RESULTS OF THE RISK ASSESSMENTS FOR LICENSED MILLS. 9-12
9.4 SUPPEMENTARY CONTROL OPTIONS AND COSTS 9-26
9.5 REFERENCES 9~52
10. DEPARTMENT OF ENERGY RADON SITES 10-1
10.1 SITE-DESCRIPTIONS 10-1
10.2 BASIS OF THE RISK ASSESSMENT 10-9
10.3 RESULTS OF THE RISK ASSESSMENT 10-11
10.4 SUPPLEMENTARY CONTROL OPTIONS AND COST 10-20
10.5 REFERENCES 10-23
vii
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11. UNDERGROUND URANIUM MINES i:L-l
11.1 GENERAL DESCRIPTION 11-1
11.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT! ! '. 11-5
11.3 RESULTS OF THE EXPOSURE AND RISK ASSESSMENT. . . 11-10
11.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS. 11-14
11.5 REFERENCES ] !11-30
12. SURFACE URANIUM MINES 12-1
12.1 GENERAL DESCRIPTION !!!!!!!!' 12-1
12.2 BASIS OF THE DOSE AND RISK ASSESSMENT. . '. . '. ' ' 12-12
12.3 RESULTS OF THE DOSE AND RISK ASSESSMENT ' 12-12
12.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS. " *12-19
12.5 REFERENCES \ !12-21
13. PHOSPHOGYPSUM STACKS 13_1
13.1 SOURCE CATEGORY DESCRIPTION. ..!!!!!!*'* 13-1
13.2 RADIONUCLIDE EMISSIONS ! . ! ! 13-8
13.3 RESULTS OF THE HEALTH IMPACT ASSESSMENT! ! ! ! ! * 13-21
13.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS. ' '13-27
13.5 REFERENCES ' * *13 39
Vlll
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LIST OF TABLES
VOLUME II: RISK ASSESSMENT
Table 2.1-1. Department of energy facilities 2-2
Table 2.1-2. Summary of doses and risks to nearby
individuals from DOE facilities due to 1986
emissions 2~3
Table 2.1-3. Distribution of fatal cancer risk
in the population 2~7
Table 2.1-4. Summary of doses and risks to the regional
population (0-80 km) around DOE facilities . . .2-8
Table 2.1-5. Baseline risk assessment for DOE facilities. . 2-14
Table 2.1-6. Risks when emissions are limited
to 3 mrem/y EDE 2-14
Table 2.1-7 Risks when emissions are limited
to 1 mrem/y EDE 2"~15
Table 2.1-8 Maximum individual risk, with Alternative 4
supplemental control strategies 2-16
Table 2.1-9 Fatal cancers/year to nearby individuals
with Alternative 4 supplemental control
techniques 2-19
Table 2.1-10 Distribution of fatal cancer risk in the
populations within 80 km with Alternative 4
supplemental control techniques 2-21
Table 2.2-1. Radionuclides released to air during
1986 from RMI 2~22
Table 2.2-2. Estimated radiation dose rates from RMI. . . . 2-23
Table 2.2-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from RMI 2~23
Table 2.3-1. Radionuclides released to air during
1986 from Los Alamos Scientific Laboratory . . 2-28
Table 2.3-2. Estimated radiation doses from the
Los Alamos Laboratory 2-29
Table 2.3-3. Estimated fatal cancer risks from the
Los Alamos Laboratory 2~29
IX
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Table 2.3-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
gopulation from the Los Alamos Scientific
Laboratory 2-29
Table 2.3-5. Effects of holdup time on the release
of air activation products from the
proposed stack serving the LAMPF beam
st°P 2-30
Table 2.4-1. Radionuclides released to air during 1986
from the Hanford Reservation 2-40
Table 2.4-2. Estimated radiation dose rates from the
Hanford Reservation 2-41
Table 2.4-3. Estimated fatal cancer risks from the
Hanford Reservation 2-41
Table 2.4-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Hanford Reservation. . . . 2-41
Table 2.5-1. Radionuclides released to air from Oak
Ridge Reservation during 1986 2-44
Table 2.5-2. Estimated radiation dose rates from the
Oak Ridge National Laboratory 2-45
Table 2.5-3. Estimated fatal cancer risks from the
Oak Ridge National Laboratory 2-45
Table 2.5-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from Oak Ridge National
Laboratory 2-45
Table 2.5-5. Anticipated new emission rate for tritium
at CRGDF 2_46
Table 2.5-6. Summary of capital and operating costs
for supplementary controls at the Oak
Ridge Reservation 2-50
Table 2.6-1. Radionuclides released to air during 1986
from Savannah River Plant. . 2-54
Table 2.6-2. Estimated radiation dose rates from the
Savannah River Plant 2-55
Table 2.6-3. Estimated fatal cancer risks from the
Savannah River Plant 2-55
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Table 2.6-4.
Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Savannah River Plant .
2-56
Table 2.7-1. Radionuclides released to air during
1986 from FMPC 2-59
Table 2.7-2. Estimated radiation dose rates from FMPC . . . 2-60
Table 2.7-3. Estimated fatal cancer risks from FMPC .... 2-60
Table 2.7-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from FMPC 2-61
Table 2.7-6. Cost estimates for acquisition and
installation of HEPA filter systems 2-64
Table 2.8-1. Radionuclide emission points stacks at
Brookhaven National Laboratory 2-67
Table 2.8-2. Radionuclides released to air during 1986
from Brookhaven National Laboratory 2-69
Table 2.8-3. Estimated radiation dose rates from the
Brookhaven National Laboratory 2-70
Table 2.8-4. Estimated fatal cancer risks from the
Brookhaven National Laboratory 2-70
Table 2.8-5. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Brookhaven National
Laboratories 2-71
Table 2.9-1. Radionuclides released to air during 1986
from Mound Facility 2-72
Table 2.9-2. Estimated radiation dose rates from the
Mound Facility 2-73
Table 2.9-3. Estimated fatal cancer risks from the
Mount Facility 2-73
Table 2.9-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Mound Facility 2-73
Table 2.10-1. R,adionuclides released to air during 1986
from all Idaho Facilities 2-78
Table 2.10-2. Estimated radiation dose rates from the
Idaho National Engineering Laboratory 2-79
XI
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Estimated ratal cancer risks from the
Idaho National Engineering Laboratory.
. . 2-79
Table 2.10-3
Table 2.10-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from INEL facilities 2-79
Table 2.11-1. Radionuclides released to air during 1986
from Lawrence Berkeley Laboratory 2-81
Table 2.11-2. Estimated radiation dose rates from the
Lawrence Berkeley Laboratory 2-82
Table 2.11-3. Estimated fatal cancer risks from the
Lawrence Berkeley Laboratory 2-82
Table 2.11-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Lawrence Berkeley
Laboratory 2-82
Table 2.12-1. Radionuclides released to air during 1986
from Paducah Gaseous Diffusion Plant 2-83
Table 2.12-2. Estimated radiation dose rates from the
Paducah Gaseous Diffusion Plant 2-84
Table 2.12-3. Estimated fatal cancer risks from the
Pacucah Gaseous Diffusion Plant 2-84
Table 2.12-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Paducah Gaseous
Diffusion Plant 2-84
Table 2.13-1. Source terms and release point
characterization 2-85
Table 2.13-2. Estimated radiation dose rates from
Lawrence Livermore Laboratory/Sandia
Livermore 2-86
Table 2.13-3. Estimated fatal cancer risks from
Lawrence Livermore Laboratory/Sandia
Livermore 2-86
Table 2.13-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from Lawrence Livermore
Laboratory/Sandia Livermore 2-86
Table 2.14-1. Radionuclides released to air during 1986
from the Portsmouth Gaseous Diffusion
Plant 2-88
xn
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Table 2.14-2.
Table 2.14-3
Table 2.14-4.
Table 2.15-1.
Table 2.15-2.
Table 2.15-3.
Table 2.15-4.
Estimated radiation dose rates from the
Portsmouth Gaseous Diffusion Plant 2-89
Estimated fatal cancer risks from the
Portsmouth Diffusion Plant 2-89
Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Portsmouth Gaseous
Diffusion Plant 2-89
Radionuclides released to air during 1986
from Argonne National Laboratory 2-90
Estimated radiation dose rates from the
Argonne National Laboratory 2-91
Estimated fatal cancer risks from the
Argonne National Laboratory 2-91
Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Argonne National
Laboratory
2-92
Table 2.16-1.
Table 2.16-2.
Table 2.16-3.
Table 2.16-4.
Table 2.17-1.
Table 2.17-2.
Table 2.17-3.
Table 2.17-4.
Table 2.18-1.
Radionuclides released to air during 1986
from Pinellas Plant 2-92
Estimated radiation dose rates from the
Pinellas Plant 2-93
Estimated fatal cancer risks from the
Pinellas Plant 2-93
Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Pinellas Plant 2-94
Radionuclides released to air during 1986
from the Nevada Test Site 2-95
Estimateed radiation dose rates from the
Nevada Test Site 2-95
Estimated fatal cancer risks from the
Nevada Test Site 2-96
Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Nevada Test Site 2-96
Radionuclides released to air during 1986
from Knolls Atomic Power Lab-Kesselring. . . . 2-97
Xlll
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Table 2.18-2. Estimated radiation dose rates from the
Knolls Lab-Kesselring 2-98
Table 2.18-3. Estimated fatal cancer risks from the
Knolls Lab-Kesselring 2-98
Table 2.18-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from Knolls Atomic Power Lab-
Kesselring 2-98
Table 2.19-1. Radionuclides released to air during 1986
from Battelle Columbus 2-100
Table 2.19-2. Estimated radiation dose rates from the
Battelle Columbus Laboratory 2-101
Table 2.19-3. Estimated fatal cancer risks from the
Battelle Columbus Laboratory 2-101
Table 2.19-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from Battelle Columbus 2-101
Table 2.20-1. Radionuclides released to air during 1986
from Fermi National Accelerator
Laboratory 2-102
Table 2.20-2. Estimated radiation dose rates from the
Fermi National Laboratory 2-103
Table 2.20-3. Estimated fatal cancer risks from the
Fermi National Laboratory 2-103
Table 2.20-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Fermi National
Laboratory 2-103
Table 2.21-1. Radionuclides released to air during 1986
from Sandia National Laboratory/Lovelace
Research Institute 2-104
Table 2.21-2. Estimated radiation dose rates from the
Sandia National Laboratory/Lovelace Research
Institute 2-105
Table 2.21-3. Estimated fatal cancer risks from the
Sandia National Laboratory/Lovelace Research
Institute 2-105
xiv
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Table 2.21-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Sandia National
Laboratory/Lovelace Research Institute . . . .2-105
Table 2.22-1. Radionuclides released to air during 1986
from Bettis Atomic Power Laboratory 2-106
Table 2.22-2. Estimated radiation dose rates from the
Bettis Atomic Power Laboratory 2-107
Table 2.22-3. Estimated fatal cancer risks from the
Bettis Atomic Power Laboratory 2-107
Table 2.22-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Bettis Atomic Power
Laboratory 2-107
Table 2.23-1. Radionuclides released to air during 1986
from Knolls Atomic Power Lab-Windsor 2-108
Table 2.23-2. Estimated radiation dose rates from the
Knolls Lab-Windsor 2-109
Table 2.23-3. Estimated fatal cancer risks from the
Knolls Lab-Windsor 2-109
Table 2.23-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Knolls Atomic Power
Lab-Windsor 2-110
Table 2.24-1. Radionuclides released to air during 1986
from Rocky Flats Plant 2-111
Table 2.24-2. Estimated radiation dose rates from the
Rocky Flats Plant 2-112
Table 2.24-3. Estimated fatal cancer risks from the
Rocky Flats Plant 2-112
Table 2.24-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Rocky Flats Plant 2-112
Table 2.25-1. Radionuclides released to air during 1986
from the Pantex Plant 2-113
Table 2.25-2. Estimated radiation dose rates from the
Pantex Plant 2-114
Table 2.25-3. Estimated fatal cancer risks from the
Pantex Plant 2-114
xv
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Table 2.25-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Pantex Plant 2-114
Table 2.26-1. Radionuclides released to air during 1986
from Knolls Atomic Power Lab-Knolls 2-115
Table 2.26-2. Estimated radiation dose rates from the
Knolls Lab-Knolls 2-116
Table 2.26-3. Estimated fatal cancer risks from the
Knolls Lab-Knolls 2-116
Table 2.26-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Knolls Atomic Power
Lab-Knolls
2-116
Table 2.27-1. Radionuclides released to air during 1986
from Ames Laboratory 2-117
Table 2.27-2. Estimated radiation dose rates from the
Ames Laboratory 2-118
Table 2.27-3. Estimated fatal cancer risks from the Ames
Laboratory 2_118
Table 2.27-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from the Ames Laboratory 2-118
Table 2.28-1. Radionuclides released to air during 1986
from Rocketdyne Division, Rockwell
International 2-119
Table 2.28-2. Estimated radiation dose rates from
Rocketdyne Division, Rockwell
International 2-120
Table 2.28-3. Estimated fatal cancer risks from
Rocketdyne Division, Rockwell
International 2-120
Table 2.28-4. Estimated distribution of the fatal
cancer risk to the regional (0-80 km)
population from Rocketdyne Division,
Rockwell International 2-121
Table 3-1. Estimated emissions from model hospitals 3-3
Table 3-2. Estimated radiation dose rates from model
hospitals 3_5
xvi
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Table 3-3. Estimated fatal cancer risks from model
hospitals 3-5
Table 3-4. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from all hospitals 3-5
Table 3-5. Effluent release rates (Ci/y) for
radiopharmaceutical manufacturers 3-7
Table 3-6. Estimated radiation dose rates from
radiopharmaceutical manufacturers 3-8
Table 3-7. Estimated fatal cancer risks from reference
radiopharmaceutical manufacturers 3-8
Table 3-8. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations from
all radiopharmaceutical manufacturers 3-9
Table 3-9. Effluent release rates (Ci/y) for laboratories . 3-11
Table 3-10. Estimated radiation dose rates from
laboratories 3-12
Table 3-11. Estimated fatal cancer risks from laboratories . 3-12
Table 3-12. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from all
laboratories 3-13
Table 3-13. Effluent release rates (Ci/y) for research
reactors 3-14
Table 3-14. Estimated radiation dose rates from research
reactors 3-15
Table 3-15. Estimated fatal cancer risks from research
reactors 3-15
Table 3-16. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from research and test reactors 3-16
Table 3-17. Effluent release rates (Ci/y) for sealed
source manufacturers 3-17
Table 3-18. Estimated radiation dose rates from sealed
source manufacturers 3-18
Table 3-19. Estimated fatal cancer risks from sealed
source manufacturers 3-19
xvi i
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Table 3-20. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from sealed source manufacturers 3-19
Table 3-21. Effluent release rates (Ci/y) for non-LWR fuel
fabricators 3-21
Table 3-22. Estimated radiation dose rates from non-LWR
fuel fabricators 3-22
Table 3-23. Estimated fatal cancer risks from non-LWR fuel
fabricators 3-22
Table 3-24. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from all non-LWR fuel fabricators. 3-22
Table 3-25. Effluent release rates for source material
licensees 3-23
Table 3-26. Estimated radiation dose rates from source
material licensees 3-24
Table 3-27. Estimated fatal cancer risks from source
material licensees 3-24
Table 3-28. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from all source material licensees 3-25
Table 3-29. Effluent release rates (Ci/y) for low-level
waste disposal facilities. 3-26
Table 3-30. Estimated radiation dose rates from low-level
waste disposal facilities 3-27
Table 3-31. Estimated fatal cancer risks from low-level
waste disposal facilities 3-27
Table 3-32. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from all low-level waste disposal facilities . . 3-27
Table 3-33. Effluent release rates (Ci/y) for DOD
facilities 3-29
Table 3-34. Estimated radiation dose rates from DOD
facilities 3-29
Table 3-35. Estimated fatal cancer risks from DOD
facilities 3-30
XVlll
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Table 3-36. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from all DOD facilities 3-30
Table 3-37. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from all NRC-licensed facilities 3-31
Table 4-1. Uranium mills licensed by the U.S. Nuclear
Regulatory Commission as of December 1988 4-4
Table 4-2. Source terms for uranium milling 4-10
Table 4-3. Areas of the tailings impoundments at uranium
mills and average radium-226 concentrations. . . 4-14
Table 4-4. Sources of meteorological data used in the
assessment of uranium milling 4-15
Table 4-5. Estimated populations living within 0 to 5 km
of active uranium milling facilities 4-16
Table 4-6. Estimated radiation dose rates from
uranium mills 4-17
Table 4-7. Estimated fatal cancer risks from uranium
mills 4-19
Table 4-8. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from uranium mills 4-19
Table 4-9. Effluent controls for process emissions 4-20
Table 4-11. Reported atmospheric radioactive emissions
for uranium conversion facilities (Ci/y) .... 4-27
Table 4-12. Atmospheric radioactive emissions assumed
for reference dry and wet process uranium
conversion facilities 4-29
Table 4-13. Radiation dose equivalent rates from
atmospheric radioactive emissions from
reference uranium conversion facilities 4-29
Table 4-14. Fatal cancer risks due to atmospheric
radioactive emissions from reference
uranium conversion facilities 4-30
Table 4-15. Estimated distribution of lifetime fatal
cancer risks projected for uranium conversion
facilities 4-31
xix
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Table 4-16. Light water reactor commercial fuel fabrication
facilities licensed by the Nuclear
Regulatory Commission as of June 1987 4-33
Table 4-17. Light water reactor commercial fuel
fabrication facilities reported annual
uranium effluent releases for 1983 through
1987 in uCi/y 4_35
Table 4-18. Atmospheric radioactive emissions assumptions
for reference fuel fabrication facility 4-37
Table 4-19. Radiation dose equivalent rates from
atmospheric radioactive emissions from model
fuel fabrication facility 4-39
Table 4-20. Fatal cancer risks due to atmospheric
radioactive emissions from reference fuel
fabrication facility 4-39
Table 4-21. Estimated distribution of lifetime fatal
cancer risks projected for all fuel
fabrication facilities 4-40
Table 4-22. U.S. nuclear power generating units operable
as of December 31, 1986 (DOE87) 4-41
Table 4-23. Geometric mean and standard deviation
by year for selected radionuclides for
boiling water reactors in the United
.States for 1981 through 1985 in uCi/y 4-49
Table 4-24. Geometric mean and standard deviation
by year for selected radionuclides for
pressurized water reactors in the United
States for 1981 through 1985 in uCi/y 4-51
Table 4-25. Atmospheric radioactive emissions assumed for
model boiling water reactor 4-53
Table 4-26. Atmospheric radioactive emissions assumed for
model pressurized water reactor 4-54
Table 4-27. Minimum, maximum, median, and 90th percentile
population densities for nuclear power reactor
sites in the United States 4-55
Table 4-28. Dose rates from model light water reactors . . . 4-56
Table 4-29. Fatal cancer risks for model light water
reactors 4-56
Table 4-30. Estimated distribution of lifetime fatal
cancer risks projected for all power reactors. . 4-58
xx
-------
Table 4-31. Doses to maximally exposed individuals
in mrem/y 4~59
Table 4-32. Summary of fatal cancer risks from atmospheric
radioactive emissions from uranium fuel cycle
facilities 4~67
Table 4-33. Estimated distribution of lifetime fatal cancer
risks for uranium fuel cycle facilities 4-68
Table 5-1. Projected generation of spent fuel 5^2
Table 5-2. Emissions from normal operations at HLW disposal
facilities 5~5
Table 5-3. WIPP discharge stacks 5-7
Table 5-4. Estimated radiation dose rates from high-level
waste disposal facilities 5-8
Table 5-5. Estimated fatal cancer risks from high-level
waste disposal facilities .5-8
Table 5-6. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from high-level waste disposal facilities. . . . 5-10
Table 6-1. Elemental phosphorus plants 6-2
Table 6-2. Radionuclide stack emissions measured at
elemental phosphorus plants (1975-1980) 6-4
Table 6-3. Measured radionuclide concentrations in
process samples at elemental phosphorus
plants - 1983-1984 results 6-6
Table 6-4. Radionuclide emissions from calciners at
elemental phosphorus plants - 1983-1984
results 6~6
Table 6-5. Measured distribution of lead-210 and
polonium-210 by particle size in calciner
stack outlet streams at elemental phosphorus
plants - 1983 results 6-7
Table 6-6. Dissolution of lead-210 and polonium-210
from particulate samples collected from
off-gas streams at FMC and Stauffer elemental
phosphorus plants 6-7
Table 6-7. Lead-210 and polonium-210 emissions measured
in calciner off-gas streams at two elemental
phosphorus plants - 1988 6-9
xxi
-------
Table 6-8. Measured distribution of lead-210 and
polonium-210 by particle size in calciner
stack inlet and outlet streams at elemental
phosphorus plants - 1988 results 6-9
Table 6-9. Estimated annual radionuclide emissions from
elemental phosphorus plants 6-10
Table 6-10. Lung clearance classification and particle
sizes used in the assessment 6-12
Table 6-11. Calciner stack emission characteristics 6-12
Table 6-12. Populations within 80 km and distances to
the maximum exposed individuals of elemental
phosphorus plants with the source of meteoro-
logical data used in dose equivalent and risk
calculations 6-13
Table 6-13. Estimated radiation dose equivalent rates
to the maximum exposed individual and to the
80-km regional population from elemental
phosphorus plants 6_15
Table 6-14. Estimated fatal cancer risks to the maximum
exposed individual and to the 80-km regional
population from elemental phosphorus plants. . . 6-17
Table 6-15. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from operating elemental phosphorus plants . . . 6-18
Table 6-16. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from idle elemental phosphorus plants 6-18
Table 6-17. Estimated Po-210 emission levels achieved
by control alternatives 6-19
Table 6-18. Estimated Pb-210 emission levels achieved
by control alternatives 6-20
Table 6-19. Capital cost of control alternatives
(1,000 1988 $) 6_21
Table 6-20. Annualized cost of control alternatives
(1,000 1988 $) 6_22
Table 7-1. Major decay products of uranium-238 7-3
Table 7-2. Major decay products of thorium-232 7-3
xxn
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Table 7-3. Typical uranium and thorium concentrations
in coal 7-4
Table 7-4. Uranium concentrations and distributions
in coal 7-5
Table 7-5. Coal ash distribution by boiler type 7-7
Table 7-6. Distribution of particulate control equipment
for bituminous coal-fired utility boilers 7-9
Table 7-7. U-238 emission factors for coal-fired utility
boilers 7-11
Table 7-8. Th-232 emission factors for coal-fired utility
boilers 7-12
Table 7-9. Enrichment factors for radionuclides 7-12
Table 7-10. Emissions for typical coal-fired utility
boilers 7"13
Table 7-11. Emissions for large coal-fired utility boilers . 7-14
Table 7-12. Estimated radiation dose rates from typical coal-
fired utility boilers 7-15
Table 7-13. Estimated radiation dose rates from large coal-
fired utility boilers 7-16
Table 7-14. Estimated fatal cancer risk from typical coal-
fired utility boilers 7-17
Table 7-15. Estimated fatal cancer risk from large coal-
fired utility boilers 7-17
Table 7-16. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from all
coal-fired utility boilers 7-18
Table 7-17. Numbers and capacities of industrial boilers . . 7-20
Table 7-18. Estimated radiation dose rates from the reference
coal-fired industrial boiler 7-23
Table 7-19. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from all
coal-fired industrial boilers 7-23
Table 8-1. Quantity of tailings and planned remedial actions
at inactive uranium mill tailings sites 8-4
Table 8-2. Summary of radon-222 emissions from inactive
uranium mill tailings disposal sites 8-6
xxiii
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Table 8-3. Estimated number of persons living within 5 km
of the centroid of tailings disposal sites for
inactive mills 8-7
Table 8-4. Estimated exposures and risks to individuals
living near inactive tailings sites after UMTRCA
disposal is completed 8_8
Table 8-5. Estimated fatal cancers per year in the regional
(0-80 km) populations around inactive tailings
disposal sites 8-10
Table 8-6. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from
inactive uranium mill tailings disposal sites. . 8-11
Table 8-7. Estimated exposures and risks to individuals
living near inactive tailings sites assuming
a 6 pCi/nr/s radon flux limit 8-12
Table 8-8. Estimated fatal cancers per year in the
regional (0-80 km) populations around inactive
tailings disposal sites assuming a 6 pCi/m2/s
radon flux limit 8-13
Table 8-9. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from
inactive uranium mill tailings disposal sites
assuming a 6 pci/m2/s radon flux limit 8-14
Table 8-10. -Estimated exposures and risks to individuals
living near inactive tailings sites assuming
a 2 pci/mvs radon flux limit 8-15
Table 8-11. Estimated fatal cancers per year in the
regional (0-80 km) populations around
inactive tailings disposal sites assuming a
2 pCi/mVs radon flux limit 8-16
Table 8-12. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations from •
inactive uranium mill tailings disposal sites
assuming a 2 pCi/m2/s radon flux limit 8-17
Table 8-13. Estimated depths of earth cover needed to
achieve given radon flux rates 8-23
Table 8-14. Major volumes and surface areas used to
calculate the costs to achieve given
radon-222 flux rates 8-25
Table 8-15. Estimated costs of reducing average radon-222
flux rate to 20 pCi/m2/s 8-26
xxiv
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Table 8-16. Estimated costs of reducing average radon-222
flux rate to 6 pCi/nr/s 8~27
Table 8-17. Estimated costs of reducing average radon-222
flux rate to 2 pCi/m^/s 8-28
Table 9-1. Operating status of licensed conventional
uranium mills as of June 1989 9-3
Table 9-2. Summary of operable tailings impoundment areas
and radium-226 content at operating and standby
mills 9"7
Table 9-3. Summary of radon source terms calculated for
operable mill tailings impoundments 9-9
Table 9-4. Summary of uranium mill tailings impoundment
areas, flux rates, and post-UMTRCA radon-222
release rates 9-10
Table 9-5. Estimated number of persons living within
5 km of the centreid of tailings impoundments
of licensed mills 9-11
Table 9-6. Estimated exposures and risks to individuals
living near operable tailings impoundments . . . 9-13
Table 9-7. Estimated fatal cancers per year in the
regional (0-80 km) populations around operable
tailings impoundments 9-15
Table 9-8. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from operable uranium mill tailings piles. . . . 9-15
Table 9-9. Estimated exposures and risks to individuals
living near licensed tailings impoundments
post-UMTRCA disposal 9-17
Table 9-10. Estimated fatal cancers per year in the
regional (0-80 km) populations around
licensed tailings impoundments post-UMTRCA
disposal 9~18
Table 9-11. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from licensed uranium mill tailings piles
post-UMTRCA disposal 9-19
Table 9-12. Estimated exposures and risks to individuals
living near licensed tailings impoundments
post-disposal to 6 pCi/m2/s ' 9-20
XXV
-------
Table 9-13. Estimated fatal cancers per year in the
regional (0-80 km) populations around licensed
tailings impoundments post-disposal to
6 PCi/m2/s 9_21
Table 9-14. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from licensed uranium mill tailings piles
post-disposal to 6 pCi/m2/s 9-22
Table 9-15. Estimated exposures and risks to individuals
living near licensed tailings impoundments
post-disposal to 2 pCi/m2/s 9_23
Table 9-16. Estimated fatal cancers per year in the
regional (0-80 km) populations around
licensed tailings impoundments post-disposal
to 2 pci/m2/s T . . . . 9-24
Table 9-17. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) populations
from licensed uranium mill tailings piles
post-disposal to 2 pCi/m2/s 9_25
Table 9-18. Estimated depths of earth cover needed to
achieve given radon flux rates 9-27
Table 9-19. Estimated costs of reducing average
radon-222 flux rate to 20 pCi/m2/s 9-30
Table 9-20. Estimated costs of reducing average
radon-222 flux rate to 6 pCi/m2/s 9-3!
Table 9-21. Estimated costs of reducing average
radon-222 flux rate to 2 pCi/m2/s 9-32
Table 9-22. Estimated total costs for new tailings control
technologies 9-36
Table 9-23. Summary of estimated radon-222 emissions
for new tailings control technologies 9-35
Table 9-24. Unit cost categories for partially
below-grade impoundments 9-43
Table 9-25. Costs for a single cell partially below-grade
new model tailings impoundment 9-45
Table 9-26. Costs for a phased design, partially
below-grade, new model tailings impoundment. . . 9-46
Table 9-27. Costs for a continuous design, partially
below-new model tailings impoundment 9-47
xxvi
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Table 9-28. Additional areas of operable impoundments to
be controlled to achieve average radon-222
flux of 20 pCi/mz/s,
9-51
Table 10-1.
Characteristics of the four raffinate
pits and activity levels of major radio-
nuclides in the currently stored materials,
. 10-4
Table 10-2. Estimated volumes of radioactive wastes
stored in Weldon Spring Quarry 10-6
Table 10-3. Volumes of contaminated soil on
the MSP storage pads 10~7
Table 10-4. Radon spurce strength, areas, and
radon flux rates at the MUMT 10-11
Table 10-5. Estimated exposures and risks to individuals
living near DOE radon sites assuming current
radon emission rates 10-12
Table 10-6. Estimated exposures and risks to individuals
living near DOE radon sites assuming
post-remediation radon emission rates 10-13
Table 10-7. Estimated fatal cancers/year to the regional
(0-80 km) populations around DOE radon sites
for current radon emission rates 10-14
Table 10-8. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the FMPC for current radon emission
rates • - -10-15
Table 10-9. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the NFSS for current radon emission
rates 10-15
Table 10-10. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the WSCP for current radon emission
rates 1Q-16
Table 10-11. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the WSQ for current radon emission
rates . . . ..10-16
Table 10-12. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the MSP for current radon emission
rates 1Q-17
XXV11
-------
Table 10-13. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the MUMT for current radon emission
rates .10-17
Table 10-14. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the FMPC for post-remediation radon
emission rates 10-18
Table 10-15. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the MSP for post-remediation radon
emission rates 10-18
Table 10-16. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around the MUMT for post-remediation radon
emission rates 10-19
Table 10-17. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around all DOE radon sites for current
radon emission rates 10-19
Table 10-18. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around all DOE radon sites for post-
remediation radon emission rates. ...... .10-20
Table 10-19. Summary of capital costs to reduce radon
emissions from DOE radon sites 10-22
Table 11-1. Currently operating underground uranium
mines in the United States 11-2
Table 11-2
Table 11-3,
Table 11-4,
Table 11-5,
Table 11-6,
Estimated annual radon-222 emissions from
underground uranium mining sources (EPA83b) . . 11-6
Radon-222 concentrations and annual release
rates in mine ventilation exhaust air 11-7
Estimated exposures and risks to individuals
living near underground uranium mines 11-11
Estimate committed fatal cancers per year
due to radon-222 emissions from underground
uranium mines
11-13
Estimated distribution of the fatal cancer
risk caused by radon-222 emissions from all
underground uranium mines
11-14
xxvi11
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Table
Table
Table
Table
Table
Table
Table
Table
11-7. Current mine ventilation exhaust vents 11-19
11-8. Estimated lifetime fatal cancer risk to the
maximum exposed individual and the committed
fatal cancers per year due to radon-222
emissions from underground uranium mines as
a function of vent stack height 11-21
11-9. Effectiveness of various stack heights 11-24
11-10. Estimated costs (dollars) to extend the
heights of the ventilation exhaust stacks
at each underground uranium mine 11-26
11-A-l. Weights of stack liner per vertical foot. . ll-A-3
ll-A-2. Weights of structural steel used ll-A-3
ll-A-3. Exhaust stack costs (dollars) for
individual stacks ll-A-4
ll-A-4. Number and size of exhaust shafts assumed
for cost estimate • • ll-A-5
Table 12-1.
Table 12-2.
Table 12-3.
Table 12-4.
Table 12-5.
Table 12-6.
Table 12-7.
Table 12-8.
Table 12-9.
Uranium ore production from surface mines,
1948-1986
12-2
Breakdown by state of surface uranium mines
with > 1,000 tons production 12-3
Federal laws, regulations, and guidelines
for uranium mining
12-5
Estimated additional uranium resources by
land status 12~6
Estimated status(a) of surface uranium
mine reclamation 12-11
Mines characterized in the field studies. . . .12-13
Estimated radon-222 emissions from surface
uranium mines 12-14
Estimated particulate emissions from surface
uranium mines
12-15
Estimated exposures and risks to individuals
living near surface uranium mines 12-17
xxix
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Table 12-10. Estimated fatal cancers per year in the
regional (0-80 km) populations due to
radon-222 emissions from surface uranium
mines
12-18
Table 12-11. Estimated distribution of the fatal cancer
risk caused by radon-222 emissions from
all surface uranium mines 12-19
Table 12-12. Estimated lifetime fatal cancer risks from
particulate emissions 12-19
Table 12-13. Estimated depths of cover to reduce radon-222
emissions at surface uranium mines 12-20
Table 12-14. Estimated costs to reduce radon emissions at
surface uranium mines 12-20
Table 13-1.
Table 13-2.
Table 13-3.
Table 13-4.
Table 13-5.
Table 13-6.
Table 13-7.
Table 13-8.
Table 13-9.
Table 13-10.
The location and characteristics of
phosphogypsum stacks in the United States . . . 13-3
Summary of the phosphogypsum stacks in
each state 13-4
Average radionuclide concentrations in
phosphogypsum, pCi/g dry weight 13-5
Results of radon-222 flux measurements
on phosphogypsum stacks in Florida 13-10
Radon-222 flux values applied to various
regions of phosphogypsum stacks 13-11
Results of radon-222 flux measurements on
phosphogypsum stacks in Idaho 13-13
Estimates of annual radon-222 emissions
from phosphogypsum stacks 13-16
Annual radionuclide emissions in fugitive
dust from a model 31-ha phosphogypsum stack . .13-17
Average net airborne radionuclide
concentrations measured at the W.R. Grace
stack
The ten highest individual lifetime risks
estimated to result from radon-222 emissions
from phosphogypsum stacks
13-18
13-22
xxx
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Table 13-11. Estimated increased risk of fatal cancer
and the dose equivalent rates from maximum
exposure to fugitive dusts for an individual
living near phosphogypsum stacks 13-24
Table 13-12. The 10 regional populations estimated to
receive the highest collective risks from
radon-222 emissions from phosphogypsum
stacks 13-24
Table 13-13. Estimated distribution of the fatal cancer
risk caused by radon-222 emissions from
phosphogypsum stacks 13-25
Table 13-14. A summary of the committed fatal cancers
due to radon-222 emissions from phospho-
gypsum stacks located in five regions in
the United States 13-26
Table 13-15. Estimated number of fatal cancers from
fugitive dust emissions for the population
living within 80 km of the model
phosphogypsum stacks 13-27
Table 13-16. Characteristics of gypsum stacks 13-30
Table 13-17. Mean characteristics of the stacks in
each group 13-35
Table 13-18. Radon emissions from grouped gypsum stacks. . .13-36
Table 13-19. Cost of mitigation 13-37
Table 13-20. Risk of cancer death 13-38
Table 13-B-l. Estimated dimensions and areas of
phosphogypsum stacks 13-B-2
Table 13-C-l.
Table 13
Table 13
-C-2.
-D-l.
Table 13-D-2.
Estimated lifetime fatal cancer risks to
nearby individuals caused by radon-222
emissions from phosphogypsum stacks .... 13-C-3
Summary of committed fatal cancers per year
within 80 km of phosphogypsum stacks. . . . 13-C-6
Estimated distribution of lifetime fatal
cancer risk caused by radon-222 emissions
from seven phosphogypsum stacks in Texas. . 13-D-2
Estimated distribution of lifetime fatal
cancer risk caused by radon-222 emissions
from 10 phosphogypsum stacks in the Bartow,
FL, region 13-D-2
xxxi
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Table 13-D-3. Estimated distribution of lifetime fatal
cancer risk caused by radon-222 emissions
from six phosphogypsum stacks in Illinois . 13-D-3
Table 13-D-4. Estimated distribution of lifetime fatal
cancer risk caused by radon-222 emissions
from seven phosphogypsum stacks in
Louisiana 13-D-3
Table 13-D-5. Estimated distribution of lifetime fatal
cancer risk caused by radon-222 emissions
from three phosphogypsum stacks in Idaho. . 13-D-4
Table 13-E-l. Values used to scale risk 13-E-4
Table 13-E-2. Cost breakdown 13-E-6
XXXll
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LIST OF FIGURES
VOLUME II: RISK ASSESSMENT
Figure 9-1. Shape and layout of the model single-cell
impoundment 9-38
Figure 9-2. Size of partially above-grade model single
cell impoundment 9-39
Figure 9-3. Size of below-grade and partially above-grade
cell of model phased impoundment 9-41
Figure 9-4. Shape and layout of model phased disposal
impoundment 9-42
Figure 11-1. General framing plan of a mine ventilation
exhaust stack 11-25
Figure 13-1. Effect of release height on individual risk
for a model stack • • .13-20
XXXlll
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1. INTRODUCTION
The purpose of this report is to serve as a background
information document in support of the Environmental Protection
Agency's (EPA's) final rules for sources of airborne emissions
of radionuclides pursuant to Section 112 of the Clean Air Act.
This report presents an analysis of the exposures and risks
caused by radionuclides emitted into the air from 12 source
categories. The analysis draws upon and updates previous
evaluations and incorporates revisions to the estimates based on
new information developed during the public comment period for
the proposed rules. Specific changes from the analyses presented
in the draft report are noted in the appropriate sections of the
text and on the AIRDOS/DARTAB/ RADRISK input sheets in Appendix
A. The report presents the Agency's most current assessment of
the risks and impacts caused by these facilities. The evaluation
covers the following source categories:
1. Department of Energy (DOE) Facilities;
2. Nuclear Regulatory Commission (NRC) Licensed and
non-DOE Federal Facilities;
3. Uranium Fuel Cycle Facilities;
4. High-Level Waste Disposal Facilities;
5. Elemental Phosphorus Plants;
6. Coal-Fired Boilers;
7. Inactive Uranium Mill Tailings;
8. Licensed Uranium Mill Tailings;
9. DOE Radon Sites;
10. Underground Uranium Mines;
11. Surface Uranium Mines; and
12. Phosphogypsum Stacks.
For each source category, the EPA is presenting the
following information:
1. A general description of the source category,
including a brief description of the processes that
lead to the emission of radionuclides to air and a
characterization of the emission controls that are
currently in use to limit such emissions;
1-1
-------
2. The basis for the exposure and risk assessment,
including radionuclide emissions data,
characteristics of the release point(s), and the
sources for the demographic and meteorological data
that were used;
3. The results of the risk assessment, including
estimates of the exposure and lifetime fatal cancer
risk to nearby individuals, the exposure and number
of committed deaths/year in the regional (0-80 km)
populations, and the distribution of the fatal
cancer risk in the regional populations; and
4. An evaluation of supplementary control options and
costs for source categories or segments of source
categories with the highest estimated risks and
impacts.
In making the risk assessments every effort has been made to
assess facilities on a site-specific basis, using measured data
for emissions and actual data on the configuration of the release
point(s) and the locations of nearby individuals. For source
categories where measured emissions data are not available,
emissions have been estimated using the bases and the assumptions
given for that source category. Where locations of nearby
individuals are not known, the assessment is made to the point of
maximum offsite concentrations. The intent of each assessment is
to provide a realistic estimate of the exposures and risks that
could be received by individuals.
For certain source categories, the number of facilities
makes such site-specific evaluations impractical. In these
instances, for example nuclear power reactors, reference (actual)
facilities are used or model facilities are defined and
evaluated. When a reference or model facility is used, the
exposure and risk estimates presented are for hypothetical
individuals and populations selected as representative of the
demography around actual facilities.
The exposures presented represent 50-year committed dose
equivalents. Estimated doses are presented for organs where the
dose represents 10 percent or more of the fatal cancer risk. For
radon exposures, both the radon concentration (pCi/1) and the
working levels (WL) are reported. The working levels include the
contribution from radon decay products, calculated as a function
of distance (see Volume I).
The fatal cancer risks for nearby or maximum individuals are
lifetime risks. They represents the probability of a typical
individual dying from a lifetime (70 year) exposure to the
concentration of radionuclides estimated at that environmental
location. Chapter 7 of Volume I discusses the uncertainties that
are associated with this assumption. The number of committed
fatal cancers per year (deaths/year) of operation is the
1-2
-------
estimated number of cancers that will occur in the exposed
population from one year's release of radionuclides. Due to the
latency period for cancers, these deaths will occur in the
future, not in the year that the release takes place.
As discussed in Chapter 7 of Volume I, modeling
uncertainties, completeness uncertainties, and parameter
uncertainties are associated with each of the exposure and risk
estimate. However, throughout this volume, exposure and risk
estimates are presented as discrete values. The reader is
referred to Chapter 7 of Volume I and the "Analysis of the
Uncertainties in the Risk Assessment Performed in Support of the
Proposed NESHAPS for Radionuclides" (EPA89) for information on
the range and distribution of the parameter uncertainties
associated with the estimates.
1-3
-------
REFERENCES
EPA89 U.S. Environmental Protection Agency, "Analysis of the
Uncertainties in the Risk Assessment Performed in Support
DC?s1ptem£er?989SHAPS f°r Radi°™cli<^" Washington,
1-4
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2. DEPARTMENT OF ENERGY (DOE) FACILITIES
2.1 OVERVIEW AND SUMMARY OF RESULTS
2.1.1 General Description of DOE Facilities
The DOE facilities source category comprises sites that are
owned by the Federal government and operated by contractors under
the supervision of the DOE. The sites addressed in this chapter
are the active DOE sites that release significant quantities of
radionuclides to the air. These facilities and their locations
are listed in Table 2.1-1. These facilities are engaged in
numerous aspects of nuclear energy. They support the nation s
nuclear weapons capability by designing and producing nuclear
weapons for the Department of Defense (DOD). They support the
commercial nuclear power sector through enrichment of uranium and
nuclear reactor development and safety programs. They are also
involved in biomedical research, environmental safety, and
nuclear waste disposal programs.
The diversity of operations at these sites makes it
difficult to assess DOE facilities on a generic basis. The major
emissions from the facilities, however, are similar and consist
largely of inert gases such as argon-41, krypton-85, krypton-88,
and xenon-133. These gases are heavier than air and only
slightly soluble in water. Tritium, oxygen-15, uranium-234, and
uranium-238 are also commonly emitted.
A site-by-site discussion of each facility is presented in
the following sections along with an estimate of the doses and
risks associated with the current (1986) releases of
radionuclides to the atmosphere. Details of the inputs supplied
to the AIRDOS-EPA/DARTAB/RADRISK risk assessment computer codes
are presented for each site in Appendix A.
Historically, the Department of Energy has been self-
regulating with respect to environmental controls. Since the
1970's limits on releases of radioactive materials have roughly
paralleled those established by the Nuclear Regulatory Commission
(NRC) In 1985, the EPA promulgated a NESHAP for DOE facilities
(40 CFR 61, Subpart H) which limits radionuclide releases to air
from any DOE facility to quantities that do not cause nearby
individuals a dose greater than 25 mrem/y to the whole body or
75 mrem/y to any organ.
The summary tables in Section 2.1 and the individual
facility discussions incorporate source terms, stack heights,
meteorology, and other model parameters that reflect comments
received from DOE and the specific facilities. Model input
parameters are described in the AIRDOS input sheets presented in
the appendix. Draft version input sheets may be compared to
these sheets to determine changes in AIRDOS input parameters.
2-1
-------
Table 2.1-1. Department of Energy facilities.
Facility Location
Los Alamos National Laboratory
Oak Ridge Reservation
Savannah River Plant
Reactive Metals, Inc.
Feed Materials Production Center
Hanford Reservation
Brookhaven National Laboratory
Mound Facility
Idaho National Engineering Laboratory
Lawrence-Berkeley Laboratory
Paducah Gaseous Diffusion Plant
Lawrence Livermore/Sandia Laboratory
Portsmouth Gaseous Diffusion Plant
Argonne National Laboratory
Pinellas Plant
Nevada Test Site
Knolls Atomic Power Laboratory
Battelle Memorial Institute
Fermi National Accelerator Laboratory
Sandia National Laboratories/Lovelace
Bettis Atomic Power Laboratory
Knolls Atomic Power Laboratory
Rocky Flats Plant
Pantex Plant
Knolls Atomic Power Laboratory
Ames Laboratory
Rockwell International
Los Alamos, New Mexico
Oak Ridge, Tennessee
Aiken, South Carolina
Ashtabula, Ohio
Fernald, Ohio
Richland, Washington
Long Island, New York
Miamisburg, Ohio
Upper Snake River,Idaho
Berkeley, California
Paducah, Kentucky
Livermore, California
Piketon, Ohio
Argonne, Illinois
Pinellas County, Florida
Nye County, Nevada
Kesselring, New York
Columbus, Ohio
Batavia, Illinois
Albuquerque, New Mexico
West Mifflin,
Pennsylvania
Windsor, Connecticut
Jefferson Co., Colorado
Amarillo, Texas
Schenectady, New York
Ames, Iowa
Santa Susana, California
2.1.2 Summary of the Dose and Risk Assessment
The following tables present the tabulated results of the
risk assessment for 27 facilities in this source category. Table
2.1-2 shows the risk figures representing the highest cancer risk
to a selected individual. Table 2.1-3 presents the aggregate
risk distribution table for all DOE facilities. Table 2.1-4
presents the population exposures and total deaths per vear for
all DOE facilities. v y
Results for each site are also tabulated and presented in
the following sections.
2-2
-------
Table 2.1-2 Summary of doses and risks to nearby individuals
from DOE facilities due to 1986 emmissions.
Site
Primary 1986
Radio- Emissions
nuclide (Ci/y)
Organ Doses
(mrem/y)
Maximum
Individual
Risk
Los Alamos
Laboratory, NM
0-15
C-ll
N-13
8 . 6E+4
1.8E+4
4.8E+3
Gonads
Remainder
Breast
Lungs
Red marrow
9.5E+0
7.4E+0
8.9E+0
8.8E+0
7.0E+0
2E-4
Oak Ridge National
Lab., TN
U-234 1.5E-1
H-3 3.1E+4
U-238 2.8E-2
Lungs
Remainder
2.2E+1
2.0E+0
8E-5
Savannah River
Plant, GA
H-3
Ar-41
4.2E+5
8.3E+4
Remainder
Gonads
Breast
Lungs
Red marrow
3.2E+0
2.6E+0
2.6E+0
2.7E+0
2.6E+0
8E-5
Reactive Metals,
Inc., OH
U-234 5.6E-4 Lungs
U-238 5.3E-3
2.5E+1
4E-5
Feed Materials
Prod. Ctr., OH
U-234 2.0E-2 Lungs
U-238 2.0E-2
1.9E+1
3E-5
Hanford Reservation, Ar-41 1.3E+5
WA Pu-238 8.9E-2
Pu-239 3.1E-3
Lungs
Remainder
Gonads
Endosteum
2.8E+0
l.OE+0
1.1E+0
6.3E+0
3E-5
Brookhaven National Ar-41 1.2E+3
Lab., NY
Gonads
Remainder
Breast
Red marrow
Lungs
8.0E-1
6.2E-1
7.2E-1
6.2E-1
6.1E-1
2E-5
2-3
-------
Table 2.1-2
Summary of doses and risks to nearby individuals
from DOE facilities (continued).
Site
Mound Facility, OH
Idaho National Eng.
Lab., ID
Lawrence Berkeley
Lab. , CA
Primary 1986
Radio- Emissions
nuclide (Ci/y)
Organ Doses
(mrem/y)
Maximum
Individual
Risk
H-3 3.6E+3
Ar-41 1.9E+3
Sb-125 9.3E-1
Kr-88 1.6E+2
H-3 7.6E+1
Remainder
Gonads
Breast
Lungs
Red marrow
Gonads
Remainder
Breast
Lungs
Red marrow
Remainder
Gonads
Red marrow
Breast
Lungs
4.1E-2
3.7E-2
3.7E-2
3.8E-2
3.7E-2
2.9E-2
2.3E-2
2.7E-2
2.4E-2
2.3E-2
1.9E-2
1.8E-2
2.5E-2
1.8E-2
1.8E-2
1E-6
6E-7
5E-7
Paducah Gaseous U-234 1.8E-4
Diff. Plant, KY U-238 1.8E-4
Lungs
2.5E-1
4E-7
Lawrence Livermore
Lab., CA
H-3
1.8E+3
Remainder
Gonads
Breast
Lungs
Red marrow
1.1E-2
1.1E-2
1.1E-2
1.1E-2
1.1E-2
3E-7
Portsmouth Gaseous U-234 2.3E-2
Diff. Plant, OH U-238 1.4E-2
Endosteum
Remainder
Red marrow
3.4E-1
3.0E-2
2.3E-2
2E-7
Argonne National oil 9.0E+1
Lab., IL H-3 5.0E+1
Lungs
Remainder
3.1E-2
2.7E-3
1E-7
2-4
-------
Table 2.1-2
Site
Summary of doses and risks to nearby individuals
from DOE facilities (continued).
Primary 1986
Radio- Emissions
nuclide (Ci/y)
Organ Doses
(mrem/y)
Maximum
Individual
Risk
Pinellas Plant, FL
Nevada Test Site,
NV
Knolls Lab-
Kesselring, NY
Battelle Memorial
Inst., OH
Fermi National Lab.,
IL
Sandia National
Lab . -Lovelace , NM
Rocky Flats Plant,
CO
H-3
Xe-133
H-3
Ar-41
CO-60
C-14
K-40
U-235
Pu-239
C-ll
Ar-41
Pb-212
U-238
Am-241
1.9E+2
3 . 6E+4
1.2E+2
1.6E-1
3.4E-6
3.4E-1
3.0E-4
2.6E-6
4.0E-7
3.4E+0
5.5E+0
8.5E-3
1.7E-5
4.8E-6
Remainder
Gonads
Breast
Lungs
Red marrow
Gonads
Remainder
Breast
Thyroid
Remainder
Red marrow
Breast
Gonads
Lungs
Lungs
Gonads
Remainder
Breast
Gonads
Remainder
Breast
Lungs
Red marrow
Remainder
Gonads
Lungs
Breast
Red marrow
Lungs
Endosteum
Remainder
4.7E-3
4.4E-3
4.4E-3
4.4E-3
4.3E-3
5.3E-3
3.5E-3
6.5E-3
1.9E-2
3.8E-3
6.9E-3
4.4E-3
2.5E-3
2.5E-3
3.1E-3
8.7E-4
7.2E-4
7.8E-4
9.2E-4
7.1E-4
8.6E-4
9.1E-4
7.0E-4
5.3E-4
5.9E-4
1.2E-3
5.4E-4
5.6E-4
6.3E-3
1.6E-2
7.5E-4
1E-7
1E-7
1E-7
2E-8
2E-8
1E-8
1E-8
2-5
-------
Table 2.1-2 Summary of doses and risks to nearby individuals
from DOE facilities (continued).
Site
Primary 1986
Radio- Emissions
nuclide (Ci/y)
Organ Doses
(mrem/y)
Maximum
Individual
Risk
Bettis Atomic Power U-234 6.0E-7
Lab., PA U-238 6.0E-7
Sb-125 3.1E-5
Lungs
4.3E-3
1E-8
Knolls Lab-Windsor, Ar-41 7.8E-2
CT
Gonads
Remainder
Breast
Red marrow
Lungs
3.8E-4
3.0E-4
3.5E-4
3.0E-4
2.9E-4
8E-9
Pantex Plant, TX
U-238 l.OE-5 Lungs
2.2E-3
4E-9
Knolls Lab-Knolls, U-234 3.3E-6 Lungs
CT
1.7E-3
3E-9
Ames Laboratory, IA H-3 7.6E-2
Remainder
Gonads
Breast
Red marrow
Lungs
1.6E-5
1.3E-5
1.3E-5
1.3E-5
1.3E-5
4E-10
Rocketdyne Rockwell, Sr-90 1.3E-5
CA
Red marrow 7.OE-6
Endosteum 1.5E-5
2E-11
2-6
-------
Table 2.1-3. Distribution of fatal cancer risk in the population.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
590,
1,000,
65,000,
0
0
o*
2*
000
000
000
0
0
0
5E-6
2E-1
3E-2
1E-2
TOTALS 67,000,000 2E-1
EPA believes there are people at this risk at two facilities
(RMI, LASL). However, we cannot quantify the number because
a site visit has not been made.
2-7
-------
Table 2.1-4 Summary of doses and risks to the regional
population (0-80 km) around DOE facilities.
Site
0-80 km
Population
Population Organ
Exposure
(person-rem/y)
Deaths/y
Los Alamos 160,000
Laboratory, NM
Oak Ridge National 850,000
Lab . , TN
Savannah River 550,000
Plant, GA
Gonads
Remainder
Breast
Lungs
Red marrow
Lungs
Remainder
Remainder
Gonads
Breast
Lungs
Red marrow
l.OE+1
1.1E+1
9.7E+0
1.1E+1
9.2E+0
4.3E+2
7.8E+1
6.7E+2
5.5E+2
5.5E+2
5.6E+2
5.5E+2
4E-3
3E-2
2E-1
Reactive Metals,
Inc., OH
1,400,000
Lungs
3.2E+1
8E-4
Feed Materials
Prod. Ctr., OH
3,300,000
Lungs
1.1E+2
3E-3
Hanford Reservation, 350,000
WA
Lungs
Remainder
Gonads
Endosteum
5.6E+1
1.7E+1
1.5E+1
1.7E+2
6E-3
Brookhaven National 5,200,000
Lab., NY
Gonads
Remainder
Breast
Red marrow
Lungs
3.8E+0
3.0E+0
3.4E+0
2.9E+0
2.9E+0
1E-3
2-8
-------
Table 2.1-4
Summary of doses and risks to the regional
population (0-80 km) around DOE facilities
(continued).
Site
0-80 km
Population
Population Organ
Exposure
(person-rem/y)
Deaths/y
Mound Facility, OH 2,900,000
Remainder 3.3E+0
Gonads 3.OE+0
Breast 3.OE+0
Lungs 3.OE+0
Red marrow 3.OE+0
3E-3
Idaho National Eng. 100,000
Lab., ID
Gonads
Remainder
Breast
Lungs
Red marrow
7.3E-2
6.3E-2
6.8E-2
6.1E-2
5.7E-2
2E-5
Lawrence Berkeley 5,000,000
Lab., CA
Remainder
Gonads
Red marrow
Breast
Lungs
7.8E-1
7.0E-1
l.OE+0
7.0E-1
7.0E-1
3E-4
Paducah Gaseous
Diff. Plant, KY
500,000
Lungs
3.1E-1
1E-5
Lawrence Livermore 5,300,000
Lab., CA
Remainder
Gonads
Breast
Lungs
Red marrow
4.2E+0
3.7E+0
3.7E+0
3.8E+0
3.7E+0
1E-3
Portsmouth Gaseous 620,000
Diff. Plant, OH
Endosteum
Remainder
Red marrow
5.7E+0
7.7E-1
4.0E-1
9E-5
Argonne National
Lab., IL
7,900,000
Lungs
Remainder
2.5E-1
2.1E-1
8E-5
2-9
-------
Table 2.1-4
Summary of doses and risks to the regional
population (0-80 km) around DOE facilities
(continued).
Site
0-80 km
Population
Population Organ
Exposure
(person-rem/y)
Deaths/y
Pinellas Plant, FL 1,900,000
Remainder
Gonads
Breast
Lungs
Red marrow
5.3E-1
4.7E-1
4.7E-1
4.7E-1
4.7E-1
2E-4
Nevada Test Site,
NV
3,500
Gonads
Remainder
Breast
Thyroid
1.2E-2
8.1E-3
1.5E-2
5.7E-2
3E-6
Knolls Lab-
Kesselring, NY
1,200,000
Remainder
Red marrow
Breast
Gonads
Lungs
3.2E-2
6.5E-2
3.7E-2
1.5E-2
1.8E-2
2E-5
Battelle Memorial 1,900,000
Inst., OH
Lungs
Gonads
Remainder
Breast
1.5E-2
6.2E-3
5.2E-3
5.7E-3
3E-6
Fermi National Lab., 7,700,000
IL
Gonads
Remainder
Breast
Lungs
Red marrow
4.1E-3
3.2E-3
3.9E-3
4.1E-3
3.2E-3
1E-6
Sandia National 500,000
Lab.-Lovelace, NM
Remainder
Gonads
Lungs
Breast
Red marrow
1.9E-2
2.1E-2
4.9E-2
1.9E-2
2.1E-2
8E-6
Rocky Flats Plant,
CO
1,900,000
Lungs 1.2E-1
Endosteum 2.OE-1
Remainder 9.3E-3
9E-6
2-10
-------
Table 2.1-4 Summary of doses and risks to the regional
population (0-80 km) around DOE facilities
(continued).
Site
0-80 km
Population
Population Organ
Exposure
(person-rem/y)
Deaths/y
Bettis Atomic Power 3,100,000
Lab., PA
Lungs
3.5E-2
1E-6
Knolls Lab-Windsor, 3,200,000
CT
Gonads
Remainder
Breast
Red marrow
Lungs
2.3E-3
4.2E-3
4.9E-3
8.1E-3
2.5E-3
2E-6
Pantex Plant, TX
260,000
Lungs
3.5E-3
7E-8
Knolls Lab-Knolls,
CT
1,200,000
Lungs
3.1E-2
1E-6
Ames Laboratory, IA
680,000
Remainder
Gonads
Breast
Red marrow
Lungs
2.3E-4
1.8E-4
1.8E-4
1.8E-4
1.8E-4
9E-8
Rocketdyne Rockwell, 8,800,000
CA
Red marrow 1.4E-3
Endosteum 3.2E-3
7E-8
2-11
-------
2>1'3 Summary of the Supplementary Control Alternatives
The facilities chosen for discussion of supplemental control
alternatives are those that yielded an effective dose equivalent
of 1 mrem/yr or higher. These facilities are:
1. Oak Ridge Reservation
2. Los Alamos Scientific Laboratory
3. Savannah River Plant
4. FMPC
Current emission control technologies and detailed
discussions of supplemental control technologies at each of these
facilities are presented in Sections 2.2 through 2.7.
Alternative 1: baseline emissions
MIR: 2E-4
Incidence: 0.22
Impact: None
Alternative 2i emissions limited to 10 mrem/y EDE.
MIR: 8.1E-5
Incidence: 0.24
Impact, alternative 1 to alternative 2:
Incremental Capital Cost: $0
Incremental Annual Operating Cost: $0
Incremental Incidence Reduction: None
All DOE facilities have baseline emissions corresponding to
an EDE of 10 mrem/y or less. Therefore, Alternative 2 is
identical to Alternative 1.
Alternative 3; emissions limited to 3 mrem/y EDE.
MIR: 4E-5
Incidence: 0.22
Impact, alternative 2 to alternative 3:
Incremental Capital Cost: $5.9 million
Incremental Annual Operating Cost: $182,000
Incremental Incidence Reduction: 0.02
To reach this limit, supplemental emission controls would be
required at two DOE facilities: Oak Ridge National Laboratory and
Los Alamos National Laboratory.
At Oak Ridge, an additional stage HEPA filter and
high-energy Venturi scrubber, at an estimated capital cost of
$2,650,000, would reduce emissions of uranium-234 and uranium-238
from the Y-12 plant. In addition, a tritiated water sieve/dryer
system, at an estimated capital cost of $1,660,000, would reduce
emissions of tritium from ORNL. These emission reductions would
be sufficient to allow ORNL to reach the Alternative B limit.
2-12
-------
At Los Alamos, beam stop modifications and a delay tunnel
and new venting stack at the Meson Physics Facility would
sufficiently reduce emissions of oxygen-15, carbon-11, and
nitrogen-13, at a capital cost of $1,600,000.
Alternative 4; emissions limited to 1.0 mrem/y EDE.
MIR: 2.4E-5
Incidence: 0.094
Impact, alternative 3 to alternative 4:
Incremental Capital Cost: $134 million
Incremental Annual Operating Cost: $8,111,000
Incremental Incidence Reduction: 0.036
To reach Alternative 4, additional emission controls would
be required at RMI, Savannah River and FMPC.
For Savannah River, additional stage HEPA filters would be
required on the F and H stacks and in the P, X, and C reactor
areas, at an estimated capital cost of $130 million.
For FMPC, HEPA filters for Plants 4, 5, and 8 and additional
dust collector and scrubber stacks, at an estimated capital cost
of $4.2 million would be required.
2.1.4 Effect of Supplementary Control Alternatives
Tables 2.1-5 through 2.1-7 present the risk distributions
for the population at risk fop the DOE facilities. Table 2.1-5
presents the risk distribution for the baseline case, which
assumes 1986 emissions with no supplemental control strategies
implemented. Table 2.1-6 presents the risk distribution for
Alternative 3, which assumes that supplemental controls have been
applied to ensure that an effective dose equivalent to nearby
individuals would be no more than 3 mrem/y at any of the DOE
facilities. Table 2.1-7 presents the risk distribution for
Alternative 4, which assumes that supplemental controls have been
applied to ensure that an effective dose equivalent to nearby
individuals would be no more than 1 mrem/y at any of the DOE
facilities.
The maximum individual risks, assuming implementation of
Alternative 4 supplemental control strategies, are presented in
Table 2.1-8.
The number of deaths per year, assuming implementation of
Alternative 4 supplemental control strategies, are presented in
Table 2.1-9.
2-13
-------
Table 2.1-5. Baseline risk assessment for DOE facilities.
Highest Lifetime Individual Fatal Cancer Risk: 1E-04
Population Risk (those within 80 km): 0.2
Distribution of Fatal Cancer Risk in Populations Within 80 km:
Risk interval Number of persons Deaths/y
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
-------
Table 2.1-7. Risks when emmissions are limited to 1 mrem/y EDE,
Highest Lifetime Individual Fatal Cancer Risk: 2E-05
Population Risk (those within 80 km): 0.09
Distribution of Fatal Cancer Risk in Populations Within 80 km:
Risk interval Number of persons Deaths/y
1E-2 to 1E-1 0 0
1E-3 to 1E-2 0 0
1E-4 to 1E-3 0 0
1E-5 to 1E-4 250,000 4E-2
1E-6 to 1E-5 540,000 4E-2
-------
Table 2.1-8,
Maximum individual risk, with Alternative 4
supplemental control strategies.
Site
Hanford Reservation, WA
Savannah River Plant, GA
Oak Ridge National Lab., TN
Brookhaven National Lab., NY
Los Alamos Laboratory, NM
Reactive Metals, Inc., OH
Feed Materials Prod. Ctr. , OH
Mound Facility, OH
Idaho National Eng. Lab., ID
Lawrence Berkeley Lab., CA
* With supplemental emission
** Nearby generic individual
Primary
Radio-
nuclide
Ar-41
Pu-238
Pu-239
H-3
Ar-41
U-234
H-3
U-238
Ar-41
0-15
C-ll
N-13
U-234
U-238
U-234
U-238
H-3
Ar-41
Sb-125
Kr-88
H-3
controls.
from population
1986
Emissions
(Ci/y)*
1.3E+5
8.9E-2
3.1E-3
4.2E+5
8.3E+4
1.5E-1
3 . 1E+4
2.8E-2
1.2E+3
8 . 6E+4
1.8E+4
4.8E+3
5.6E-4
5.3E-3
2.0E-2
2.0E-2
3 . 6E+3
1.8E+3
9.3E-1
1.4E+2
7.6E+1
run.
Maximum
Individual
Risk**
3E-5
2E-5
2E-5
2E-5
2E-5
1E-5
1E-5
1E-6
6E-7
5E-7
2-16
-------
Table 2.1-8,
Maximum individual risk, with Alternative 4
supplemental control strategies (continued).
Site
Paducah Gaseous Diff. Plant,
KY
Lawrence Livermore Lab . , CA
Primary
Radio-
nuclide
U-234
U-238
H-3
Portsmouth Gaseous Diff. Plant, U-234
OH U-238
Argonne National Lab., IL
Pinellas Plant, FL
Nevada Test Site, NV
Knolls Lab-Kesselring, NY
Battelle Memorial Inst., OH
Fermi National Lab., IL
Sandia National Lab. -Lovelace,
NM
Rocky Flats Plant, CO
* With supplemental emission
** Nearby generic individual
C-ll
H-3
H-3
Xe-133
H-3
Ar-41
CO-60
C-14
K-40
U-235
PU-239
C-ll
Ar-41
Pb-212
1986
Emissions
(Ci/y)*
1.8E-4
1.8E-4
2 . OE+3
2.8E-2
l.OE-2
9.0E+1
5.0E+1
1.9E+2
3 . 6E+4
1.2E+2
1.6E-1
3.4E-6
3.4E-1
3.0E-4
2.6E-6
4.0E-7
3.4E+0
5.5E+0
8.5E-3
U-238 1.7E-5
Am-241 4.8E-6
controls.
from population run.
Maximum
Individual
Risk**
4E-7
3E-7
2E-7
1E-7
1E-7
1E-7
1E-7
2E-8
2E-8
1E-8
1E-8
2-17
-------
Table 2.1-8,
Maximum individual risk, with Alternative 4
supplemental control strategies (continued).
Site
Bettis Atomic Power Lab., PA
Knolls Lab-Windsor, CT
Pant ex Plant, TX
Knolls Lab-Knolls, CT
Ames Laboratory, IA
Rocketdyne Rockwell, CA
* With supplemental emission
** Nearby generic individual
Primary
1986
Radio- Emissions
nuclide (Ci/y)*
U-234
U-238
Sb-125
Ar-41
U-238
U-234
H-3
Sr-90
controls.
from population
6.0E-7
6.0E-7
3.2E-5
7.8E-2
l.OE-5
3.3E-6
7.6E-2
1.3E-5
run.
Maximum
Individual
Risk**
1E-8
8E-9
4E-9
3E-9
4E-10
2E-11
2-18
-------
Table 2.1-9. Fatal cancers/year to nearby individuals, with
Alternative 4 supplemental control technologies.
0-80 km «
site Population Deaths/y
Los Alamos Laboratory, NM
Oak Ridge National Lab., TN
Savannah River Plant, GA
Reactive Metals, Inc., OH
Feed Materials Prod. Ctr. , OH
Hanford Reservation, WA
160,000
550,000
550,000
1,400,000
3,300,000
350,000
2E-3
7E-3
8E-2
7E-5
9E-4
6E-3
Brookhaven National Lab., NY 5,200,000 1E-3
Mound Facility, OH 2,900,000 3E-3
Idaho National Eng. Lab., ID 100,000 2E-5
Lawrence Berkeley Lab., CA 5,000,000 3E-4
Paducah Gaseous Diff. Plant, KY 500,000 1E-5
Lawrence Livermore Lab., CA 5,300,000 1E-3
Portsmouth Gaseous Diff. Plant, OH 620,000 9E-5
Argonne National Lab., IL 7,900,000 8E-5
Pinellas Plant, FL 1,900,000 2E-4
* In population within 80 km.
2-19
-------
Table 2.1-9. Fatal cancers/year to nearby individuals, with
4 SU^lemental control technologies
Site °~80
Population Deaths/y*
Nevada Test Site, NV
Knolls Lab-Kesselring, NY
Battelle Memorial Inst., OH
Fermi National Lab. , IL
Sandia National Lab. -Lovelace, NM
Rocky Flats Plant, CO
Bettis Atomic Power Lab., PA
Knolls Lab-Windsor, CT
Pantex Plant, TX
Knolls Lab-Knolls, CT
Ames Laboratory, IA
Rocketdyne Rockwell, CA
* In population within 80 km.
• .
3,500
1,200,000
1,900,000
7,700,000
500,000
1,900,000
3,100,000
3,200,000
260,000
1,200,000
680,000
8,800,000
2E-5
3E-6
IE-6
8E-6
9E-6
IE-6
2E-6
7E-8
IE-6
9E-8
7E-8
2-20
-------
Table 2.1-10. Distribution of fatal cancer risk in the
populations within 80 km with Alternative 4
supplemental control technologies.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
250,
540,
66,000,
0
0
0
0
000
000
000
0
0
0
0
4E-2
4E-2
1E-2
Totals 67,000,000 1E-1
2.2 RMI COMPANY
2.2.1 Description and Existing Controls
2.2.1.1 Site Description
RMI Company (RMI), formerly Reactive Metals, Inc., 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 1980 U.S. Census, the
population within 80 km of the facility is about 1.4 million.
2.2.1.2 Major Release Points,and Existing Emission
Control Technology
RMI 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. The RMI facility also conducts activities an an
NRC licensee. Releases from both DOE and NRC activities are
included in this assessment.
2.2.2 Basis for the Dose and Risk Assessment
2.2.2.1 Source Terms and Release Point Characterization
The only radioactive material released to the air from RMI
is insoluble natural uranium. The total airborne releases, in
Ci/y, from all sources during 1986 are listed below in Table
2.2-1.
2-21
-------
Table 2.2-1. Radionuclides released to air during 1986 from RMI.*
Nuclide Release Rate (Ci/y)
U-234 5.6E-4
U-235 4.4E-5
U-238 5.3E-3
* Ajusted, see text,
Releases from the RMI plant consist of natural, depleted,
and slightly enriched uranium. During 1986, the year for which
the assessment is made, control technology upgrades consisting of
HEPA filters were begun at RMI. These upgrades were completed on
stack 4 during 1986 and reduced the emissions for that stack from
approximately 12,000 juCi for the first half of the year to
0.06 /xCi during the second half. The emissions shown in Table
2.2-1 were used to assess the risk. They reflect the emissions
during 1986 adjusted to account for the addition of HEPA filters
on stack 4. Continued upgrades of the effluent controls during
1987, 1988, and the discontinuation of stacks without HEPA
filtration have further reduced emissions. In 1988, RMI reports
a total uranium release of 7E-4 Ci/y, approximately a factor of
10 lower than the source term used in this assessment (RMI89).
To evaluate the health impact from the operation of RMI,
releases from the facility were assumed to be from six stacks
with heights given in the Appendix. The released uranium-234 was
assumed to be in equilibrium with its daughters thorium-234 and
protactinium-234m. Default particle sizes (1.00 AMAD) and
solubility class Y were assumed based on information from RMI
(RMI89).
2.2.2.2 Other Parameters Used in the Assessment
The nearest individual was assumed to be located 310 m from
the release point (RMI86).
Meteorological data used in the assessment are from Erie,
Pennsylvania. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data. Food
consumption rates appropriate to an urban location were used.
2.2.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-234 (52
percent) and uranium-238 (46 percent). The predominant exposure
pathway is inhalation for uranium-234 and uranium-238.
The results of the dose and risk assessment are presented in
Tables 2.2-2 through 2.2-4. Table 2.2-2 presents the doses
2-22
-------
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.2-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.2-4 presents the estimated distribution of fatal cancer risk to
the regional population.
Table 2.2-2. Estimated radiation dose rates from RMI.
Organ
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
2.5E+1
3.2E+1
Table 2.2-3. Estimated fatal cancer risks from RMI.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
4E-5
8E-4
Table 2.2-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from RMI.
Totals
Number of Persons
1,400,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
to
to
to
to
to
to
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
1
98,000
1,400,000
0
0
0
0
6E-7
2E-4
5E-4
8E-4
2.2.4 Supplementary Controls
As noted in Section 2.2.2.1, RMI has recently completed the
upgrade of its effluent control system which was begun in 1986.
This has consisted of addition of HEPA filters on stacks 1 and 4
and the discontinuation of unfiltered stacks.
2-23
-------
2.2.4.1 Emission Reduction
The upgrade of the effluent control system has resulted in a
reduction of uranium emissions. During 1986, when only stack 4
was retro-fittee for nalf the year, total uranium releases were
1.7E-2 Ci/y. During 1988, with the upgrade complete, total
uranium releases were 7E-4 Ci/y, a reduction of 96 percent.
2.2.4.2 Costs of Supplementary Controls
... N° ^ta were provided by RMI on the costs of the additional
effluent controls. Further reductions could be achieved bv
placing additional HEPA filters in series. No estimates of the
costs or efficiencies of such additional controls have been made.
2.3 LOS ALAMOS NATIONAL LABORATORY
2-3.1 Description and Existing Controls
2.3.1.1 Site Description
Los Alamos National Laboratory is one of the prime research
and development centers for DOE's nuclear weapons program. This
facility is located about 100 km north-northeast of Albuquerque
New Mexico. In addition to defense-related activities, programs
include research in the physical sciences, energy resources
environ- mental studies, and biomedical applications of '
radiation.
Radionuclides are released from 13 technical areas at this
site. These areas contain research reactors that produce
materials for use in high-temperature chemistry applications
weapons systems development, nuclear safety program development
accelerator operations, biomedical research, development of
isotope separation processes, and waste disposal.
2.3.1.2 Major Release Points and Existing Emission
Control Technology
The following sections describe the emission control
technology currently in use at the six sources being evaluated
Possible application of additional control technology, the
effects of such improvements on discharge rates, and the costs of
such improvements are also discussed. Generic information on the
emission control technology for the nonspecific or minor sources
is also provided (Mo86).
2.3.1.2.1 Omeoa West Reactor Stack
The Omega West research reactor, located in TA-2, is used
for a wide variety of experimental programs. The reactor is a
heterogeneous water-cooled tank-type reactor, with a maximum
power level of 8 MWth.
2-24
-------
Argon-41 (t 1/2 =1.8 hr) was the only radionuclide above the
limits of detectability released to the atmosphere from the Omega
West reactor stack in 1986. The argon-41 is produced by neutron
activation of the natural argon in air. Process air streams and
part of the building ventilation exhaust are discharged to the
atmosphere from the reactor stack, which is located about 300 m
from the reactor. The total air flow to the stack is about
28.3 m3/roin. The stack is approximately 0.2 m in diameter, and
its height is approximately 46 m above ground level. The stack
is continuously monitored. Charcoal cartridges are installed in
the process air stream to remove any radioiodine present. There
is no technology in place to remove argon-41 from the air stream
flowing to the stack. Some reduction in the argon-41 level is
provided by delay (approximately one hour) as the air flows from
the reactor building to the stack.
2.3.1.2.2 LAMPF Main Stack
The Clinton P. Anderson Los Alamos Meson Physics Facility
(LAMPF) in TA-53 consists primarily of a linear proton
accelerator, approximately 800 m long, designed to produce an
800 MeV proton beam with an average intensity of one milliampere.
The proton beam and secondary particles produced when the
energetic protons strike a target are used in a wide variety of
experimental programs. Fields of investigation include medium
energy nuclear physics, biophysics, radiochemistry, and cancer
therapy.
Interaction of the proton beam and secondary particles with
air produces several activation products. These activation
products, which include beryllium-7, carbon-11, nitrogen-13,
oxygen-15, argon-41, and tritium, were the only radionuclides
released to the atmosphere from the LAMPF facility in 1986. The
activation products are discharged to the atmosphere from the
LAMPF main stack. The main stack receives the air flow from a
single fan exhaust system. Air flow to the main stack is about
480 m3/roin. The stack has a diameter which varies from about
1.5 m to 0.9 m at the top. The stack height is about 30.5 m
above ground level.
Air flowing to the LAMPF stack is passed through a single
stage of HEPA filtration to remove particulates. There is no
technology in place to remove gaseous radionuclides from the air
stream. Areas where the air activation products are produced are
continuously ventilated to remove the radionuclides as they are
formed. Due to the short half-lives of some of the activation
products formed, some reduction in the radionuclide release is
obtained by decay due to holdup as the air flows from the various
source points to the stack. The extent of the reduction will
depend on the radionuclides. In the case of oxygen-15
(t 1/2 =2.0 min), the holdup could reduce the release
significantly. In the case of tritium (t 1/2 =12.3 yr) and
2-25
-------
beryllium-? (t 1/2 =53.3 days), the holdup would have
essentially no effect on the releases.
2.3.1.2.3 Stack FE-6-HP Site
The tritium handling facility is located at the HP site
(TA-33). A wide variety of experimental programs involving the
use of tritium is carried out at the facility. Large amounts of
tritium are released to the atmosphere from the facility stack
(FE-6). A single fan-exhaust system is used to ventilate the
facility and feeds to the FE-6 stack. More than 84 percent of
the tritium discharged to the atmosphere at LANL is released from
Stack FE-6.
The average air flow to the stack is about 200 m3/min. The
stack is 0.61 m in diameter, and the height above ground level is
about 23 m.
The tritium handling facility is scheduled to be replaced in
several years. Physical containment of the tritium during
experimental activities is the principal method for controlling
tritium emissions from the tritium handling facility stack. Work
areas are ventilated to maintain the tritium concentration, due
to leaks, below the concentration guide for controlled areas. A
dryer system is used to remove tritiated water from the air
flowing to the stack.
2.3.1.2.4 South Stack-Wing 3 - CMR
The Chemistry Metallurgy Research Building (CMR) located in
TA-3 is a large multiwinged building in which a wide variety of
research programs is carried out. Each wing of the facility is
equipped with one or two stacks to handle the wing's air flow.
Small amounts of radionuclides are discharged to the atmosphere
from most of the building stacks. Wing 3 houses a variety of
analytical chemistry groups which provide services for the entire
laboratory. Approximately 55 percent of the plutonium released
to the atmosphere at LANL in 1986 was discharged from the south
stack of Wing 3 of the facility. No other radionuclides were
detected in the stack air flow in 1986.
The air flow to the stack comes from a single fan and
exhaust system (FE-19) serving a number of laboratories. The air
flow to the stack is about 1,400 m3/min. The stack has a
diameter of about 1 m, and the height above ground level is about
17 m. The air flowing to the south stack of Wing 3 of the
Chemistry Metallurgy Research Building is passed through a
two-stage prefilter and a single-stage bag filter prior to
discharge from the stack. It is estimated that the filter system
removes 90 to 95 percent of the particulates.
2-26
-------
2.3.1.2.5 Main Stack - Building 3-DP Site
Building 3 at the DP site (TA-21) is used for enriched
uranium recovery operations. Small amounts of uranium are
discharged to the atmosphere from several stacks used to
ventilate the building. Uranium-235 released from the main stack
of the building accounted for about 55 percent of the total
uranium released to the atmosphere at LANL in 1986. The chemical
form of the uranium released from the stack is unknown. No other
radionuclides were detected in the air leaving the stack.
The main building stack serves to ventilate building work
areas using a single fan-exhaust system (FE-1). Air flow to the
stack is 480 m3/ndn. The stack is about 1 m in diameter and
about 15 m above ground level. There is no equipment in place to
reduce emissions from the main stack of Building 3, except for
local HEPA filters in gloveboxes.
2.3.1.2.6 Core Wina Stack Radiochemistry Site
The radiochemistry site in TA-48 is used for a variety of
programs involving radioactive materials. Laboratory hoods,
glove boxes, and "hot cells" are used to contain the radioactive
materials. Small quantities of radioactive materials are
released to the atmosphere from several stacks at the facility.
About 87 percent of the mixed fission products (MFP) released to
the atmosphere at LANL in 1986 were released from the Core Wing
Stack, which is one of the stacks used to ventilate the
radiochemistry facility.
Two fan-exhaust systems (FE-45 and FE-46) discharge into the
Core Wing Stack. A number of glove boxes are serviced by the two
fan-exhaust systems. Total air flow to the stack is about
1,400 m3/rain, with the air flow almost equally divided between
the two fan-exhaust systems. The Core Wing Stack has a diameter
of about 1.5 m and a height of approximately 21.3 m above ground
level. The glove boxes which discharge to the two fan-exhaust
systems serving the Core Wing Stack are provided with a single
stage of HEPA filters.
2.3.2 Basis for the Dose and Risk Assessment
2.3.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed below in Table 2.3-1.
In modeling the site, all releases were assumed to be made
from the LAMPF, since this is the major source of dose. The
releases were assumed from a 30.5-m stack. Default particle
sizes (1.00 Amad) and solubility classes (Class D for carbon-11,
nitrogen-13, and oxygen-15) were assumed.
2-27
-------
Table 2.3-1. Radionuclides released to air during 1986 from
Los Alamos Scientific Laboratory.
Nuclide Release Rate (Ci/y)
Ar-41
C-ll
H-3
1-131
N-13
0-14
O-15
P-32
Pu-238
Pu-239
Sr-90
U-235
U-238
7.3E+2
1.8E+4
1.1E+4
3.8E-5
4.8E+3
2.6E+3
8 . 6E+4
7.0E-5
9.9E-5
1.1E-4
2.6E-3
7.1E-4
1.4E-4
2.3.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Santa
Fe, New Mexico. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data.
Nearby individuals were located 750 m from the assumed release
point (Em87). Food consumption rates appropriate to an urban
location were used.
2.3.3 Results of the Dose and Risk Assessment
The major contributors to exposure are oxygen-15
(57 percent), carbon-11 (29 percent), and nitrogen-13
(7 percent). The predominant exposure pathway is air immersion.
The results of the dose and risk assessment are presented in
Tables 2.3-2 through 2.3-4. Table 2.3-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.3-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.3-4 presents the estimated distribution of fatal cancer risk to
the regional population.
2-28
-------
Table 2.3-2
Organ
Estimated radiation doses from the Los Alamos
Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Gonads
Remainder
Breast
Lungs
Red marrow
9.5E+0
7.4E+0
8.9E+0
8.8E+0
7.0E+0
l.OE+1
1.1E+1
9.7E+0
1.1E+1
9.2E+0
Table 2.3-3. Estimated fatal cancer risks from the Los Alamos
Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
2E-4
4E-3
Table 2.3-4.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Los
Alamos Scientific Laboratory.
Risk Interval
Totals
Number of Persons
160,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
2,
100,
50,
0
0
0
1
500
000
000
0
0
0
3E-6
9E-4
2E-3
7E-4
4E-3
2.3.4 Supplementary Controls
2.3.4.1 LAMPF Main Stack
The results of the dose and risk assessment show that
98 percent of the dose is due to emissions of oxygen-15,
carbon-11, and nitrogen-13, short-lived air activation products
from the LAMPF Main Stack.
2-29
-------
A permanent committee was formed at LANL several years ago
to review LAMPF operations (Em87, Mo85). One objective of the
committee is to evaluate potential methods for reducing releases
of airborne radioactivity from LAMPF operations. One plan
currently under consideration is to enclose one of the primary
beam stop areas, which is a major producer of air activation
products. The enclosed area would not be vented during
accelerator operation. Venting would be done only after the
accelerator shuts down and the short-lived radioisotopes have had
a chance to decay. The overall effectiveness of the proposed
modification for reducing airborne emissions from LAMPF has not
been determined. If the plan is implemented, construction of the
enclosure will start within two years (Mo86).
The large air flow to the LAMPF main stack (about
480 m /min) makes it very difficult to use any existing
technology to remove the gaseous activation products from the air
stream. The most realistic approach would be to provide
additional holdup time to allow some decay of the short
half-lived radionuclides, as indicated above. Extremely large
air storage volumes would be required to reduce radionuclide
releases significantly. For example, if an atmospheric pressure
air storage system having a storage volume of 9,300 m3 were
applied to the air flowing to the LAMPF stack, the additional
holdup time provided would be about 19.4 minutes. Table 2.3-5
presents the reductions in radionuclide emissions as a function
of holdup time.
Table 2.3-5. Effect of holdup time on the release of air
activation products from the proposed stack
serving the LAMPF beam stop.
Fraction of the Radionuclide Generated at the
Beam Stop Released to the Atmosphere
Single Tank Dual Tank
(20 min. additional (40 min. additional
Radionuclide holdup time) holdup time)
Oxygen-15
Carbon-11
Nitrogen-13
0.00108
0.505
0.25
1.18E-6
0.255
0.0625
As a result, total emissions from the stack would be reduced
from about 109,000 Ci/y to about 5,000 Ci/y at the same level of
programmatic activities.
The air storage tank would be constructed of carbon steel
and located on a concrete pad adjacent to the LAMPF stack,
2-30
-------
high.
estimated capital cost for an atmospheric pressure ar
r year The capital cost of air storage systems of
of the size ratio (Mo86) .
2.3.4.2 Omega West Reactor Stack
The Omega West research reactor, located in TA-2, is a
-0
NRC standards for research reactors
ss ck
The tota! air flow to the stack is about 28.3 m3/min. The stack
is aDoroximately 0.2 m in diameter, and its height is
approximate^ 46 m above ground level. The stack is continuously
monitored.
The argon-41 released from the reactor stack can be reduced
SecaT^^^
<-s url s^rage 55S2 ^^^^^^X^^S in
the argon-41 emissions at a normal airflow of 28.3 ^n are
given in Table 2.3-5. The use of a pressurized air storage
lystem would reduce the storage volume required for a given
decontamination factor (DF) but would probably increase the
overall cost of the system.
2.3.4.3 Stack FE-6-HP Site
The tritium handling facility is located at the HP site
fTA-33) A wide variety of experimental programs involving the
use of tritium is carried out at the facility °
u k
tritium are released to the atmosphere from the facility s£ack
(FE-6) . A single fan-exhaust system is used to ventilate the
facility and feeds to the FE-6 stack. More than 84 percent of
?he tritium discharged to the atmosphere at LANL is released from
Stack FE-6.
2-31
-------
The average air flow to the stack is about 200 mVmin. The
aboSt 23 m.'61 m ^ diameter' and the hei(?ht ab°ve ground level is
several6 trltium nandlin9 facility is scheduled to be replaced in
<- •4--Tl?eJchemical for* °f the tritium is unknown, but since any
tritiated water should be removed by the dryer, the tritium is
probably present as molecular hydrogen.
The large volume of air flowing to Stack FE-6 and the very
low concentration of tritium in the air make effective reduction
? Sf •tritium released from the stack both difficult and costly.
In addition, because the tritium handling facility is to be
replaced in a few years, it is difficult to justify large
expenditures for additional emission control technology.
Assuming the tritium is present in the air stream primarily
as molecular hydrogen, adequate removal of the tritium from the
air would require its conversion to water. A drying step would
then be required to remove the tritiated water from the air prior
to discharge. Subsequent recovery of the tritiated water from
the dryer and its final disposal would present additional
problems. A risk analysis would have to be carried out to
determine if disposal of the tritiated water would present less
of a risk then release of the tritium, as molecular hydrogen to
the atmosphere.
If removal of tritium from the air flowing to Stack FE-6
becomes necessary, a recovery system similar to the emergency
tritium cleanup system (ETC) which is used at the Tritium Systems
Test Assembly (TSTA) at LANL could probably be used. The ETC
system is designed to process air at the rate of about 39 m3/min.
Therefore, a similar system for Stack FE-6 would have to be
designed for air flow about five times as large (200 m3/min)
The ETC system was not intended for continuous operations, but
only for emergency use. However, the system could probably be
designed for continuous use.
2.3.4.4 South Stack-Wing 3 - CMR
The Chemistry Metallurgy Research (CMR) Building located in
TA-3 is a large multiwinged building, housing a wide variety of
research programs. Each wing of the facility is equipped with
one or two stacks to handle the wing air flow. Small amounts of
radionuclides are discharged to the atmosphere from most of the
building stacks. Wing 3 houses a variety of analytical chemistry
groups which provide services for the entire laboratory. The
air flow to the stack comes from a single fan and exhaust system
(FE-19) serving a number of laboratories. The air flow to the
stack is about 1,400 mymin. The stack has a diameter of about
1 m, and the height above ground level is about 17 m.
2-32
-------
The chemical form and isotopic composition of the plutonium
discharged are unknown.
Because the amount of plutonium released from the stack in
question and its effect on the environment are already very
small, additional equipment to reduce the plutonium release
probably would result in only slight decreases in the total risks
due to this facility. If additional reductions are necessary,
however, they could be attained by installing a HEPA filter
system in addition to or in place of the existing bag filter
system. A bank of at least 48 HEPA filters, measuring 61 cmx
61 cm x 30 cm would be needed to handle the air flow. The HEPA
filter system would provide at least a 99 percent reduction in
the plutonium release from the stack.
2.3.4.5 Main Stack - Building 3-DP Site
Building 3 at the DP site (TA-21) is used for enriched
uranium recovery operations. Small amounts of uranium are
discharged to the atmosphere from several stacks used to
ventilate the building. Uranium-235 released from the main stack
of the building accounted for about 34 percent of the total
uranium released to the atmosphere at LANL in 1986. The chemical
form of the uranium released from the stack is unknown. No other
radionuclides were detected in the air leaving the stack.
The main building stack serves to ventilate building work
areas using a single fan-exhaust system (FE-1). Air flow to the
stack is 480 mVmin. The stack is about 1 m in diameter, and the
height of the stack is about 15 m above ground level.
The amount of uranium released from the main stack of
Building 3 is already very small, and its effect on the
environment is minimal. If reductions become necessary, however,
a filter system could probably be installed. A HEPA filter
system would be preferred. A bank of at least 18 HEPA filters
measuring 61 cm x 61 cm x 30 cm would be required to handle the
air flow to the stack.
Installation of a HEPA filter system would provide at least
a 99.9 percent reduction in the uranium release from the stack.
The system would consist of three modules, each rated at
250 mymin, with two modules in operation and one module in
standby. Each module would consist of nine HEPA filters, two
dampers, and one 300 nr/min blower.
2.4 HANFORD RESERVATION
The Hanford Reservation was established in 1943 as a
Plutonium production facility for nuclear armaments. Information
used to evaluate the facility was obtained from DOE and Hanford
reports (Mo84, PNL87). Plutonium production has decreased, and
other programs have filled the gap, such as management and
storage of radioactive wastes, reactor operations, fuel
2-33
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ea^^
of each area are described briefly. activities
2.4.1.1 100 Area
for wMnh1^ — J contains the nine plutonium production reactors
for which the site was originally developed. Eight of these
reactors are currently shut-down. Operating facilities durtng
1986 include the N-Reactor and the 1706 Laboratory, which
provides support services for the reactor. N-Reactor has
subsequently been shut-down pending the resolution of safety
concerns . •*
2.4.1.2 200 Area
Activities conducted in the 200 Area include fuel
processing nuclear waste treatment and storage, equipment
decontamination, and research. Plutonium reclamation from spent
fuel is performed at the PUREX Plant in this area
2.4.1.3 300 Area
The major facilities in the 300 Area are the Hanford
Engineering Development Laboratory, the fuel fabrication
facility, and the Life Sciences Laboratory. The Hanford
Engineering Development Laboratory, the largest operation in this
area, supports all activities of the development program tor the
fast breeder reactor. Life science research in this area
includes plutonium inhalation studies and other programs
investigating the physiological effects of radioactive materials.
2.4.1.4 400 Area
4-v, J"1? °?ly facility currently in operation in the 400 Area is
the Fast Flux Test Facility. When the Fuel Materials Examination
Facility currently under construction is completed, the 400 Area
will be the center of the Hanford breeder reactor research
program.
2-4-2 Maior Release Points and Existing Emission
Control Technology
2.4.2.1 Stack 116-N Serving the 105-N Reactor
Building
Argon-41, which constitutes the primary airborne radioactive
emission from N-Reactor, is produced from the leakage of air into
the reactor system and subsequent activation of the stable arqon
in the air. Noble gases and volatile fission products, such as
xenon-133 and iodine-131, come from leaks in fuel element
claddings. Nonvolatile particulate fission and activation
2-34
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products, such as cobalt-60, europium-154, and molybdenum-99,
become airborne as a result of the primary coolant contacting
exposed surfaces, then drying and becoming suspended in air
currents.
The ventilation systems in 105-N are separated into five
zones based on their potential for contamination with airborne
radioactive material. The 116-N stack is the main discharge
coint for airborne radioactive material from N-Reactor.
?mmedia?ely preceding the 116-N stack is the 117-N filter and
diversion facility through which the exhaust air is routed prior
^release to the stack. The stack exhausts to the atmosphere
61 m above ground level.
The 117-N facility contains four separate air filtration
cells. The air from Zones I, II, and III of the 105-N building
enters through three separate ducts. Air from Zone I Passes
through two filtration cells, air from Zone II passes through a
third filtration cell, and air from Zone III normally bypasses
the filter cells as it is routed through the facility. In the
event of an emergency, however, Zone III exhaust can be combined
with Zone II exhaust to provide filtration for Zone III exhaust.
The fourth filtration cell is on standby for emergency backup.
The first, second, and fourth filtration cells are composed
of a series of three filter bank stages. The first stage is an
aluminum mesh screen used as a moisture separator to protect the
remaining filters in the event of entrained moisture in the air
stream. The second stage is a high-efficiency particulate air
(HEPA) filter. Minimum efficiency for removal of particulate
matter larger than 0.3 microns is 99.97 percent. These filters
are routinely tested for efficiency. The third stage contains
granular activated charcoal which removes 95 percent of the
inorganic halogen gases in the air stream.
The third filtration cell contains two stages, a HEPA filter
and an activated charcoal absorber.
Zones IV and V serve offices, administration areas, and the
reactor control room. Ventilation air from these areas is
exhausted through roof exhausters without treatment.
2.4.2.2 PUREX Main Stack No. 291-A-l
The four sources of gases that exhaust through the 61-m-high
291-A-l main stack of the Hanford PUREX facility are: the declad
and dissolver off-gas system, the process off-gas system, the
Plutonium oxide conversion facility off-gas system, and the
canyon ventilation system.
2.4.2.2.1 Declad and Dissolver Off-Gas System
The PUREX facility has the capability to process irradiated
fuel to separate and recover plutonium, uranium, and neptunium.
2-35
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In the head-end of the process, the cladding is chemicallv
removed from the fuel elements and the fuel is" tn^ dissolved in
The decontamination factor (DF) for the silver reactor
averages 100. The cell B silver reactor has a iTJI-J deep
packing bed of 1.3 cm ceramic saddles, while the cells A Snd C
of I60 3?-m de! hbed o °'88~m ^^ b °f I'3~°m "^les on top
T-JI ^V* ^ T T * j- ^ t*\*t\,Axco • J.X1S So.ClClJ.GS 3IT6 OO3u6Cl
witn silver nitrate. Iodine-129 and iodine-131 are removed in
?llv®r reactor- When the efficiency falls, the silver
?£!!/£»ve2enera.Ked Wltl? fresh Silver nitrate solution
then baked on the packing. When a reactor becomes
ground 1S replaced and sent to a low-level waste burial
^o KrSm ^he Si^ver react°r, the declad gases pass through two
deep-bed glass fiber filters in series. The gases are then
exhausted through the main stack, 291-A-l.
™^ DUmjng the diss°lution step, the gases follow a similar
path. The ammonia scrubber does not operate during dissolution
,S? rS^iS"*1"9 th? S8COnd glass fiber filter ar« routSd to Sie
90 ^J^J 5^in WhlCl> ^W° acid abs°rberS in series remove
90 percent of the remaining iodine and 90-92 percent of the
1 Sent to the
Krypton-85 is a major radionuclide released during the
declad and dissolving processes. There is no cleanup of
krypton-85 at PUREX. p
2.4.2.2.2 Process Off-Gas System
The PUREX process produces off-gases from condensers and
combined and routed
heater to 160 °C and pass through a silver reactor that removes
radioactive iodine that remained in solution during the fuel
dissolving process and that evolves during processing steps
This silver reactor has a very low efficiency. From the silver
reactor, the gases pass through a deep-bed glass fiber filter and
from there to the ventilation system No. 1 air tunnel
2-36
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2.4.2.2.3 Plutonium Oxide Conversion Facility Off-Gas
System
While the PUREX plant has been on standby, a plutonium oxide
facility has been added to the plant.
Off-gases from the plutonium nitrate storage vessels and the
prereduction tank pass through a heater and the* through two
stages of HEPA filtration. There is a combined flow of about
1,583 1/min at 60°C. Blowers deliver these gases to the
ventilation system No. 1 air tunnel.
Off-gases from the calciner pass through a porous stainless
steel filter at a flow rate of about 186 1/min at 157'C to remove
plutonium oxide particles. These gases, along with the off-gases
from the filtrate concentrator and the vessel vent gases from the
oxide rework facility, are fed to a scrubber to remove nitric
acid The off-gases from the vacuum header pass through a vacuum
tank*and are combined with the scrubber off-gases. The combined
gases then pass through two vacuum dropout tanks in series to
remove entrained liquids. The combined gas flow of about
400 1/min then goes through a heater and two stages of HEPA
filtration in series. A vacuum pump delivers the gases to the
blowers that exhaust to the ventilation system No. 1 air tunnel.
2.4.2.2.4 Canvon Ventilation System
Ventilation system No. 1 provides ventilation air for the
process cells in the PUREX canyon. Added to this air are the
gases from the process off-gas cleanup system and from the
plutonium oxide conversion facility off-gas treatment system.
The combined gases are exhausted through filters at a flow
rate of 3,570 m3/ndn. TWO glass fiber filters and one HEPA
filter are installed in parallel. Each unit is designed to
handle the full canyon ventilation air flow. Unit one, which was
installed in 1955, now has marginal capacity because of the
accumulation of solids. Unit two is run in parallel with unit
one. Unit three is on standby. The filters are installed
underground. When they are no longer usable, they will be sealed
and left in place. Recent tests have shown the two fiberglass
filters to have efficiencies greater than 99.95 percent for
0.3 micron particles. Unit three is designed to remove
99.97 percent of the 0.3 micron particles from the ventilation
air. Fans deliver the filtered gases to the PUREX main stack,
291-A-l.
2.4.2.3 Combined Exhaust, Buildings 405, 4621E, 4717;
Building 491-S, and Building 4717
Radioactive gases generated in the Fast Flux Test Facility
(FFTF) are a result of neutron activation of the reactor cover
gas or are released from the fuel through defective fuel
cladding. These gases are processed through the Radioactive
2-37
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(RAPS) and "leased to the atmosphere
a™ exhaust' There are about 200-280 1/min of
to no Jn?? ? fource'. Effluent from cells and spaces subject
to potential contamination is processed through the Cell
™™£?£ J™ Jrocessing System (CAPS) before release through the
combined exhaust. The CAPS contributes about 1,700-2,000 1/min
ovh*n V Ombln*d fxhaust other contributions to the combined
re ab°^ 10° m /min from ^he normal heating and
ind ;y8^ a"d ab°Ut 57° m /min from the containment
and ventilating system.
»„/. x GaS6S ff?m the f ission
-------
week from the gas tag sample trap, and (4) effluent from the
RAPS, if radiation monitors detect radioactivity above
1E-3 microcuries per cubic centimeter.
As with the RAPS, the gases are drawn into a vacuum tank and
through filters to remove moisture and oils by two compressors,
one online and one on standby, and thence into a surge and delay
tank for decay of argon-41. The CAPS input flow normally has a
very low radioactivity level («lE-7 uCi/cc) . In normal
operation, the gases from the surge and delay tank are then
routed to the combined exhaust. If radiation monitors detect
radiation, the gases are routed to the cold box. Two drying
units dry the gas to a dewpoint of -68°C or less. The liquid
from the drying unit may contain some tritium and is sent to the
liquid waste system. Two liquid-nitrogen-cooled charcoal-delay
beds in series provide decay time for short-lived radionuclides.
If the gases exiting the charcoal-delay beds have a radioactivity
of less than 5E-3 uCi/cm3, they are routed to the combined
exhaust. If the radioactivity exceeds this limit, the gases are
routed back to the CAPS vacuum tank for another pass through the
CAPS.
2.4.3 Basis for the Dose and Risk Assessment
2.4.3.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.4-1.
In modeling the site, all releases were assumed to be made
from the 200 Area, since this is the major source of dose, and
the nearest individual at risk is assumed to be 15,000 m from the
source (PNL87). The releases were assumed from a 10-m stack.
Default particle sizes (1.00 AMAD for plutonium-238) and
solubility classes (Class Y for plutonium-238) were assumed.
2.4.3.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Moses
Lake/Grant, Washington. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 15,000 m from the assumed
release point.Food consumption rates appropriate to a rural
location were used.
2.4.3 Results of the Dose and Risk Assessment
The major contributors to exposure are argon-41 (61 percent)
and plutonium-238 (33 percent). The predominant exposure
pathways are inhalation for uranium-238 and air immersion for
argon-41.
2-39
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Table 2.4-1. Radionuclides released to air during 1986 from the
Hanford Reservation.
Nuclide Release Rate (Ci/y)
Am-241
Ar-41
Ce-144
Co-60
Cs-137
Cs-138
H-3
1-129
1-131
1-132
1-133
1-135
Kr-85
Kr-85m
Kr-87
Kr-88
La-140
Mo-99
Nb-95
Pb-212
Pm-147
Pu-238
Pu-239
Pu-241
Rb-88
Ru-106
Sn-113
Sr-90
Tc-99
U-234
U-235
U-236
U-238
Xe-133
Xe-135
Zr-95
5.3E-4
1.3E+5
2.6E-3
1.1E-2
8.0E-3
1.9E+3
8.7E+1
5.3E-1
5.6E-1
2.6E-1
2.3E+0
3.5E-1
5.3E+5
3.3E+2
8 . 5E+2
3 . 6E+2
3.4E-2
9.6E-2
3.5E-3
1.8E-1
1.2E-2
8.9E-2
3.2E-3
1.4E-2
3 . 6E+2
4.5E-1
1.8E-1
1.2E-3
2.0E-4
6.8E-5
8.4E-6
5.4E-7
4.2E-5
6.7E+1
1.3E+3
4.0E-3
The results of the dose and risk assessment are presented in
Tables 2.4-2 through 2.4-4. Table 2.4-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.4-5 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.4-4 presents the estimated distribution of fatal cancer risk to
the regional population.
2-40
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Table 2.4-2,
Organ
Estimated radiation dose rates from the Hanford
Reservation.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
Remainder
Gonads
Endosteum
Red marrow
Breast
2.8E+0
l.OE+0
1.1E+0
6.3E+0
1.2E+0
9.4E-1
5.6E+1
1.7E+1
1.5E+1
1.7E+2
2.3E+1
1.2E+1
Table 2.4-3. Estimated fatal cancer risks from the Hanford
Reservation.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
3E-5
6E-3
Table 2.4-4.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Hanford
Reservation.
Risk Interval
Totals
Number of Persons
350,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
5
140
210
0
0
0
0
,200
,000
,000
0
0
0
0
1E-3
4E-3
1E-3
6E-3
2.4.5 Supplementary Controls
The N-Reactor shutdown in 1987 has reduced emissions of
argon-41 and plutonium-238 sufficiently to lower the estimated
maximum exposure below 1 mrem/y. Therefore, additional emission
controls for airborne radionuclides are not discussed.
2-41
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2.5 OAK RIDGE RESERVATION
2-5.1 Description and Existing Control
The Oak Ridge Reservation (ORR) , located in eastern
Tennessee occupies approximately 15,000 ha in a valley between
the Cumberland and southern Appalachian mountain ranges. The ORR
4e; 3 us t southwest of the city of Oak Ridge and about 24 km west
of Knoxville, Tennessee. The reservation is bounded on the
northeast, southeast, and southwest by the Clinch River.
2.5.1.1 Site Description
piHr, TS%1?aj0f ^ilities at the ORR are the Y-12 plant, the Oak
Ridge National Laboratory (ORNL) , and the Oak Ridge Gaseous
Diffusion Plant (ORGDP) . In addition to these major facilities,
the Oak Ridge Associated Universities and the Comparative Animal
Research Laboratory are also located at the site.
The Y-12 plant, located adjacent to the city of Oak Ridge
is a manor nuclear weapons production facility, processing '
enriched uranium, its major missions include fabricating nuclear
weapons components, processing source and special nuclear
material, and providing support to the weapons design
laboratories. While the actual processes employed at the Y-12
plant are classified, the activities associated with these
missions include production of lithium compounds, recovery of
enriched uranium from scrap materials, and fabrication of uranium
and other materials into finished parts and assemblies.
Fabrication operations include vacuum casting, arc melting,
powder compaction, rolling, forming, heat treating, and
machining.
The ORNL is a large multipurpose research laboratory where
basic and applied research in all areas relating to energy is
conducted. The ORNL facilities include nuclear reactors,
chemical pilot plants, research laboratories, and radioisotope
production laboratories.
The significant airborne radioactive emissions from the ORNL
are from the Central Radioactive Gas Disposal Facility (CRGDF)
and the Tritium Target Fabrication Building. The CRGDF is
equipped with charcoal filters for radioiodines and HEPA filters
for particulate emissions. There are no controls for the noble
gases krypton and xenon or for tritium. The Tritium Target
Fabrication Building also releases tritium without effluent
control .
Until the summer of 1985, the ORGDP 's primary mission was to
provide enriched uranium for use in nuclear reactors. The ORGDP
uses the gaseous diffusion process. The facility was placed in
"ready standby" in August 1985. Since that time, the decision
has been made to shut down permanently the enrichment cascade.
ORGDP is also involved in developing and demonstrating more
2-42
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energy-efficient and cost-effective methods of enriching uranium,
such as the gas centrifuge process and the atomic vapor laser
isotopic separation (AVLIS) system. However, the gas centrifuge
process was shut down in 1985, and the work on AVLIS has been
significantly reduced.
2.5.1.2 Major Release Points and Existing Emission
Control Technology
There are approximately 350 process exhaust stacks at the
Y-12 plant, of which approximately 85 serve operations with the
potential to release uranium to the atmosphere. Although actual
emission controls are classified, it is known that the majority
of the stacks serving uranium operations are equipped with
particulate control devices such as HEPA and fabric filters.
The purge cascade was the largest source of airborne
radioactive emissions at the ORGDP. Effluents from the purge
cascade were passed through sodium fluoride traps, alumina traps,
and potassium hydroxide (KOH) scrubbers.
2.5.2 Basis for the Dose and Risk Assessment
2.5.2.1 Source Terms and Release Point Characterization
The airborne emissions from all facilities at ORR are
summarized in Table 2.5-1. These emissions data were obtained
from the DOE's Effluent Information System and the Annual
Environmental Monitoring Report for 1986 (Or87a, Or87c).
In modeling the site, all releases were assumed to be made
from the Y-12 plant, since this is the major source of uranium.
Data on the actual stacks at the Y-12 Plant are classified.
Therefore, the releases were assumed from a 10-m stack, with a
flow of 200 cfm (Mo86).
Default particle sizes (1.00 AMAD) were assumed. The
uranium-234 was assumed to be one-half solubility class W and
one-half solubility class Y.
2.5.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Knoxville, Tennessee. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located in the city of Oak Ridge,
750 m from the assumed release point. Food consumption rates
appropriate to a rural location were used.
2-43
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Table 2.5-1,
Radionuclides released to air from Oak Ridge
Reservation during 1986.
1986 Emissions
(Curies/year)
Nuclide
Y-12
ORNL
ORGDP
Other
Total
C-14
Cu-64
Ga-67
H-3
1-125
1-131
Kr-85
P3-234M
Tc-99
TC-99M
Th-234
Tl-201
U-234
U-234
U-235
U-236
U-238
Xe-133
Y-90
3 . 1E+4
3.6E-2
1.1E+4
3.7E-4
1.3E-2 1.2E-1
3.7E-4
7.0E-2 7.4E-3
7.7E-2
6.4E-3
8.0E-6
2.8E-2 3.6E-4
5.2E+4
l.OE-4
2.0E-6
3.0E-6
4.0E-3
1.5E-5
1.3E-4
3.0E-6
5.0E-6
2.0E-5
l.OE-4
2.0E-6
3.0E-6
3 . 1E+4
1.5E-5
3.6E-2
1 . 1E+4
3.7E-4
1.3E-1
3.0E-6
3.7E-4
5.0E-6
7.7E-2
7.7E-2
6.4E-3
8.0E-6
2.8E-2
5.2E+4
2.0E-5
2.5.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-234
(40 percent), tritium (35 percent), and uranium-238 (13 percent).
The predominant exposure pathway is inhalation for uranium-234
and uranium-238, and ingestion for tritium.
The results of the dose and risk assessment are presented in
Tables 2.5-2 through 2.5-4. Table 2.5-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.5-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population.
Table 2.5-4 presents the estimated distribution of fatal cancer
risk to the regional population.
2-44
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Table 2.5-2.
Organ
Estimated radiation dose rates from the Oak Ridge
National Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
Remainder
2.2E+1
2.0E+0
4.3E+2
7.8E+1
Table 2.5-3. Estimated fatal cancer risks from the Oak Ridge
National Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/year
8E-5
3E-2
Table 2.5-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from Oak Ridge
National Laboratory.
Totals
Number of Persons
850,000
Deaths/year
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
28,000
760,000
60,000
0
0
0
0
8E-3
3E-2
8E-4
3E-2
2.5.4 Supplementary Controls
The emission control technology (ECT) currently used to
reduce airborne radioactive emissions at facilities in the major
Oak Ridge areas was described in Section 2.5.1. Potential
additional emission control technologies are described in the
following sections (Mo86).
2-45
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2.5.4.1 Additional Emission Control Technology for the ORNL
Central Radioactive Gas Disposal Facility
The major portion of the radiological hazard from the gas
disposal facility is due to the emission of tritium. Practical
control technology exists for removal of these materials from low
flow rate air streams only. Because of the high rate of emission
(64 m/sec) from the stack of this facility, additional control
technology must be implemented before the individual source
stream is diluted with ventilation air or other gas streams.
Much of the tritium emission is in the form of tritiated
water. This portion can be removed by passing the source stream
through a dryer containing molecular sieve materials for water
removal and then regenerating the adsorber material with heat. A
pair of such dryers, operated alternately, will provide for the
continuous removal of tritiated water from the source. Table
2.5-5 presents the expected emission rate for tritium at the
CRGDF if this additional control technology is implemented.
Table 2.5-5. Anticipated new emission rate for tritium at
CRGDF.
Present Emission Postulated ECT New Emission
Rate (Ci/y) Removal Efficiency Rate (Ci/v)
3.12E+4 90% 3.12E+3
The cost of an emission control system for the removal of
tritiated water is estimated at $1.66 million. This includes
$1 million for construction, $0.2 million for engineering, and a
$0.46 million contingency. These cost estimates are highly
dependent upon the ease of incorporating the potential controls
into the existing gas handling system. It is possible that the
existing gas handling system would have to be completely replaced
to accommodate more controls.
2.5.4.2 Additional Emission Control Technology for the ORNL
Tritium Target Fabrication Building
Tritiated water can be removed from the gaseous exhaust by
passing the exhaust air stream through a dryer containing
molecular sieve materials for water removal and then regenerating
the adsorber material by the application of heat. A pair of such
driers, operated alternately, would provide for the continuous
drying of the exhaust and the collection of tritiated water for
storage or further processing.
Analytical information concerning the gases present in the
stack exhaust indicates that only about 1 percent of the tritium
is in the form of tritiated water. At this time, it is not
2-46
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practical to remove tritium in the form of hydrogen gas from the
large gaseous stream flow emanating from the Tritium Target
Fabrication Building.
It is postulated that over 90 percent of the tritiated water
would be collected by the application of the additional
technology; however, since the tritiated water represents only a
small portion of the total tritium from this facility, the
present emissions (1.2 x 103 Ci/y) would not be significantly
reduced.
The cost of an emission control system specifically designed
for the removal of tritiated water is estimated at $1.66 million.
This includes $1 million for construction, $0.2 million for
engineering, and a $0.46 million contingency.
2.5.4.3 Additional Emission Control Technology for the Y-12
Plant (Uranium Product Recovery)
There are three sources of emissions from uranium product
recovery. The major source is the West Head House, Building
9212. The emission controls described here apply to this
facility.
Installation of an additional stage of HEPA filters would
reduce the amount of particulate emission and uranium-234 and
uranium-238 that bypasses the present ECT system, if the present
ductwork can be adapted or expanded to allow incorporation of
more HEPA filters downstream of the existing filter system.
HEPA filters are estimated to remove at least 99.95 percent
of particulate materials in a single pass. It has been shown,
however, that uncollected materials have a lower collection
efficiency when passed through a second HEPA filter stage.
Collection efficiency estimates for such a second stage may vary
due to the size distribution of the original particulates. It is
postulated that a second HEPA filter installed in series will
remove 99 percent of the remaining particulates and reduce the
amount of uranium-234 from 0.154 Ci/y to 0.093 Ci/y, and reduce
the amount of uranium-238 from 2.8E-2 Ci/y to 3.0E-3 Ci/y.
The cost of the control devices presently installed in the
uranium product recovery facility is $55,000. The estimated cost
for installation of backup HEPA filtration within the existing
system is an additional $20,000. The present annual operating
cost is $14,640. Based upon the assumption that the air
capacityof the system can be maintained by the existing fan
system, additional power and HEPA changeout requirements would
increase the operating cost about 20 percent.
If significant structural additions or modifications are
necessary for proper operation and maintenance of the expanded
air control system, then significant cost increases can be
anticipated. In addition to the HEPA filter cost, modifications
2-47
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that include ductwork, blowers, dampers, instrumentation, and
electrical work would increase the cost to about $455,000.
Engineering costs of about $115,000 and a 35 percent contingency
would raise the total project cost to over $800,000. Major
structural additions will further increase the cost. Operating
costs are expected to double with the implementation of this
modified system.
2.5.4.4 Additional Emission Control Technology for the Y-12
Plant (Uranium Product Preparation)
Replacing the existing scrubber with a high-energy venturi
scrubber and adding a backup stage of HEPA filtration would
reduce the emission of uranium-234 from this facility, if the
present ductwork can be modified or expanded to allow
incorporation of these changes.
Based upon the arguments presented in Section 2.5.4.3, about
99 percent of the particulate emission would be removed by the
addition of a second HEPA filter stage. In addition, the use of
a high-energy venturi scrubber would improve the collection
efficiency of the scrubber system by 20 percent and would provide
higher efficiency (98-99 percent) for removal of particulates
below 1 micron. By implementing the additional ECT, the emission
of uranium-234 would be reduced from 2.98E-2 Ci/y to less than
2.38E-4 Ci/y.
The cost of the control devices already installed in the
uranium product preparations C-I wing building is $46,300. The
estimated additional cost for adding a high-efficiency scrubber,
including demisters, is $15,000 ($11,000 capital plus $4,000
installation). The estimated additional cost for backup HEPA
filtration is $9,000. These estimates are based upon the
assumption that the existing fan system is capable of maintaining
the necessary pressures and flows with the added ECT.
The present annual operating cost of $6,880 is expected to
increase 30 percent due to the power necessary to maintain high
differential pressures in the venturi and provide flow through
both HEPA filters.
If significant structural additions or modifications and
other equipment such as special nitric acid scrubbers are
necessary for proper operation and maintenance of the expanded
air control system, then significant cost increases can be
anticipated. In addition to the HEPA filter cost, modifications
that include ductwork, blowers, dampers, instrumentation, and
electrical work would increase the cost to about $200,000.
Engineering costs of about $80,000 and a 35 percent contingency
would raise the total project cost to about $400,000. Major
structural additions would further increase the cost. Operating
costs are expected to double with the implementation of this
modified system.
2-48
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2.5.4.5 Additional Emission Control Technology for the Y-12
Plant (Uranium Fuel Element Fabrication)
The fabrication process is located in the C Wing of Building
9212. Installation of HEPA filters would significantly reduce
the amount of particulate uranium-234 emitted from this facility,
if the present ventilation system can be modified or expanded to
allow installation of HEPA filters downstream of the roughing
filters.
HEPA filters collect almost 100 percent of the airborne
particulate materials from airstreams containing typical size
distributions of suspended materials. It is estimated that
99.95 percent of the materials that pass the roughing filters
will be removed by a single pass through HEPA filtration. Based
upon this assumption, the installation of HEPA filters would
reduce the annual emission of uranium-234 from 1.73E-2 Ci to less
than 8.7E-6 Ci.
The uranium fuel element fabrication facility is now served
by a large ventilation system which exhausts air at the rate of
23.6 m3/sec. A similarly sized system which includes the
addition of HEPA filters is installed at the Y-12 plant uranium
denitrator. The difference in cost between these facilities is
$41,000, which is postulated as the cost to add HEPA filters to
the fabrication facility. This is based upon the assumption that
the air capacity of the system can be maintained by the existing
fan system. The cost of additional power requirements and the
cost of HEPA filter replacement will double operating costs to
about $50,000 per year.
If significant structural additions or modifications such as
air coolers are necessary for proper operation and maintenance of
the expanded air control system, then significant cost increases
can be anticipated. In addition to the HEPA filter cost,
modifications that include ductwork, blowers, dampers,
instrumentation, and electrical work would increase the costs to
about $825,000. Engineering costs of about $200,000 and a
35 percent contingency would raise the total project cost to
$1,450,000. Major structural additions would further increase
the cost. Operating costs are expected to double with the
implementation of this modified system.
2.5.4.6 Additional Emission Control Technology for the Oak
Ridge Gaseous Diffusion Plant (Purge Cascade)
The Purge Cascade is part of the Oak Ridge Gaseous Diffusion
Plant K-27 process area. All diffusion plant process buildings
are three-story, steel frame with 6-mm transite side panels
(preformed concrete). The Purge Cascade is intended to separate
light gases from UF6 and vent them to the atmosphere through the
emission control devices. Emissions from this building represent
the largest hypothetical risk from the Oak Ridge Gaseous
Diffusion Plant.
2-49
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Radioactive emissions from the ORGDP Purge Cascade consist
mainly of gaseous and particulate uranium and technetium
fluorides that pass through existing abatement equipment. A new,
low-energy venturi scrubber is planned for installation
downstream of the existing spray scrubber to reduce mist
carry-over and thus help mitigate equipment corrosion problems.
This new scrubber should also reduce airborne emissions somewhat
by removing more airborne particulate and droplet materials;
however, quantification of the scrubbing action is not precise.
It is dependent upon the gaseous solubility and upon the
effectiveness of the mixing and impinging action. Addition of
this device is estimated to remove about 50 percent of the
remaining radioactive emissions.
The cost of the emission control devices now installed at
the Purge Cascade is $1.25 million. The estimated additional
cost for purchase of a low-energy venturi is $13,000. The added
annual operating cost for this installation is estimated to be
minor ($1,300) compared to the present annual operating cost of
$300,000. Installation costs, which are sensitive to the amount
of modification necessary to incorporate the added device, were
not estimated.
Table 2.5-6. Summary of capital and operating costs for
supplementary controls at the Oak Ridge
Reservation.
Facility Plant
Capital Operating
Nuclide Control Technology Cost($K) Cost($K)
ORNL
CRGDF
Y-12
Y-12
U Prod.
Recovery
U Prod.
Prepara-
tion
Y-12
H-3
U-234
U-238
U-234
U-238
U Fuel U-234
Element U-238
Fabrication
Tritiated water/ $1,660 $ 0
sieve dryer
system
Additional stage $800 $29
HEPA filters and
high-energy venturi
scrubbers
Additional stage $400 $13
HEPA filters and
high-energy venturi
scrubbers
Additional stage $1,450 $50
HEPA filters and
high-energy venturi
scrubbers
TOTALS: $4,310 $92
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2.6 SAVANNAH RIVER PLANT
2.6.1 Site Description
The facilities at the Savannah River Plant are used
primarily to produce plutonium and tritium, the basic materials
required for nuclear weapons. Materials for medical and space
applications are also manufactured here, however. The Savannah
River Plant is situated along the Savannah River at a site 35 km
southeast of Augusta, Georgia. The site covers about 770 km2.
Operations are grouped into five major areas (designated the
100, 200, 300, 400, and 700 Areas) according to their operational
function in the plutonium manufacture/recovery process.
2.6.1.1 100 Area - Nuclear Production Reactors
Three production reactors were in operation. The three
reactors produce plutonium and tritium by irradiation of uranium
and lithium. Heavy water is used both as a neutron moderator and
as a primary coolant. All three reactors have been subsequently
shut-down pending the resolution of safety issues and other
operational problems.
2.6.1.2 200 Area - Separations and Waste Management Facilities
Nuclear fuel reprocessing occurs in this area. Plutonium is
recovered from irradiated uranium by the PUREX solvent-extraction
process. Enriched uranium and plutonium-238 are recovered from
other irradiated materials by a solvent-extraction procedure
similar to the PUREX process.
2.6.1.3 300 Area - Fuel and Target Fabrication
Tubular fuel and target elements are produced by cladding
depleted uranium fuel in aluminum or aluminum/lithium shells. A
low-power reactor and a subcritical test reactor are then used to
test for assembly defects.
2.6.1.4 400 Area - Heavy Water Production and Recovery
Heavy water is produced from river water by distillation and
extraction. Heavy water is also recovered from contaminated
reactor coolant. Heavy water is transported from this area to
the 100 Area for use in the production reactors.
2.6.1.5 700 Area - The Savannah River Laboratory
Research and process development work is performed at the
Savannah River Laboratory. Major activities in this area include
fabrication of fuel element and target prototypes; fabrication of
radioisotopic sources for medical, space, and industrial
applications; thermal and safety studies of reactor operations;
and applied research in physics and the environmental sciences.
2-51
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2.6.2 Maior Release Points and Existing Emission
Control Technology
Radionuclides are released into the atmosphere from a number
of facilities on the SRP site (Ze87, Mo84). Each operating area
has one or more discharge stacks that have emission control
equipment installed. Monitoring systems record data on a
real-time or a near real-time basis. All stack release data are
reported annually. The largest quantities of radionuclides are
released from the fuel reprocessing areas (F and H Areas). The
three production reactor stacks (C, K, and P) release the next
largest quantities, followed in descending order of quantities of
radionuclide emissions by the heavy water rework plant, the
Savannah River Laboratory, and the fuel and target fabrication
plant.
Tritium is released from six facilities, with the tritium
facilities (232-H, 234-H, 238-H) contributing about 66 percent of
the total tritium dose; the reactor areas (105-C, 105-K, and
105-P) contribute about 10 percent, 16 percent, and 7 percent,
respectively; the Moderator Rework Unit (420-D) contributes about
0.6 percent; and the Savannah River Laboratory contributes less
than 0.01 percent.
Argon-41 is released exclusively at the operating reactors
in roughly equal proportions.
Carbon-14 is released from the three operating reactors and
from the separations plants in F and H Areas in approximately
equal proportions.
In terms of radiation dose to the offsite population, the
principal sources are the H Area tritium facilities, followed in
order of decreasing contribution by 105-K, 105-C, 105-P, and the
F and H Areas separations plants. The contributions from other
source locations are negligible (less than 1 percent).
2.6.2.1 200-H Area Tritium Facility Stacks
Releases of tritium from the four stacks associated with the
tritium facilities in the 200-H Area constitute the principal
sources of radioactive emissions at SRP.
The emission control system uses a long transit volume (the
"Serpentine") as a means to capture and hold air flows from
process hoods that contain accidental releases of tritium, so
that the contained tritium can be removed from the air before
discharge to the stack. A nominal air flow continually passes
through the Serpentine to the stack line. Air from the process
hoods also normally flows to the stack line. When an in-line ion
chamber detects a preset level of tritium in the hood outflow,
the Serpentine inlet from the process hood is opened, and the
hood flow is diverted to the Serpentine. The volume of the
holdup line is sufficient to prevent loss of the tritium burst to
2-52
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the stack. An ionization chamber near the end of the Serpentine
detects the tritium concentration as it exits the Serpentine. If
the concentration is greater than a preset limit, the volume that
exceeds the limit is subsequently diverted and processed through
the Hopcalite stripper and zeolite beds to remove the tritium.
If the concentration is less than the preset limit, the trapped
air volume is discharged to the stack.
The system uses a holdup tank into which batches of inert
gases or air from various operational activities are placed for
eventual processing through a Hopcalite stripper and two zeolite
beds.
The efficiency of the Hopcalite stripper varies with
operating conditions (oxidizer bed temperature, oxygen and
hydrogen content in the gases to be treated) and can range from a
few percent to nearly 100 percent. The actual average efficiency
of the strippers at SRP is classified information and cannot be
reported here.
2.6.2.2 Production Reactor Area Stacks
Releases of radioactivity into the atmosphere at the three
production reactors are the next largest contributors to the
offsite population dose resulting from operations at the SRP.
Actual releases will vary from reactor to reactor, year by year,
depending upon activities.
A ventilation system typical of the production reactors is
described below. The filter system consists of inlet prefilters
to remove particulates from incoming air, moisture separators to
remove entrained moisture droplets from the outgoing air stream,
particulate (HEPA) filters to remove particulate material, and
charcoal filters to remove iodines. There are no provisions for
reducing the emission of tritium, noble gases, or carbon-14.
Monitoring equipment at the 61-m reactor stacks includes
continuous Kanne chambers and dehumidifier samplers for
monitoring tritium emission, a continuous noble gas monitor
utilizing a Ge-Li detector/multichannel analyzer system, a
continuous charcoal filter for monitoring radioiodines, and a
continuous filter paper sampler for particulate monitoring.
2.6.2.3 200-F and 200-H Area Separation Plants
Releases of radioactivity to the 291-F and 291-H and
associated stacks (221-F and 221-H facilities) are principally
carbon-14, noble gases, and small amounts of iodine.
Effluent control equipment on the 200-F Area ventilation
systems consists principally of particulate filters: fiberglass,
HEPA, and sand filters. Silver nitrate beds are used for
scrubbing iodine from the dissolver offgas stream.
2-53
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2.6.3 Basis for the Dose and Risk Assessment
2.6.3.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed below in Table 2.6-1.
Table 2.6-1. Radionuclides released to air during 1986 from
Savannah River Plant.
Nuclide
Release Rate (Ci/y)
Am-241
Ar-41
C-14
Ce-141
Ce-144
Cm-244
Co-60
Cs-134
Cs-137
H-3
1-129
1-131
Kr-85
Kr-85m
Kr-87
Kr-88
Nb-95
Os-185
Pu-238
Pu-239
Ru-103
Ru-106
Se-75
Sr-89
Sr-90
U-234
U-238
Xe-131m
Xe-133
Xe-135
Zr-95
1.9E-4
8.3E+4
5.6E+1
1.9E-5
1.1E-2
2.8E-5
8.0E-6
6.9E-4
3.0E-3
4.2E+5
8.8E-2
2.6E-2
7 . 1E+5
2.0E+3
1.4E+3
2.4E+3
9.2E-3
1.4E-4
2.0E-3
2.9E-4
3.5E-3
5.9E-2
2.1E-5
9.2E-4
1.4E-3
1.6E-3
1.6E-3
3.0E-1
1.1E+4
2 . 6E+3
4.4E-3
In modeling the site, all releases were assumed to be made
from the F-separations area. The releases were aggregated to
five stacks: Stack 1 is the 100 Area (60 m): all nuclear
production reactors; Stack 2 is the 200 Area (61 m): plutonium
and uranium separation; Stack 3 is the 300 Area (10 m): Fuel and
Target Fabrication; Stack 4 is the 400 Area (10 m): Heavy Water
2-54
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Recovery and Production; Stack 5 is not used; and Stack 6 is the
700 Area (50 m): Laboratory. Default particle sizes (1.00 AMAD)
and solubility classes were assumed.
2.6.3.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Augusta/Bush, Georgia. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 15,000 m from the assumed
release point (Ze87). Food consumption rates appropriate to a
rural location were used.
2.6.4 Results of the Dose and Risk Assessment
The major contributors to exposure are tritium (77 percent)
and argon-41 (18 percent). The predominant exposure pathways are
inhalation, ingestion, and air immersion.
The results of the dose and risk assessment are presented in
Tables 2.6-2 through 2.6-4. Table 2.6-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.6-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure as well
as estimated deaths per year in the regional population. Table
2.6-4 presents the estimated distribution of fatal cancer risk to
the regional population.
Table 2.6-2
Organ
Estimated radiation dose rates from the Savannah
River Plant.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Remainder
Gonads
Breast
Lungs
Red marrow
3.2E+0
2.6E+0
2.6E+0
2.7E+0
2.6E+0
6.7E+2
5.5E+2
5.5E+2
5.6E+2
5.5E+2
Table 2.6-3. Estimated fatal cancer risks from the Savannah
River Plant.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
8E-5
2E-1
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Table 2.6-4. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Savannah
River Plant.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
^.
-
-
-
-
-
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
550,000
0
0
0
0
0
0
2E-1
0
0
Totals 550,000 2E-1
2.6.5 Supplementary Controls
This section examines specific sources of radionuclide
emissions, and existing control systems, discusses current
discharge rates, suggests additional control equipment and
anticipated reduction in emissions, and estimates costs of the
suggested additional equipment (Mo86).
2.6.5.1 Additional Emission Control Technology for the
200-H Area Tritium Facility Stacks
Releases of tritium from the four stacks associated with the
Tritium Facilities in the 200-H Area constitute the principal
sources of radioactive emissions at SRP. They resulted in a
radiation dose to the offsite population of about 67 man-rem
during 1981. This dose represents about 57 percent of the total
population dose from SRP emissions.
The efficiency of the catalytic oxidizer system might be
improved by replacing the Hopcalite (80 percent MnO2 - 20 percent
CuO) beds with a palladium catalyst. Recycling the effluent
gases through the stripper combined with hydrogen swapping will
also improve the efficiency of the stripper. The SRP staff has
estimated that recycling could reduce normal tritium emissions by
25 percent. The cost of the system improvements is estimated to
be about $65 million. The system lifetime is estimated to be
about 15 years.
2.6.5.2 Additional Emission Control Technology for the
Production Reactor Area Stacks
Releases of radioactivity into the atmosphere at the three
production reactors are the next largest contributors to the
offsite population dose resulting from operations at the SRP.
Actual releases vary from reactor to reactor,.year by year,
.depending upon activities.
2-56
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Tritium emissions from the heavy water moderated reactors
could be reduced by (1) replacing tritiated moderator with fresh
moderator, (2) minimizing evaporation losses from the moderator,
and (3) removing tritium from the existing moderator. While none
of these approaches is classified as emission control technology,
they are operational in that they attempt to prevent tritium in
the ventilation system rather than attempting to remove the
tritium from the effluent air stream.
The first approach is not particularly viable. The effect
would be only temporary since the tritium levels in the moderator
build up with each year of reactor operation.
The second approach is normal operating practice and is
already carried out to the extent feasible.
The third approach would use either vapor phase catalytic
exchange with cryogenic distillation (CE-CD) or a thermal cycle
absorption process (TCAP). These processes have the potential
for reducing tritium emissions at the production reactors by
about 90 percent once steady-state operation is achieved after
about 6 years. SRP staff estimate capital costs for a CE-CD
system are to be in the $20-40 million range. Estimated annual
operating cost would be in the $1.5 to $2 million range, with an
estimated operating life of 30 years. No estimates are currently
available for the cost of a TCAP system.
Releases of argon-41 at the production reactors could be
reduced by installing a holdup volume into which the air
containing the argon-41 (from the annular cavity around the
reactor tank) could be routed, thus allowing the radioactivity to
decay to insignificant levels. A possible system would use an
existing 1,893-m tank in the emergency core cooling system. An
air flow of 1.4 to 4.3 m3/minute into an effective storage volume
of 707 m3 is expected to reduce argon-41 emissions by about 60
percent. The feasibility of utilizing the 1,893-m3 tank for this
purpose is being actively investigated. The capital cost of this
proposed system is small, since mostly existing systems and
eguipment would be used.
No other systems for reducing emissions from the production
reactors are presently under consideration.
2.6.5.3 Additional Emission Control Technology for the 200-F
and 200-H Area Separation Plants
Releases of radioactivity to the 291-F and 291-H and
associated stacks (221-F and 221-H facilities) are principally
carbon-14, noble gases, and small amounts of iodine.
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Carbon-14, the noble gases, and iodine contribute nearly all
of the radiation dose from the separations plants. An absorber
system utilizing flaked barium hydroxide octahydrate to form
barium carbonate, thus capturing the carbon-14, could be
installed. In addition, one of several techniques for capturing
the noble gases (particularly krypton-85) could also be
installed. These techniques, cryogenic distillation,
fluorocarbon absorption, and absorption on mordenite beds, all
have decontamination factors of about 100. The iodine removal
capability of the existing iodine absorber beds utilizing silver
nitrate could be improved if the beds were converted to silver
mordenite, moved from the dissolver off-gas system, and installed
in the vessel vent system.
SRP staff estimates that an integrated off-gas treatment
system utilizing the above techniques would cost about $50
million per plant and would have annual operating costs of about
$3 million.
2.7 FEED MATERIALS PRODUCTION CENTER
2.7.1 Description and Existing Controls
2.7.1.1 Site Description
The Feed Material Production Center, located 32 km northwest
of Cincinnati, Ohio, produces uranium metal and other materials
for DOE facilities. The uranium may be natural, depleted, or
enriched with respect to uranium-235.
Raw materials are processed in the following manner. The
material is first dissolved in nitric acid and separated by
liquid organic extraction. The recovered uranium is reconverted
to uranyl nitrate, heated to form uranium trioxide, reduced to
uranium dioxide with hydrogen, and reacted with hydrogen fluoride
to form uranium tetrafluoride. Purified metal is made by
reacting the uranium tetrafluoride with metallic magnesium in a
refractory-lined vessel.
The U.S. DOE Effluent Information System Nuclide Database
Master List for 1986 reports emissions in 1986 from eight plants
at the FMPC (EIS86). These emissions are listed in Table 2.7-1.
The emissions are identified as natural uranium in the form of
particulates. Each plant at the FMPC has several stacks.
DOE forecasts indicate increased use of the FMPC in support
of increased work at other DOE sites (We87, Mo84). The actual
magnitude of this increased FMPC production depends on the needs
of other DOE sites but could reasonably be expected to double the
1981 production. A corresponding increase in total uranium
emissions would therefore be expected, assuming no change in
emission control technology.
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2.7.1.2 Major Release Points and Existing Emission
Control Technology
Emission control technology at the FMPC differs from that of
other sites in two major aspects: (1) emissions are essentially
all particulates, with natural uranium being the predominant
radionuclide; and (2) each plant at the FMPC has multiple stacks,
each with its own emission control device and each providing
ventilation to a specific area or specific equipment within a
given plant.
Chemical and radioactive emissions at the FMPC are
controlled by wet scrubbers, bag-type dust collectors, and
electrostatic precipitators. The radioactive emissions from the
various plants are essentially all particulate emissions.
Emissions from Plants 4, 5, and 8 are controlled by the bag-type
dust collectors or wet scrubbers.
Bag-type dust collectors are installed on many of the
stacks. The dust collectors for these particular stacks have
been shown to have total system efficiencies of >99.9 percent
over a 2-year period. Most of the material losses occur because
of cloth bag ruptures or other malfunctions that allow the dust
to bypass the filter.
Stack emissions are constantly sampled using a permanently
installed in-stack sampling system. These systems require the
collection of about 1 g of material before the collection filters
are removed for analysis. A continuous stack monitoring device
that will be used in addition to the existing stack samplers has
been installed on selected stacks. The results to date indicate
that the new stack monitoring device is very sensitive to small
quantities of material loss; it has detected minor leaks in dust
collection bags that, prior to its installation, had gone
undetected until a buildup of material on the stack sampler was
found.
2.7.2 Basis for the Dose and Risk Assessment
2.7.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed below in Table 2.7-1.
Table 2.7-1. Radionuclides released to air during 1986 from
FMPC.
Nuclide Release Rate (Ci/y)
U-234 2.0E-2
U-238 2.0E-2
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In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 Amad) were
assumed. The uranium-234 and uranium-238 emissions were assumed
to be 1/3 Class D, 1/3 Class W, and 1/3 Class Y.
2.7.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Covington/GTR Cincinnati, Ohio. The 0-80 km population
distribution was produced using the computer code SECPOP and 1980
Census Bureau data. Nearby individuals were located 800 m from
the assumed release point (We87). Food consumption rates
appropriate to an urban location were used.
2.7.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-234
(53 percent) and uranium-238 (48 percent). The predominant
exposure pathway is inhalation for uranium-234 and uranium-238.
The results of the dose and risk assessment are presented in
Tables 2.7-2 through 2.7-4. Table 2.7-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.7-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.7-4 presents the estimated distribution of fatal cancer risk to
the regional population.
Table 2.7-2. Estimated radiation dose rates from FMPC.
Organ Nearby Individuals Regional Population
(mrem/y) (person-rem/y)
Lungs 1.9E+1 1.1E+2
Table 2.7-3. Estimated fatal cancer risks from FMPC.
Nearby Individuals Regional (0-80 km) Population
Lifetime Fatal Cancer Risk Deaths/y
3E-5 3E-3
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Table 2.7-4. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from FMPC.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
85
4,100
3,300,000
0
0
0
0
2E-5
1E-4
3E-3
Totals 3,300,000 3E-3
2.7.4 Supplementary Controls
The U.S. DOE Effluent Information System Nuclide Database
Master List for 1986 reports emissions in 1986 from eight plants
at the FMPC. Although the major emission sources (stacks) differ
each year, Plants 4, 5, and 8 are consistently the greatest
source of emissions. The emissions are identified as natural
uranium in the form of particulates (EIS86, Mo84).
As mentioned, DOE forecasts indicate increased FMPC
production, perhaps as much as double the 1981 production. A
corresponding increase in total uranium emissions would therefore
be expected, assuming no changes are made in the existing
emission controls.
2.7.4.1 Emission Control Technology
The FMPC has over 50 dust collection stacks in either full-
or part-time operation. The operating stacks already use very
efficient dust collection systems. Additional improvement in
reducing operational releases is expected by using Goretex fabric
bags rather than wool bags and by using administrative controls
in conjunction with the continuous stack monitor. Approximately
20 additional stacks have either been abandoned or placed on
standby status. Extensive repair and refurbishment would be
needed to return the abandoned and standby dust collection stacks
to operation.
However, neither the use of improved fabric bags in the
existing baghouses, nor installation of continuous radionuclide
stack monitors will insure reductions in uranium particulate
emissions at the FMPC. Reductions in emissions to lower levels
will require the installation of secondary air cleaning systems
on the primary emission sources located in Plants 4, 5, and 8.
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2.7.4.1.1 Proposed Emission Control Equipment:
It is proposed that HEPA filter systems be installed, in
addition to the existing emission control technology, on each of
the emission sources from Plants 4, 5, and 8 to reduce their
particulate emissions. By definition, each individual HEPA
filter must have a minimum particle removal efficiency
>99.97 percent for particles 0.3 urn diameter.
It has been assumed each system will use redundant HEPAs,
each sized for the stated airflow. Filter housings and ductwork
are stainless steel. Inlets to the HEPA systems are from
existing baghouses or scrubbers.
Placement of the proposed HEPA filter systems depends on:
(1) available existing space in Plants 4, 5, and 8; (2) space
that could be made available by removal of obsolete and unneeded
existing emission controls; and (3) allowable floor or roof live
loads at the locations proposed for installation of the HEPA
filter systems. The floor loading attributed to the proposed
systems is very light and for most of the filter systems would
require the addition of only minor secondary steel for support.
However, the Plant 5 perimeter appears heavily loaded and may
require the additional filter systems to be located outside the
existing structure, i.e., a new structure or structures may be
required for the filter systems installed in Plant 5.
2.7.4.1.2 Existing and Proposed Stack Monitoring
Systems
Radionuclide emissions at the FMPC are essentially all
natural uranium in the form of particulates. Emission particle
sizes and particle densities have not been reported.
Each stack at the FMPC has an in-stack sampler to determine
the quantity of particulates emitted. The sampler collects
particulates on a filter paper which is periodically removed and
the quantity of uranium collected determined by chemical
analyses. Each stack sampler is operated under isokinetic
conditions so that total stack emissions can be determined from
the quantity of material collected by the stack sampler.
The FMPC has installed new, continuous stack monitors on the
following stacks: Plant 4, Stacks G4-2, G4-12, and G4-14; Plant
5, Stacks G5-250, G5-260, and G5-261; Plant 8, Stack G43-27; and
Plant 9, Stack G9N1-1039. The continuous stack monitors are
pancake-type Geiger-Muller probes installed to monitor the back
side of the filter paper used in the in-stack particulate
sampler. The continuous stack monitors provide information in
real-time on stack emissions. The new monitors can be alarmed
for rate-of-rise of radioactivity detected and coupled to
automatic shutoff of the process equipment. The rate-of-rise
alarm on the continuous stack monitor indicates the failure of
the existing primary emission control device (baghouse or
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scrubber) to control emissions adequately. The usual cause of
alarms for existing baghouses is a break or tear in a bag. The
new continuous stack monitors have shown they can detect small
leaks in bags that would have gone unnoticed until a buildup of
material on the in-stack sampler was observed.
Thus, engineered controls to shut down a given process as a
result of using the continuous stack monitor are possible. The
FMPC already has administrative controls to shut down processes
in order to replace leaking bags in the existing baghouses.
However, the reliability of coupling process shutdown to the
continuous stack monitors is presently unknown. In addition, the
FMPC has stated that some processes cannot be shut down during
certain operational phases.
The use of the continuous stack monitor is highly
recommended as a method to detect leaks in bags or excessive
emissions from either the baghouses or scrubbers. However,
installation of the continuous stack monitor cartnot insure
reductions in emissions; secondary particulate emission control
devices are also required.
The continuous stack monitors are best used in their
existing configuration, i.e., real-time detection of emissions
prior to the secondary particulate emission control devices.
This configuration allows rapid detection and repair of
deficiencies in the primary emission control devices and should
reduce the rate of particulate loading on the HEPA filter systems
proposed as the secondary emission controls.
A second in-stack sampler (filter paper collector)
downstream of the final emission control device is also
recommended for uranium inventory control and determination of
actual emissions to the environment. If possible, this in-stack
sampler should be analyzed to correlate with annual reporting
requirements.
2.7.4.2 Estimated Cost for Emission Control Technology
The FMPC has plans to obtain and install 14 additional
continuous stack monitors at an estimated cost of $105K ($7.5K
per continuous stack monitor). The acquisition of 14 additional
continuous stack monitors would allow installation of a
continuous stack monitor on each of the stacks that currently do
not have one, plus on other selected stacks.
A summary of the cost estimates for the acquisition and
installation of the conceptual design HEPA filter systems for
each of the stacks is given in Table 2.7-5.
2.7.4.2.1 Effect of Proposed Equipment
Reductions in emissions from the existing emission control
devices based on the installation of continuous stack monitors
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Table 2.7-5. Cost estimates for acquisition and installation of
HEPA filter systems.
HEPA Filter
Installation Total Cost(a>
Stack No. Cost ($ Thousands) ($ Thousands)
G4-2
G4-5(a)
G4-7
G4-14(a)
G5-249
G5-254
G5-256
G5-260(a)
G5-261
-------
Architect-Engineer services are typically about 25 percent
of all other costs. Thus, total costs for the proposed secondary
emission controls may be expected to be about 2-1/2 times greater
than the costs shown.
A total secondary emission control cost estimate for the
seven stacks is approximately $2.3 million. This estimate is
less than half the estimate provided by the FMPC for the six
stacks having the greatest emissions in 1986. Direct comparison
of the present cost estimates for a specific stack to those of
the FMPC is not possible because FMPC provided no details for its
estimates. The FMPC has estimated a cost of approximately $14M
to install secondary emission controls on all presently operating
stacks (Mo86). In either case, the cost estimates are
approximate values, subject to revision based on additional
information.
2.7.4.2.2 Operation and Maintenance Costs
Addition of continuous stack monitors, as planned by the
FMPC, will result in the need for their periodic maintenance.
These maintenance needs are not expected to be excessive,
although the addition of one full-time-equivalent instrument
technician may be required. Regular operations personnel are
expected to be responsible for standard operation of the
monitors. No unusual operating or maintenance costs are
predicted as a result of the installation of additional
continuous stack monitors.
HEPA filter replacement costs have been estimated to be
$94,000 per year for the seven stacks having the greatest
emissions and $111,000 per year for all fourteen stacks. The
filter replacement cost estimate is based on an average cost of
$350 per filter (stainless steel housing) and the total number of
filters to be replaced per year (Mo86).
The FMPC currently has no facilities to process
uranium-loaded HEPA filters of the size and quantity proposed in
order to recover the uranium. Additional costs for this
operation have not been estimated.
If the HEPA filters are discarded, they would have to be
disposed of as low specific-activity radioactive waste, i.e.,
sent to a low-level radioactive waste burial ground. Costs for
the packaging, transport, and burial of discarded HEPA filters
have not been estimated.
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2.8 BROOKHAVEN NATIONAL LABORATORY
2.8.1 Description and Existing Controls
2.8.1.1 Site Description
Studies conducted at Brookhaven Laboratories pertain to the
use, environmental effects, and transport of both nuclear and
nonnuclear energy materials. Other research programs include
applied nuclear studies involving various radioisotopes and
investigations of the physical, chemical, and biological effects
of radiation. Brookhaven Laboratory is located in the center of
Long Island, about 113 km from New York City.
The equipment and facilities used to support the research
projects conducted at Brookhaven include several reactors,
particle accelerators, and laboratories. Point and area sources
of radionuclide releases at Brookhaven include:
o The 40-MW High-Flux Beam Reactor (HFBR)
o The Alternating Gradient Syncrotron, a proton accelerator
used in ultra-high energy particle physics research
o The Brookhaven Linac Isotope Production Facility (BLIP)
o The Chemistry Linac Irradiation Facility (CLIF)
o The Brookhaven Medical Research Reactor
o The Van de Graaff accelerator
o Various chemistry and medical research laboratories
Most of the airborne radionuclide emissions from Brookhaven
originate from the High-Flux Beam Reactor, the Brookhaven Linac
Isotope Production Facility, and the Van de Graaff research
generator. Lesser emissions are from the chemistry and medical
research centers.
Because very small quantities of radionuclides are released
from the Hazardous Waste Management Area, the assessments of
exposure and health risk at the Brookhaven site are based on
airborne releases from the remaining six effluent stacks.
Process descriptions, effluent data, and site information were
obtained from reports prepared by Brookhaven Laboratories and DOE
studies (Mo84, Mi87b).
2.8.1.2 Major Release Points and Existing Emission
Control Technology
In this section, the points of discharge that contribute
most to the airborne radionuclide emissions at the BNL site are
discussed.
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Table 2.8-1. Radionuclide emission points stacks at Brookhaven
National Laboratories.
Stack
Location Height (m)
Brookhaven Linac Isotope Production Facil., Bldg-931 18
High-Flux Beam Reactor Hot Laboratory 98
Hazardous Waste Management Area 10
Medical Research Reactor Building-491 45
Chemistry Building-555 Unknown
Medical Research Center Unknown
Van de Graaff Accelerator Building-901 18
2.8.1.2.1 HFBR Stack
The principal radionuclides discharged from the HFBR stack
are tritium (from the HFBR) and xenon-127 and small amounts of
unidentified radionuclides that emit beta and gamma radiation
(from the Hot Laboratory). Tritium is the most prevalent
radionuclide discharged.
The HFBR facility (Building 750) is ventilated by about
566 m3/min Of air, all of which is filtered through absolute HEPA
filters to remove particulates and radioactivity before being
discharged from the 98-m stack. In addition, procedural and
administrative controls have been implemented to detect tritium,
prevent its leakage, and reduce the release of tritiated water
vapor from the HFBR stack. Since 1977, yearly replacement of a
portion of the heavy water (moderator and coolant) has reduced
the annual tritiated water vapor released from the HFBR by
approximately 50 percent.
The hot area of the Hot Laboratory (Building 801) consists
of five semihot cells, three chemical processing hot cells, and
three high-level hot cells for handling multicurie amounts of
radioactive materials. Each cell is equipped with its own
roughing exhaust air filter, as well as a backup HEPA filter in
the exhaust line leading to the stack. The three chemical
process cells have a separate exhaust air system that uses a NaOH
scrubber and charcoal filter to remove radioiodines. The small
amount of xenon-127 released is diluted after release from the
stack. All effluents from the Hot Laboratory are exhausted to
the 98-m HFBR stack.
2.8.1.2.2 Brookhaven LINAC Isotope Production Facility
The targets used for the production of desired radionuclides
in the BLIP facility are sealed so that no radioactivity can
escape from them during normal operation. However, oxygen-15 and
tritium are formed by the incident protons in the target cooling
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water. Larger release rates of oxygen-15 in relation to the
other gases result because it is swept out with absorbed oxygen
in the cooling water. The absorbed oxygen is formed by the
radiolytic formation of stable oxygen. The airborne effluents
from the BLIP facility undergo HEPA filtration to remove any
particulates prior to monitoring and release from an 18-m stack.
The oxygen-15 and tritium currently receive no treatment prior to
discharge from the stack (Mo86).
2.8.1.2.3 Brookhaven Medical Research Reactor
The principal radioactive gas discharged during routine
operations of the BMRR is 110-minute half-life argon-41, which is
produced in the cooling air in the reactor's graphite reflectors.
At a full power level of 3 MW, a release rate of about 3 Ci/hr
has been established by direct measurements. The operation of
the BMRR is administratively controlled to a daily limit of
24 MWhr. Currently, it is operated intermittently for
short-lived activation irradiation. The BMRR is enclosed in a
containment building that is maintained under negative pressure
to prevent inadvertent releases to the outside. Air flow from
the building is passed through HEPA and charcoal filters to
remove particulates before being vented to the atmosphere via a
45-m stack.
2.8.1.2.4 Research Van de Graaff Accelerator
The principal radionuclide discharged to the atmosphere from
the Research Van de Graaff Accelerator is tritium. Currently,
about 95 percent of the release is in gaseous form and about 5
percent is tritiated water vapor. The air control system in this
facility is designed to function as a closed system. During
normal operation, a low-pressure pump is used to maintain
negative pressure on the system. The output of this pump is
routed through a catalytic recombiner where the tritium gas is
converted to tritiated water vapor which is passed through a
dessicant for removal. Spent dessicants are periodically removed
and transported offsite for disposal with other low-level solid
waste. When the accelerator is shut down for maintenance, the
negative pressure is removed and air at atmospheric pressure is
allowed to fill the system. Upon completion of maintenance, the
system is pumped down to a negative pressure. During these
times, the flow exceeds the capacity of the recombiner and the
excess flows are routed directly to the stack via a by-pass line.
When tritium ions are being accelerated, about 200 Ci/month of
tritium gas is used. Of the total tritium used, about 50 percent
is trapped by the dessicant and about 50 percent is released from
the 18-m stack attached to Building 901.
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2.8.2 Basis for the Dose and Risk Assessment
2.8.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.8-2.
Table 2.8-2. Radionuclides released to air during 1986 from
Brookhaven National Laboratory.
Nuclide Release Rate (Ci/y)
Ar-41
Ba-133
Be-7
Br-82
C-14
Co-57
Cr-51
Fe-55
H-3
Hg-203
1-125
1-126
1-131
1-133
Mn-54
0-15
P-32
Ru-103
S-35
Sb-122
Se-75
Sn-113
Sn-117m
Tc-99
Tc-99m
Tl-201
Xe-125
Xe-127
Xe-131m
Zn-65
1.2E+3
2.7E-6
1.8E-6
7.8E-3
7.7E-4
2.2E-5
1.1E-4
5.1E-3
1.6E+2
1.2E-6
5.2E-4
3.2E-4
5.1E-4
1.8E-4
l.OE-5
1.5E+2
2.5E-4
1.2E-5
5.7E-4
3.0E-7
2.0E-5
2.0E-4
4.2E-5
l.OE-4
2.0E-4
2.1E-5
8.8E-5
5.7E-4
6.8E-6
1.3E-6
In modeling the site, all releases were aggregated to six
stacks: Stack 1 is Chemistry Building #555, with a stack height
of 17 m; Stack 2 is the Van De Graaff Building 901, with a stack
height of 18 m; Stack 3 is the HFBR Hot Lab, with a stack height
of 98 m; Stack 4 is the Hazardous Waste Management Area, with a
stack height of 10 m; Stack 5 is the MRC Buildings 490 and 491,
with a stack height of 14 m; and Stack 6 is the BLIP Building
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931, with a stack height of 18 m. Default particle sizes (1.00
AMAD) and solubility classes were assumed.
2.8.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Lawrence, New York. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 750 m from the assumed
release point. Food consumption rates appropriate to an urban
location were used.
2.8.3 Results of the Dose and Risk Assessment
The major contributor to exposure is argon-41 (94 percent).
The predominant exposure pathway is air immersion.
The results of the dose and risk assessment are presented in
Tables 2.8-3 through 2.8-5. Table 2.8-3 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.8-4 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.8-5 presents the estimated distribution of fatal cancer risk to
the regional population.
Table 2.8-3
Organ
Estimated radiation dose rates from the Brookhaven
National Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Gonads
Remainder
Breast
Red marrow
Lungs
8.0E-1
6.2E-1
7.2E-1
6.2E-1
6.1E-1
3.8E+0
3.0E+0
3.4E+0
2.9E+0
2.9E+0
Table 2.8-4. Estimated fatal cancer risks from the Brookhaven
National Laboratories.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
2E-5
1E-3
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Table 2.8-5. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the
Brookhaven National Laboratories.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
800
1,800
5,200,000
0
0
0
0
2E-4
6E-5
9E-4
Totals 5,200,000 1E-3
2.8.4 Supplementary Controls
Ninety-four percent of the risk estimated for BNL results
from the release of Argon-41 from the BMRR. Argon-41 emissions
could be reduced by the addition of a hold-up tank to allow the
argon-41 to decay.
2.9 MOUND FACILITY
2.9.1 Description and Existing Controls
2.9.1.1 Site and Release Point Description
The Mound Facility, located in Miamisburg, Ohio, about 16 km
southwest of Dayton, Ohio, has a variety of active programs.
These include research and development, processing of solid
wastes for tritium recovery, fabrication and testing of weapons
components, production of stable isotopes for the market, and
manufacture of radioisotopic heat sources for military and
aerospace applications.
The principal emissions of tritium and plutonium emanate
from nine buildings, designated as HH, SW, H, PP, R, SM, WD, WDA,
and 41. Buildings HH and SW, which contain the tritium recovery
and reprocessing facilities, are the sole release points of
tritium. Plutonium is released from the other facilities as a
result of heat source production and waste disposal operations.
2.9.2 Basis for the Dose and Risk Assessment
2.9.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.9-1.
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Table 2.9-1. Radionuclides released to air during 1986 from
Mound Facility.
Nuclide Release Rate (Ci/y)
H-3 3.6E+3
Pu-238 5.8E-6
Pu-239 1.4E-7
U-234 7.5E-8
U-238 8.4E-8
In modeling the site, all releases were assumed to be made
from a single 61-m stack. Default particle sizes (1.00 AMAD) and
solubility classes were assumed.
2.9.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Dayton,
Ohio. The 0-80 km population distribution was produced using the
computer code SECPOP and 1980 Census Bureau data. Nearby
individuals were located 1,500 m from the assumed release point
(Mi87b). Food consumption rates appropriate to an urban location
were used.
2.9.3 Results of the Dose and Risk Assessment
The major contributor to exposure is tritium (98 percent).
The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.9-2 through 2.9-4. Table 2.9-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.9-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.9-4 presents the estimated distribution of fatal cancer risk to
the regional population.
2.10 IDAHO NATIONAL ENGINEERING LABORATORY
2.10.1 Description and Existing Controls
2.10.1.1 Site Description
The Idaho National Engineering Laboratory is a reactor
testing facility in southeastern Idaho, about 56 km west of Idaho
Falls. The following four contractors operate facilities here:
EGho, Inc.; Allied Chemical Corporation; Argonne West Laboratory;
and Westinghouse Electric Corporation.
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Table 2.9-2,
Organ
Estimated radiation dose rates from the Mound
Facility.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Remainder
Gonads
Breast
Lungs
Red marrow
4.1E-2
3.7E-2
3.7E-2
3.8E-2
3.7E-2
3.3E+0
3 . OE+0
3.0E+0
3. OE+0
3. OE+0
Table 2.9-3. Estimated fatal cancer risks from the Mound
Facility.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
1E-6
3E-3
Table 2.9-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Mound
Facility.
Totals
Number of Persons
2,900,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
0
1,000
2,900,000
0
0
0
0
0
2E-5
3E-3
3E-3
EGc., operates several test reactors. These reactors
provide operating information for the development of reactor
safety programs, for determination of the performance of reactor
materials and equipment, and occasionally, for use in research
performed by private organizations, other activities include
disassembly and reassembly of large radioactive reactor
components, preparation of test specimens for use in various
operating reactors, and waste handling.
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Fuel processing is the major operation that Allied Chemical
conducts at this site. Its Idaho Chemical Processing Plant
stores irradiated fuel and reprocessed fuel and converts
high-level radioactive liquid waste to solid form.
Westinghouse operates the Naval Reactor Facility at the
Idaho Laboratory. This facility serves as a testing area for
prototype naval reactors and as a disassembly and inspection area
for expended reactor cores.
Argonne West operates the experimental Breeder Reactor, the
transient Reactor Test Facility, and the Zero Power Physics
Reactor.
2.10.1.2 Major Release Points and Existing Emission
Control Technology
2.10.1.2.1 Advanced Test Reactor (ATR)
The ATR has an operational thermal-power level rating of
150 MW. It is designed for use in developing advanced cores and
fuel system materials for commercial power programs. The ATR is
a light-water-moderated and cooled system that employs the flux
concentration principle (flux traps) to achieve higher neutron
flux levels.
Ventilation air from the ATR is discharged from a 76-m stack
with no waste treatment system employed. The stack is monitored
on a continuous basis for particle and gaseous activity. Noble
gases, such as argon, krypton, and xenon, are released. The
airflow rate of the stack is 1,275 m3/roin.
2.10.1.2.2 Idaho Chemical Processing Plant (ICPP)
The ICPP is used to process highly enriched-irradiated
nuclear reactor fuel elements in order to recover uranium. Fuel
elements from INEL reactors (test and research), other research
reactors (domestic and foreign), and U.S. Navy ship propulsion
reactors have been reprocessed. Airborne emissions from the ICPP
are largely attributable to off-gases from the process
dissolvers, process vessels, analytical facilities, sample
stations, waste solvent burner, New Waste Calcining Facility
(NWCF), and ventilation air. The New Waste Calcining Facility is
used to convert radioactive liquid waste from the ICPP to a
solid, using a fluidized bed calcination process.
The atmospheric protection system (APS) serves as a final
cleanup facility for most ventilation systems and the process
off-gas systems within the ICPP. The APS is divided into three
treatment sections: (1) ventilation air treatment, (2) nitrogen
oxide-bearing off-gas treatment, and (3) hydrogen-rich off-gas
treatment.
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The vessel off-gas treatment section of the APS facilitates
treatment of the process off-gases from: (1) continuous process
modification dissolver off-gas (CPMDOG), (2) vessel off-gas, and
(3) the New Waste Calcining Facility. This section of the APS
consists of a condenser, demister, superheater, prefilter, final
filter, and blowers. The system is constructed of stainless
steel for acid resistance.
A single-story 15.8 x 6.1 m building attached to the
southeast corner of the HEPA building, CPP-649, contains the APS
cleanup system and blowers for the VOG process off-gases. The
cleanup portion of the system (condenser, demister, superheater,
and prefilter) is in the east part of the building. Some valves
that may require opening or closing during operation are equipped
with reach rods that penetrate the shielding wall.
The demister consists of two 10-cm-thick stainless steel
mesh elements contained in a stainless steel chamber.
The prefilter is constructed of five separate fiberglass
beds supported on stainless steel screens. Contained in a 3.7 x
2.1 x 4.0 m stainless steel housing, the prefilter has a water
line for flushing the filter medium. The prefilter can be
bypassed during flushing. The flush water drains to the process
equipment waste (PEW) evaporator feed tank. The three HEPA
filters are housed in caissons equipped with dampers for
individual filter isolation. The HEPA filters are made of acid-
and moisture-resistant materials. The HEPA filters are equipped
with knife-edge seals to prevent leakage.
Two stainless steel blowers exhaust the VOG streams to the
main stack. Only one blower is required for normal operation.
The operating blower is switched automatically to emergency power
during commercial power outages; the standby blower starts
automatically on failure of the operating blower to maintain
necessary vacuum. The blowers are provided with automatic air
operated valves to isolate the unit not in operation.
The ventilation exhaust filter system, a portion of the APS,
consists of a deep-bed fiberglass prefilter in series with
standard HEPA filters. The prefilter is located in an
underground reinforced concrete vault (CPP-756), measuring 12.2 x
27.4 x 4.3 m. The vault includes a system for backwashing the
prefilter medium. Over-temperature protection for the filters is
provided by a fog-spray system located upstream of the prefilter.
This system actuates on high-gas temperature in the duct and
cools the gas and protects the filters from an in-cell fire. A
bypass duct is provided around the prefilter for use during
washing of the filter medium.
The ventilation air ducts from the various buildings join
before entering the prefilter distribution plenum. The
distribution plenum extends the full length of the west side of
the vault and distributes air, via flow slots, into each of four
bays.
2-75
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The floor of the underground vault is sloped to the north;
four troughs drain condensate or flush water to the north edge of
the vault. From there, another trough carries the water to a
1,893-1 capacity collection sump located in the northeast corner
of the vault. The sump is equipped with a high-level alarm and a
sampler. From the sump, the liquid and associated solids are
jetted to the PEW evaporator feed tank, WL-102.
The south wall of the vault has six viewing ports for
inspection of the vault and filters. No lights are provided in
the vault;-portable lighting is used when needed.
The roof of the vault is 0.3 m below grade and covered with
about 0.6 m of earth for radiation shielding. The roof and earth
cover are sloped to allow proper drainage, and the vault is of
leaktight construction. The cracks between the removable
interlocking concrete blocks are caulked, and a butyl rubber
membrane covers the entire roof of the vault. Insulation board
overlays the membrane to prevent damage by the soil.
The prefilter has an area of 279 m2 and has a maximum flow
rate of 4,245 m/min. The prefilter is designed for gas upflow
through five layers of varying density, separately supported,
packed fiberglass. The five individual layers are separated and
supported by stainless steel wire screens. The screens are
mounted on Amercoat-painted carbon steel frames and wired to
support pipes spaced at 0.9-m intervals. The prefilter frame is
attached to Unistrut embedded in the concrete walls; voids in the
Unistrut and other openings are caulked with fiberglass to
prevent bypassing of the filter medium.
Water spray systems are provided to flush particulates from
the fiberglass deep-bed prefilters if the pressure drop becomes
excessive. There are three spray lines, located at different
elevations, to provide thorough washing of the filter medium.
Each of the three spray lines consists of five 1.2-cm
diameter Type 304 stainless steel pipes; the bottom line is
equipped with spray nozzles directed upward and the two upper
lines have holes drilled in the lower portion of the pipes to
supply flush water to the filter. To reduce water supply and
removal requirements for flushing the ventilation air prefilter,
flushing is done in sections. The spray system piping is stubbed
off outside the ventilation air prefilter vault for later
connection to a water supply, if required. The fiberglass
deep-bed prefilters will not require replacement during the
design lifetime of 20 years (from 1975). However, with the
estimated dust loading in the ventilation air, the prefilter
should last about 75 years without flushing or replacement.
Ventilation air from the prefilter is discharged through a
concrete duct to the HEPA filters located in a building adjacent
to the prefilter vault. The two-story reinforced concrete
structure measures 23.5 x 10.1 and is 7.9 m high. The first
.story of the structure begins 2.4 m below grade.
2-76
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The HEPA filters are 9.4 x 9.4 x 4.3 cm units, each rated at
42.5 m3/min, with an initial pressure drop of 2.5 cm of water.
The filters are housed in caissons for ease of maintenance and
filter replacement.
From the HEPA filters, the ventilation air flows through
three ventilation fans and is exhausted to the stack. The
ventilation fans are direct drive and installed in parallel to
provide the motive force for discharging the ventilation air to
the stack. The fans are housed in a 6.6 x 14.6 m addition on the
east side of the existing fan building (CPP-605). The fans are
of carbon steel construction with backward airfoil blades.
During normal operation, one or two of the three fans is operated
on commercial power. If the operating fan fails during normal
operation, the second and third fans can be started manually on
commercial power. Automatic switching of an operating fan to
emergency power, during commercial power outages, is provided by
manual preselection. Each fan is provided with a damper that
closes automatically if the fan stops. The dampers can be opened
either with a wrench or via a pressurized N2 system if the need
arises.
2.10.2 Basis for the Dose and Risk Assessment
2.10.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.10-1.
In modeling the site, all releases were assumed to be made
from the ICPP, since this is the major source of uranium. The
releases were assumed from aim stack. Default particle sizes
(1.00 AMAD) and solubility classes (Class W for antimony-125)
were assumed.
2.10.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Pocatello, Idaho. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 15,000 m from the assumed
release point (Ho87). Food consumption rates appropriate to a
rural location were used.
2.10.3 Results of the Dose and Risk Assessment
The major contributors to exposure are argon-41 (51
percent), antimony-125 (32 percent), and krypton-88 (8 percent).
The predominant exposure pathways are air immersion for argon-41
and ground surface for antimony-125 and krypton-88.
2-77
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Table 2.10-1.
Radionuclides released to air during 1986 from all
Idaho Facilities.
Nuclide
Release Rate (Ci/y)
Ar-41
Ba-139
Ba-140
Br-82
C-14
Co-60
Cs-134
Cs-137
Cs-138
Gd-153
H-3
Hg-203
1-129
1-131
Kr-85
Kr-85m
Kr-87
Kr-88
La-140
Mn-54
Nb-95
Pu-238
Ru-103
Sb-125
Se-75
Sr-85
Sr-90
Te-132
Xe-133
Xe-135
Xe-135m
Xe-138
Y-90
1.9E+3
7.5E+0
1.8E-6
l.OE-2
3.3E-1
4.4E-4
l.OE-4
2.4E-3
9.4E-1
9.8E-6
3.6E+1
1.4E-4
1.8E-1
7.4E-4
1.1E+4
7.1E+1
1.5E+2
1.6E+2
1.8E-6
8.7E-5
5.2E-7
1.6E-5
2.0E-7
9.3E-1
1.1E-4
3.2E-8
1.9E-6
6.0E-8
5.2E+2
4 . 1E+2
3.2E+0
4 . 1E+2
3.1E-8
The results of the dose and risk assessment are presented in
Tables 2.10-2 through 2.10-4. Table 2.10-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.10-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.10-4 presents the estimated distribution of fatal cancer risk
to the regional population.
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Table 2.10-2
Organ
Estimated radiation dose rates from the Idaho
National Engineering Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Gonads
Remainder
Breast
Lungs
Red marrow
2.9E-2
2.3E-2
2.7E-2
2.4E-2
2.3E-2
7.3E-2
6.3E-2
6.8E-2
6.1E-2
5.7E-2
Table 2.10-3. Estimated fatal cancer risks from the Idaho
National Engineering Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
6E-7
2E-5
Table 2.10-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from INEL
facilities.
Totals
Number of Persons
100,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
0
0
100,000
0
0
0
0
0
0
2E-5
2E-5
2.11 LAWRENCE BERKELEY LABORATORY
2.11.1 Description and Existing Controls
2.11.1.1 Site Description
Lawrence Berkeley Laboratory (LBL) is situated upon a
hillside above the main campus of the University of California,
Berkeley. The 130-acre site is located on the west-facing slope
2-79
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of the Berkeley Hills, at elevations ranging from 500 to
1,500 feet above sea level. LBL is located in an urban
environment on land owned by the University. The LBL site is
bordered on the north by predominately single family homes and on
the west by multiunit dwellings, student residence halls, and
commercial districts. The population within an 80-km radius of
the Laboratory is approximately 5.2 million (1980 census).
The Laboratory's activities are located both onsite and
offsite. There are 67 buildings on the LBL hillside site, plus
additional facilities located on the University campus, notably
the Donner Laboratory of Biology and Medicine and the Melvin
Calvin Laboratory. The onsite space consists of 1,350,000 gross
square feet (gsf) in about 60 buildings: 1,307,000 in DOE
buildings and trailers and 43,000 in University-owned buildings.
These facilities include four large accelerators, several
small accelerators, several radiochemical laboratories, and the
Tritium Labeling Laboratory. The large accelerators include the
Bevatron, the Super HILAC, the 224-cm Sector-Focused Cyclotron
and the 467-cm Cyclotron. '
The tritium facility was designed to accommodate kilocurie
quantities of tritium as a labeling agent for chemical and
biomedical research. Radiochemical and radiobiological studies
in many laboratories typically use millicurie quantities of
various radionuclides.
2.11.1.2 Major Release Points and Existing Emission
Control Technology
Each laboratory box exhaust system includes a group of HEPA
filters and/or gas traps. The tritium facility has a tritium
recovery system in which unused tritium gas is circulated over
hot copper oxide and the resultant water is trapped in a liquid
nitrogen dewar, drained from the system, and packaged for
disposal. This recovery system can be isolated from the labeling
and storage system, and the tritium can be circulated
continuously in a closed loop until the tritium concentration has
dropped to an acceptable level for discharge to the atmosphere
via the laboratory exhaust manifold. Silica gel traps are used
to reduce the level of tritium discharged.
The purge ventilation system of the LBL tritium facility
consists of an air evacuation system that draws air through
inside filters into a vent pipe to the outside of the facility
where it then undergoes mechanical forcing. This forcing vents
the air through a vertical exhaust stack elevated 9 m above a
hill directly behind the facility, giving an effective stack
height of 18.3 m.
2-80
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2.11.2 Basis for the Dose and Risk Assessment
2.11.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.11-1.
Table 2.11-1. Radionuclides released to air during 1986 from
Lawrence Berkeley Laboratory.
Nuclide
Release Rate (Ci/y)
H-3
1-125
1-131
PU-239
Sr-90
7.6E+1
3.7E-3
1.2E-3
7.4E-9
5.8E-5
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes were assumed.
2.11.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Oakland,
California. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data.
Nearby individuals were located 250 m from the assumed release
point (Sc87). Food consumption rates appropriate to an urban
location were used.
2.11.3 Results of the Dose and Risk Assessment
The major contributor to exposure is tritium (90 percent).
The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.11-2 through 2.11-4. Table 2.11-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.11-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.11-4 presents the estimated distribution of fatal cancer risk
to the regional population.
2-81
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Table 2.11-2,
Organ
Estimated radiation dose rates from the Lawrence
Berkeley Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Remainder
Gonads
Red marrow
Breast
Lungs
1.9E-2
1.8E-2
2.5E-2
1.8E-2
1.8E-2
7.8E-1
7.0E-1
l.OE+0
7.0E-1
7.0E-1
Table 2.11-3.
Estimated fatal cancer risks from the Lawrence
Berkeley Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/year
5E-7
3E-4
Table 2.11-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the
Lawrence Berkeley Laboratory.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
5,000,000
5,000,000
0
0
0
0
0
0
3E-4
3E-4
2.12 PADUCAH GASEOUS DIFFUSION PLANT
2.12.1 Site Description
The DOE operation at the Paducah Gaseous Diffusion Plant
consists of a uranium enrichment facility and a uranium
hexafluoride manufacturing complex. The plant is located 6 km
south of the Ohio River in McCrasken County, Kentucky.
2-82
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The primary activity at this site is the diffusion cascade
for the enrichment of uranium in fissionable uranium-235 content.
All stages of the enrichment cascade take place in five buildings
on the site. The manufacturing facility produces uranium
hexafluoride from uranium oxide feedstocks.
2.12.2 Basis for the Dose and Risk Assessment
2.12.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.12-1.
Table 2.12-1. Radionuclides released to air during 1986 from
Paducah Gaseous Diffusion Plant.
Nuclide Release Rate (Ci/y)
Tc-99 8.8E -3
U-234 1.8E -4
U-238 1.8E -4
In modeling the site, all releases were assumed to be made
from a 10-m stack, with a flow of 200 cfm. Default particle
sizes (1.00 AMAD) and solubility classes (Class Y for uranium-234
and uranium-238) were assumed.
2.12.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Paducah/Barkley, Kentucky. The 0-80 km population distribution
was produced using the computer code SECPOP and 1980 Census
Bureau data. Nearby individuals were located 1,500 m from the
assumed release point (Mo86). Rural food consumption rates were
used.
2.12.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-234 and
uraniura-238 (99 percent). The predominant exposure pathway for
both is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.12-2 through 2.12-4. Table 2.12-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.12-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population.
Table 2.12-4 presents the estimated distribution of fatal cancer
risk to the regional population.
2-83
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Table 2.12-2,
Organ
Estimated radiation dose rates from the Paducah
Gaseous Diffusion Plant.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
2.5E-1
3.1E-1
Table 2.12-3. Estimated fatal cancer risks from the Paducah
Gaseous Diffusion Plant.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
4E-7
1E-5
Table 2.12-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Paducah
Gaseous Diffusion Plant.
Totals
Number of Persons
500,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
—
-
-
-
-
—
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
0
0
500,000
0
0
0
0
0
0
1E-5
1E-5
2.13 LAWRENCE LIVERMORE LABORATORY
2.13.1 Site Description
The Lawrence Livermore National Laboratory, situated 64 km
east of San Francisco, California, is primarily a nuclear weapons
research and development center. Other activities, however,
include research programs in laser isotope separation, laser
fusion, magnetic fusion, biomedical studies, and nonnuclear
energy.
Two accelerators, the Insulated Core Transfer Accelerator
and the Electron Positron Linear Accelerator, are used in support
2-84
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of the fusion and neutron physics research programs. The Light
Isotope Handling Facility supports research in the area of light
isotopes. The remaining facilities at this site deal with
equipment decontamination and waste disposal.
2.13.2 Basis for the Dose and Risk Assessment
2.13.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.13-1.
Table 2.13-1. Radionuclides released to air during 1986 from
Lawrence Livermore Laboratory/Sandia Livermore.
Nuclide Release Rate (Ci/y)
H-3 1.8E+3
N-13 9.0E+1
0-15 9.0E+1
Pu-239 7.0E-9
Sr-90 1.3E-7
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes were assumed.
2.13.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Fairfield/Travis, California. The 0-80 km population
distribution was produced using the computer code SECPOP and 1980
Census Bureau data. Nearby individuals were located 3,500 m from
the assumed release point (Mo86). Food consumption rates
appropriate to a rural location were used.
2.13.3 Results of the Dose and Risk Assessment
The major contributor to exposure is tritium (98 percent).
The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.13-2 through 2.13-4. Table 2.13-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.13-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.13-4 presents the estimated distribution of fatal cancer risk
to the regional population.
2-85
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Table 2.13-2,
Estimated radiation dose rates from Lawrence
Livermore Laboratory/Sandia Livermore.
Organ
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Remainder
Gonads
Breast
Lungs
Red marrow
1.1E-2
1.1E-2
1.1E-2
1.1E-2
1.1E-2
4.2E+0
3.7E+0
3.7E+0
3.8E+0
3.7E+0
Table 2.13-3. Estimated fatal cancer risks from Lawrence
Livermore Laboratory/Sandia Livermore.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
3E-7
1E-3
Table 2.13-4. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from Lawrence
Livermore Laboratory/Sandia Livermore.
Risk Interval
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
5,300,000
5,300,000
0
0
0
0
0
0
1E-3
1E-3
2.14 PORTSMOUTH GASEOUS DIFFUSION PLANT
2.14.1 Description and Existing Controls
2.14.1.1 Site and Release Point Description
The Portsmouth Gaseous Diffusion Plant, situated in Pike
County, Ohio, about 1.6 km east of the Scioto River, is operated
2-86
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by Goodyear Atomic Corporation. The primary activity at this
site is the diffusion cascade for the enrichment of uranium in
fissionable uranium-235 content. All stages of the enrichment
cascade take place in five buildings on the site. The
manufacturing facility produces uranium hexafluoride from uranium
oxide feedstocks.
The most significant release point, which accounts for about
84 percent of total emissions, is the X-326 Top Purge Vent.
The DOE Effluent Information System Report for 1986
identifies the following major specific sources for the
Portsmouth Plant: the X-326 Building Top and Side Purge Vent, the
X-330 Building Cold Recovery Facility, and the X-333 Building
Cold Recovery Facility (EIS86).
The radioisotopes in these releases are uranium and its
daughters plus technetium-99, a long-lived fission product. The
technetium-99 results from introducing uranium feed from
reprocessed irradiated nuclear reactor fuel.
2.14.1.2 Emission Control Technology
The main control technologies presently used at Portsmouth
are:
o Cold trapping (the UF6 is removed by freezing)
o Sodium fluoride absorption
o Activated alumina absorption
These methods are primarily useful in preventing the release
of uranium. They are also effective on uranium decay daughters
and on the fission-product isotope technetium-99.
The X-326 Purge Vent is the major source of radionuclide
emissions to the atmosphere at Portsmouth. The existing control
device is the purge cascade itself, which removes the bulk of the
UF6. The remaining light gases are sent through an alumina trap
and diluted with an air jet exhauster before venting.
There are four purge vents. Each vent is 23 m high, 47 cm
apart. The diameter of each vent is 10 cm. Each vent has a flow
rate of 4.72 x 10-2 m3/s at ambient temperature.
2.14.2 Basis for the Dose and Risk Assessment
2.14.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.14-1.
2-87
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Table 2.14-1. Radionuclides released to air during 1986 from
the Portsmouth Gaseous Diffusion Plant.
Nuclide Release Rate (Ci/y)
Pa-234m 1.4E-2
Tc-99 1.2E-1
Th-234 1.4E-2
U-234 2.3E-2
U-235 1.2E-3
U-236 3.4E-5
U-238 1.4E-2
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) were
assumed, and the uranium was assumed to have a D solubility
class.
2.14.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Huntington, West Virginia. The 0-80 km population distribution
was produced using the computer code SECPOP and 1980 Census
Bureau data. Nearby individuals were located 1,500 m from the
assumed release point (Oa87a). Food consumption rates
appropriate to a rural location were used.
2.14.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-234 and
uranium-238 (96 percent). The predominant exposure pathway for
both is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.14-2 through 2.14-4. Table 2.14-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.14-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.14-4 presents the estimated distribution of fatal cancer risk
to the regional population.
2-88
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Table 2.14-2
Organ
Estimated radiation dose rates from the Portsmouth
Gaseous Diffusion Plant.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Endosteum
Remainder
Red marrow
3.4E-1
3.0E-2
2.3E-2
5.7E+0
7.7E-1
4.0E-1
Table 2.14-3. Estimated fatal cancer risks from the Portsmouth
Gaseous Diffusion Plant.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
2E-7
9E-5
Table 2.14-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the
Portsmouth Gaseous Diffusion Plant.
Number of Persons
Totals
620,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
0
0
620,000
0
0
0
0
0
0
9E-5
9E-5
2.15 ARGONNE NATIONAL LABORATORY
2.15.1 Site Description
Argonne National Laboratory is an energy research and
development center that performs investigations in basic physics,
chemistry, materials science, the environmental sciences, and
biomedicine. Argonne also plays an important role as a nuclear
and nonnuclear engineering center. The laboratory complex is
located in Dupage County, Illinois, 43 km southwest of Chicago.
2-89
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Argonne National Laboratory has the following principal
nuclear facilities:
(1) 10- and 200-kW research reactors
(2) A critical assembly reactor
(3) A 60-inch cyclotron
(4) A prototype, superconducting, heavy ion linear
accelerator
(5) Van de Graaff and Dynamitron-type charged-particle
accelerators
(6) A high-energy neutron source
(7) Cobalt-60 irradiation sources
(8) Laboratories engaged in work with multicurie quantities
of the actinide elements
The 200-kW JANUS research reactor and the laboratory
handling area (hot cells) are the main sources of radionuclide
releases from the Argonne complex.
Specific details of the site activities and emissions are
available from annual emission reports prepared by the
laboratory, the DOE Effluent Information System, and
environmental monitoring studies conducted by DOE (Mo84, EPA84,
EIS86).
2.15.2 Basis for the Dose and Risk Assessment
2.15.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.15-1.
Table 2.15-1. Radionuclides released to air during 1986 from
Argonne National Laboratory.
Nuclide Release Rate (Ci/y)
Ar-41
C-ll
Cs-134
Cs-137
H-3
1-129
1-131
Kr-85
Nb-95
Pu-239
Rn-220
Sb-125
Zr-95
1.4E+0
9.0E+1
2.0E-7
4.9E-7
5.0E+1
1.6E-5
1.5E-6
1.7E+0
1.5E-8
5.6E-9
7.0E+3
3.4E-5
7.5E-9
2-90
-------
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class D for carbon-11) were assumed.
2.15.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Midway
Airport, Illinois. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data.Nearby individuals were located 750 m from the assumed
release point. Urban food consumption rates were used.
2.15.3 Results of the Dose and Risk Assessment
The major contributors to exposure are carbon-11 and
tritium. The predominant exposure pathway is inhalation for
carbon-11 and air immersion for tritium.
The results of the dose and risk assessment are presented in
Tables 2.15-2 through 2.15-4. Table 2.15-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.15-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.15-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.15-2. Estimated radiation dose rates from the Argonne
National Laboratory.
Organ Nearby Individuals Regional Population
(mrem/y) (person-rem/y)
Lungs
Remainder
3.1E-2
2.7E-3
2.5E-1
2.1E-1
Table 2.15-3. Estimated fatal cancer risks from the Argonne
National Laboratory.
Nearby Individuals Regional (0-80 km) Population
Lifetime Fatal Cancer Risk Deaths/y
1E-7 8E-5
2-91
-------
Table 2.15-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Argonne
National Laboratory.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
7,900,000
7,900,000
0
0
0
0
0
0
8E-5
8E-5
2.16 PINELLAS PLANT
2.16.1 Site Description
The Pinellas Plant, located 10 km northwest of
St. Petersburg, Florida, is a major facility engaged in the
production of nuclear weapons. Although descriptions of the
principal operations resulting in atmospheric releases of
radioactive materials could not be found in the literature, they
are 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. The heat sources are
triple-encapsulated to prevent release of plutonium to the
atmosphere.
Emissions of radionuclides were identified from three
sources: the main stack, laboratory stack, and building stack.
2.16.2 Basis for the Dose and Risk Assessment
2.16.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.16-1.
Table 2.16-1. Radionuclides released to air during 1986 from
Pinellas Plant.
Nuclide
Release Rate (Ci/y)
H-3
Kr-85
1.9E+2
4.6E+0
2-92
-------
In modeling the site, all releases were assumed to be made
from a 10-m stack.
2.16.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Tampa,
Florida. The 0-80 km population distribution was produced using
the computer code SECPOP and 1980 Census Bureau data. Nearby
individuals were located 1,500 m from the assumed release point.
Food consumption rates appropriate to a rural location were used.
2.16.3 Results of the Dose and Risk Assessment
The major contributor to exposure is tritium (100 percent).
The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.16-2 through 2.16-4. Table 2.16-2 presents the
dosesreceived by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.16-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.16-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.16-2. Estimated radiation dose rates from the Pinellas
Plant.
Organ Nearby Individuals Regional Population
(mrem/y) (person-rem/y)
Remainder
Gonads
Breast
Lungs
Red marrow
4.7E-3
4.4E-3
4.4E-3
4.4E-3
4.3E-3
5.3E-1
4.7E-1
4.7E-1
4.7E-1
4.7E-1
Table 2.16-3. Estimated fatal cancer risks from the Pinellas
Plant.
Nearby Individuals Regional (0-80 km) Population
Lifetime Fatal Cancer Risk Deaths/y
1E-7 2E-4
2-93
-------
Table 2.16-4..
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the
Pinellas Plant.
Number of Persons
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
Totals
0
0
0
0
0
0
1,900,000
1,900,000
0
0
0
0
0
0
2E-4
2E-4
2.17 NEVADA TEST SITE
2.17.1 Site Description
The Nevada Test Site lies about 100 km northwest of Las
Vegas, Nevada, in Nye County. This facility, which is part of
DOE|sweapons research and development complex, is responsible for
design, maintenance, and testing of nuclear weapons. Other
activities at this site include development of new nuclear energy
technologies and radioactive waste disposal.
Radionuclide emissions result primarily from underground
tests of nuclear weapons. Sources of these releases include
drill-back operations, tunnel ventilation, leakage of gases from
underground test sites, and resuspension of contaminated soils.
2.17.2 Basis for the Dose and Risk Assessment
2.17.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.17-1.
In modeling the site, all releases were assumed to be made
from a single point source, since the nearest individual is 70 km
from the site (Mo86). The releases were assumed from a 10-m
stack. Default particle sizes (1.00 AMAD) and solubility classes
were assumed.
2-94
-------
Table 2.17-1.
Radionuclides released to air during 1986 from the
Nevada Test Site.
Nuclide
Release Rate (Ci/y)
H-3
1-131
1-133
Kr-85
Xe-133
Xe-133M
Xe-135
1.2E+2
2.4E+0
9.6E-6
4.3E+0
3 . 6E+4
5.8E-2
4.1E-2
2.17.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Yucca
Flats, Nevada. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data.
Nearby individuals were located 70,000 m from the assumed release
point. Food consumption rates appropriate to a rural location
were used.
2.17.3 Results of the Dose and Risk Assessment
The major contributors to exposure are xenon-133 (81
percent) and tritium (10 percent). The predominant exposure
pathways are air immersion and ingestion.
The results of the dose and risk assessment are presented in
Tables 2.17-2 through 2.17-4. Table 2.17-2 presents the
dosesreceived by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.17-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure as well
as estimated deaths per year in the regional population. Table
2.17-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.17-2.
Organ
Estimated radiation dose rates from the Nevada
Test Site.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Gonads
Remainder
Breast
Thyroid
5.3E-3
3.5E-3
6.5E-3
1.9E-2
1.2E-2
8.1E-3
1.5E-2
5.7E-2
2-95
-------
Table 2.17-3. Estimated fatal cancer risks from the Nevada Test
Site.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
1E-7
3E-6
Table 2.17-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Nevada
Test Site.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
3,500
3,500
0
0
0
0
0
0
3E-6
3E-6
2.18 KNOLLS LABORATORY - KESSELRING
2.18.1 Site Description
The Kesselring site, occupying a 1,579-ha site, is located
near West Milton, New York, approximately 27 km north of
Schenectady. The surrounding area is rural and sparsely
populated; about 1.08 million people live within 80 km.
The Kesselring site has four pressurized water reactor
plants and associated support facilities used for training.
Particulate and gaseous activity contained in the primary coolant
may become airborne from reactor coolant discharges and sampling
operations and during laboratory operations.
At the Kesselring site, exhaust air from reactor coolant
discharges, sampling, and laboratory operations is passed through
HEPA filters, monitored, and released from elevated stacks.
2.18.2 Basis for the Dose and Risk Assessment
2.18.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.18-1.
2-96
-------
Table 2.18-1. Radionuclides released to air during 1986 from
Knolls Atomic Power Lab-Kesselring.
Nuclide Release Rate (Ci/y)
Ar-41
C-14
Co-60
H-3
Kr-83m
Kr-85
Kr-85m
Kr-87
Kr-88
Xe-131m
Xe-133
Xe-135
1.6E-1
3.4E-1
3.4E-6
8.0E-2
7.0E-4
2.0E-6
2.0E-3
1.9E-3
4.0E-3
9.2E-4
2.2E-2
2.3E-2
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class Y for cobalt-60) were assumed.
2.18.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Albany/CO, New York. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 250 m from the assumed
release point (Mo86). Food consumption rates appropriate to an
urban location were used.
2.18.3 Results of the Dose and Risk Assessment
The major contributors to exposure are argon-41
(69 percent), cobalt-60 (12 percent), and carbon-14 (7 percent).
The predominant exposure pathways are air immersion for argon-41
and cobalt-60, and ground surface for carbon-14.
The results of the dose and risk assessment are presented in
Tables 2.18-2 through 2.18-4. Table 2.18-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.18-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population.
Table 2.18-4 presents the estimated distribution of fatal cancer
risk to the regional population.
2-97
-------
Table 2.18-2,
Organ
Estimated radiation dose rates from the Knolls
Lab-Kesselring.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Gonads
Remainder
Breast
Red marrow
Lungs
2.5E-3
3.8E-3
4.4E-3
6.9E-3
2.5E-3
1.5E-2
3.2E-2
3.7E-2
6.5E-2
1.8E-2
Table 2.18-3. Estimated fatal cancer risks from the Knolls
Lab-Kesselring.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
1E-7
2E-5
Table 2.18-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from Knolls
Atomic Power Lab-Kesselring.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
1,200,000
1,200,000
0
0
0
0
0
0
2E-5
2E-5
2.19 BATTELLE COLUMBUS LABORATORY
2.19.1 Site Description
Battelle Columbus Laboratory (BCL) conducts various
NRC-licensed activities, as well as activities under Department
of Energy contracts (Sw87).
2-98
-------
BCL operates two complexes in the Columbus Ohio, area. The
first site is the King Avenue Site, which consists of 4 ha near a
residential area in Columbus. The Ohio State University
intramural sports practice field borders the site to the north.
The second site is the Nuclear Sciences Area of the West
Jefferson site, which is located about 27 km west of the King
Avenue laboratories. This site occupies about 5 ha on a 405-ha
tract of land. Approximately 1.5 million people live within
80 km of the laboratory.
The King Avenue site has a uranium-235 processing facility
located within Building 3. This building also houses the melting
facility and powder metallurgy laboratory. The uranium
processing facility manages all transactions involving nuclear
material at the King Avenue site. However, handling of contract
and licensed material has been very limited since 1977, and
monitoring of airborne emissions was discontinued in 1975.
At the West Jefferson site, activities at the Nuclear
Sciences Area include operations in the JN-1 hot cell (where
irradiated reactor fuel elements are studied) and materials
accountability and storage operations, conducted at the JN-2
vault. The JN-4 plutonium laboratory, where research was
conducted on uranium-235/plutonium-239 nitride reactor fuel, is
being decommissioned.
2.19.2 Basis for the Dose and Risk Assessment
2.19.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.19-1.
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class D for K-40, Class Y for uranium-235 and
plutonium-239) were assumed.
2.19.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Columbus, Ohio. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data.
Nearby individuals were located 750 m from the assumed release
point (Mo86). Food consumption rates appropriate to an urban
location were used.
2-99
-------
Table 2.19-1. Radionuclides released to air during 1986 from
Battelle Columbus.
Nuclide Release Rate (Ci/y)
Ac-228
Be-7
Bi-214
Co-57
Co-60
Cs-134
Cs-137
1-131
K-40
Kr-85
Pb-212
Pb-214
Pu-239
Sb-125
Sr-90
Tl-208
U-235
l.OE-5
1.2E-5
2.4E-5
1.4E-6
3.7E-6
1.5E-6
3.0E-6
8.7E-7
3.0E-4
7.6E+0
3.0E-6
1.5E-5
4.0E-7
5.1E-6
5.8E-7
2.6E-6
2.6E-6
2.19.3 Results of the Dose and Risk Assessment
The major contributors to exposure are potassium-40
(61 percent), uranium-235 (24 percent), and plutonium-239 (10
percent). The predominant exposure pathways are ground surface
for potassium-40 and inhalation for uranium-235 and
plutonium-239.
The results of the dose and risk assessment are presented in
Tables 2.19-2 through 2.19-4. Table 2.19-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.19-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population.
Table 2.19-4 presents the estimated distribution of fatal cancer
risk to the regional population.
2.20 FERMI NATIONAL LABORATORY
2.20.1 Site Description
The Fermi National Accelerator Laboratory is principally
involved with basic research in high-energy physics. Another
important activity involves the treatment of cancer patients with
neutrons released by the second stage of the accelerator. The
Fermi complex is located east of Batavia, Illinois, in the
greater Chicago area.
2-100
-------
Table 2.19-2,
Organ
Estimated radiation dose rates from the Battelle
Columbus Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
Gonads
Remainder
Breast
3.1E-3
8.7E-4
7.2E-4
7.8E-4
1.5E-2
6.2E-3
5.2E-3
5.7E-3
Table 2.19-3. Estimated fatal cancer risks from the Battelle
Columbus Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
2E-8
3E-6
Table 2.19-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from Battelle
Columbus.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
1,900,000
1,900,000
0
0
0
0
0
0
3E-6
3E-6
The accelerator at the Fermi Laboratory, a proton
synchrotron, routinely operates at energies up to 400 GeV
(billion electron volts). The proton beams produced in the
accelerator are used in three different onsite experimental
facilities: (1) the Meson area, (2) the Neutrino area, and (3)
the Proton area. Radionuclides are produced in these areas and
by the accelerator when either the proton beam itself or
secondary particles interact with air.
Another source of radionuclides at Fermi Laboratory is a
magnet-debonding oven, where failed magnets for the accelerator
2-101
-------
are baked at high temperatures to break down the adhesives that
help form the magnets.
2.20.2 Basis for the Dose and Risk Assessment
2.20.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.20-1.
Table 2.20-1. Radionuclides released to air during 1986 from
Fermi National Accelerator Laboratory.
Nuclide Release Rate (Ci/y)
C-ll 3.4E+0
H-3 3.0E-3
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class D for C-ll) were assumed.
2.20.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Midway
Airport, Illinois. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 1,500 m from the assumed
release point (Ba87, Mo84). Food consumption rates appropriate
to a rural location were used.
2.20.3 Results of the Dose and Risk Assessment
The major contributor to exposure is carbon-11 (100
percent). The predominant exposure pathway is air immersion.
The results of the dose and risk assessment are presented in
Tables 2.20-2 through 2.20-4. Table 2.20-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.20-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.20-4 presents the estimated distribution of fatal cancer risk
to the regional population.
2-102
-------
Table 2.20-2
Organ
Estimated radiation dose rates from the Fermi
National Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Gonads
Remainder
Breast
Lungs
Red marrow
9.2E-4
7.1E-4
8.6E-4
9.1E-4
7.0E-4
4.1E-3
3.2E-3
3.9E-3
4.1E-3
3.2E-3
Table 2.20-3. Estimated fatal cancer risks from the Fermi
National Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
2E-8
1E-6
Table 2.20-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Fermi
National Laboratory.
Number of Persons
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
Totals
0
0
0
0
0
0
7,700,000
7,700,000
0
0
0
0
0
0
1E-6
1E-6
2.21 SANDIA NATIONAL LABORATORY
2.21.1 Site Description
The operations at Sandia National Laboratories near
Albuquerque, New Mexico, include weapons testing, arming and
fusing nuclear weapons, and developing modifications to delivery
systems (De87, Mo84). The major facilities include the Sandia
Pulsed Reactor and the Annular Core Pulsed Reactor (both of which
2-103
-------
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, all of which are located in
Technical Areas (TA) I and V. TA-I, in the northwest corner of
the site, also houses research and design laboratories. TA-III
is the site of the Sandia low-level radioactive waste dump.
2.21.2 Basis for the Dose and Risk Assessment
2.21.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/yr, from all sources
during 1986 are listed in Table 2.21-1.
Table 2.21-1. Radionuclides released to air during 1986 from
Sandia National Laboratory/Lovelace Research
Institute.
Nuclide Release Rate (Ci/y)
Ar-41 5.5E+0
H-3 1.3E-1
Pb-212 8.5E-3
In modeling the site, all releases were assumed to be from a
10-m stack. Default particle sizes (1.00 AMAD) and solubility
classes (Class D for lead-212) were assumed.
2.21.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Albuquerque/Sunpt, New Mexico. The 0-80 km population
distribution was produced using the computer code SECPOP and 1980
Census Bureau data. Nearby individuals were located 3,500 m from
the assumed release point. Urban food consumption rates were
used.
2.21.3 Results of the Dose and Risk Assessment
The major contributors to exposure are argon-41 (74 percent)
and lead-212 (26 percent). The predominant exposure pathways are
air immersion for argon-41 and inhalation for lead-212.
The results of the dose and risk assessment are presented in
Tables 2.21-2 through 2.21-4. Table 2.21-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.21-3 presents the estimated lifetime fatal
2-104
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cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.21-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.21-2.
Organ
Estimated radiation dose rates from the Sandia
National Laboratory/Lovelace Research Institute.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Remainder
Gonads
Lungs
Breast
Red marrow
5.3E-4
5.9E-4
1.2E-3
5.4E-4
5.6E-4
1.9E-2
2.1E-2
4.9E-2
1.9E-2
2 . 1E-2
Table 2.21-3.
Estimated fatal cancer risks from the Sandia
National Laboratory/Lovelace Research Institute.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
1E-8
8E-6
Table 2.21-4.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Sandia
National Laboratory/Lovelace Research Institute.
Risk Interval
Totals
Number of Persons
500,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
0
0
500,000
0
0
0
0
0
0
8E-6
8E-6
2-105
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2.22 BETTIS ATOMIC POWER LABORATORY
2.22.1 Site Description
The Bettis Atomic Power Laboratory is situated on an 0.8 km2
tract in West Mifflin, Pennsylvania, approximately 12 km south of
Pittsburgh. This facility designs and develops nuclear power
reactors.
2.22.2 Basis for the Dose and Risk Assessment
2.22.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.22-1.
Table 2.22-1. Radionuclides released to air during 1986 from
Bettis Atomic Power Laboratory.
Nuclide Release Rate (Ci/y)
Co-60
Cs-137
1-129
1-131
Kr-85
Rn-220
Sb-125
Sr-90
U-234
U-238
Xe-131m
Xe-133
1.7E-6
1.7E-6
1.8E-6
6.9E-6
9.4E-1
6i3E-2
3.1E-5
1.7E-6
6.0E-7
6.0E-7
1.5E-4
3.8E-7
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class Y for uranium-234 and uranium-238,
Class W for antimony-125) were assumed.
2.22.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Pittsburgh, Pennsylvania. The 0-80 km population distribution
was produced using the computer code SECPOP and 1980 Census
Bureau data. Nearby individuals were located 250 m from the
assumed release point. Food consumption rates appropriate to a
rural location were used.
2-106
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2.22.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-234 and
uranium-238 (69 percent) and antimony-125 (10 percent). The
predominant exposure pathways are inhalation for uranium-234 and
uranium-238, and ground surface for antimony-125.
The results of the dose and risk assessment are presented in
Tables 2.22-2 through 2.22-4. Table 2.22-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.22-5 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.22-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.22-2.
Organ
Estimated radiation dose rates from the Bettis
Atomic Power Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
4.3E-3
3.5E-2
Table 2.22-3. Estimated fatal cancer risks from the Bettis
Atomic Power Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
1E-8
1E-6
Table 2.22-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Bettis
Atomic Power Laboratory.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
TOTALS
0
0
0
0
0
0
3,100,000
3,100,000
0
0
0
0
0
0
1E-6
1E-6
2-107
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2.23 KNOLLS LAB - WINDSOR
2.23.1 Site Description
The Windsor site consists of only 4 ha near Windsor,
Connecticut, about 8 km north of the city of Hartford. The area
is a rural farming and industrial region along the Farmington
River. Approximately 3.1 million people live within 80 km.
2.23.2 Basis for the Dose and Risk Assessment
2.23.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.23-1.
Table 2.23-1. Radionuclides released to air during 1986 from
Knolls Atomic Power Lab-Windsor.
Nuclide Release Rate (Ci/y)
Ar-41
C-14
Co-60
H-3
Kr-83M
Kr-85
Kr-85M
Kr-87
Kr-88
X6-131M
Xe-133
Xe-133M
Xe-135
7.8E-2
4.7E-2
2.6E-7
1.1E-2
5.1E-5
2.3E-7
1.9E-4
1.4E-4
3.6E-4
l.OE-5
1.9E-3
6.6E-5
1.8E-3
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes were assumed.
2.23.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Hartford/Bradley, Connecticut. The 0-80 km population
distribution was produced using the computer code SECPOP and 1980
Census Bureau data. Nearby individuals were located 250 m from
the assumed release point. Food consumption rates appropriate to
a rural location were used.
2-108
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2.23.3 Results of the Dose and Risk Assessment
The major contributor to exposure is argon-41 (93 percent).
The predominant exposure pathway is air immersion.
The results of the dose and risk assessment are presented in
Tables 2.23-2 through 2.23-4. Table 2.23-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.23-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.23-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.23-2. Estimated radiation dose rates from the Knolls
Lab-Windsor.
Organ Nearby Individuals Regional Population
(mrem/y) (person-rem/y)
Gonads
Remainder
Breast
Red marrow
Lungs
3.8E-4
3.0E-4
3.5E-4
3.0E-4
2.9E-4
2.3E-3
4.2E-3
4.9E-3
8.1E-3
2.5E-3
Table 2.23-3. Estimated fatal cancer risks from the Knolls
Lab-Windsor.
Nearby Individuals Regional (0-80 km) Population
Lifetime Fatal Cancer Risk Deaths/y
8E-9 2E-6
2-109
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Table 2.23-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Knolls
Atomic Power Lab-Windsor.
Totals
Number of Persons
3,200,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
0
0
3,200,000
0
0
0
0
0
0
2E-6
2E-6
2.24 ROCKY FLATS PLANT
2.24.1 Site Description
Activities at the Rocky Flats Plant, located in Jefferson
County, Colorado, about 26 km from Denver, are restricted to
fabrication and assembly of components for nuclear weapons and
the support of these operations (Se88).
Fabrication operations include reduction rolling, blanking,
forming, and heat treating. Assembly operations include
cleaning, brazing, marking, welding, weighing, matching,
sampling, heating, and monitoring. Solid residue generated
during plutonium-related operations is recycled through one of
two plutonium-recovery processes. Process selection depends on
the purity and plutonium content of the residue. Both processes
produce a plutonium nitrate solution from which the metal can be
extracted. The recovered plutonium is returned to the storage
vault for use in foundry operations. A secondary objective of
the process is the recovery of americium-241.
Radionuclides are released from short stacks and building
vents at this plant. Building 771, Main Plenum, was selected for
comparison purposes and calculations. This point releases
54 percent of the plutonium-239 and -240 and 3 percent of the
uranium-233, -234, and -235 emitted at Rocky Flats. The most
significant release point for uranium is from a single duct in
Building 883, which releases approximately 19 percent of the
total uranium emissions from the plant.
2-110
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2.24.2 Basis for the Dose and Risk Assessment
2.24.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.24-1.
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class Y for uranium-238, Class W for
americium-241) were assumed.
Table 2.24-1. Radionuclides released to air during 1986 from
Rocky Flats Plant.
Nuclide Release Rate (Ci/y)
Am-241
H-3
Pu-233
Pu-234
Pu-238
Pu-239
Pu-240
U-233
U-234
U-238
4.8E-6
2.2E-1
1.7E-8
1.7E-8
9.8E-7
1.5E-5
1.5E-5
4.3E-6
4.3E-6
1.7E-5
2.24.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Denver/Stapleton, Colorado. The 0-80 km population distribution
was produced using the computer code SECPOP and 1980 Census
Bureau data. Nearby individuals were located 750 m from the
assumed release point (Se88). Food consumption rates appropriate
to a rural location were used.
2.24.3 Results of the Dose and Risk Assessment
The major contributors to exposure are uranium-238
(35 percent) and americium-241 (45 percent). The predominant
exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.24-2 through 2.24-4. Table 2.24-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.24-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.24-4 presents the estimated distribution of fatal cancer risk
to the regional population.
2-111
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Table 2.24-2,
Organ
Estimated radiation dose rates from the Rocky
Flats Plant.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
Endosteum
Remainder
6.3E-3
1.6E-2
7.5E-4
1.2E-1
2.0E-1
9.3E-3
Table 2.24-3. Estimated fatal cancer risks from the Rocky Flats
Plant.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
1E-8
9E-6
Table 2.24-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Rocky
Flats Plant.
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
1,900,000
1,900,000
0
0
0
0
0
0
9E-6
9E-6
2.25 PANTEX PLANT
2.25.1 Site Description
The Pantex Plant, located 30 km northeast of Amarillo,
Texas, is a nuclear weapons assembly and disassembly plant.
Because most radioactive materials handled during the assembly of
nuclear weapons are contained in sealed vessels, normal
operations involving these materials do not result in major
releases of radionuclides (La88).
2-112
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2.25.2 Basis for the Dose and Risk Assessment
2.25.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.25-1.
Table 2.25-1. Radionuclides released to air during 1986 from
Pantex Plant.
Nuclide Release Rate (Ci/y)
H-3 1.3E-1
U-238 l.OE-5
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class Y for uranium-238) were assumed.
2.25.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Amarillo, TX. The 0-80 km population distribution was produced
using the computer code SECPOP and the 1980 Census Bureau data.
Nearby individuals were located 1,500 m from the assumed release
point (La88). Food consumption rates appropriate to a rural
location were used.
2.25.3 Results of the Dose and Risk Assessment
The major contributor to exposure is uranium-238
(94 percent). The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.25-2 through 2.25-4. Table 2.25-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.25-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.25-4 presents the estimated distribution of fatal cancer risk
to the regional population.
2-113
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Table 2.25-2. Estimated radiation dose rates from the Pantex
Plant.
Organ
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
2.2E-3
3.5E-3
Table 2.25-3. Estimated fatal cancer risks from the Pantex
Plant.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
4E-9
7E-8
Table 2.25-4. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Pantex
Plant.
Risk Interval
Number of Persons
Deaths/y
1E-1 - 1E+0
1E-2 - 1E-1
1E-3 - 1E-2
1E-4 - 1E-3
1E-5 - 1E-4
1E-6 - 1E-5
< 1E-6
Totals
0
0
0
0
0
0
260,000
260,000
0
0
0
0
0
0
7E-8
7E-8
2.26 KNOLLS LAB - KNOLLS
2.26.1 Site Description
Knolls Atomic Power Laboratory has facilities at three
separate sites: Knolls, Kesselring, and Windsor. Development
ofnuclear reactors and training of operating personnel are the
major efforts at the Knolls Laboratory. The Knolls and
Kesselring complexes are located near Schenectady, NY, and the
Windsor site is near Windsor, Connecticut.
Operations at the Knolls site involving radioactive
materials are serviced by controlled exhaust systems that
2-114
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discharge through elevated stacks. Exhaust air is passed through
HEPA and carbon filters and is continuously sampled prior to
release. Small amounts of krypton-85 generated by examination of
irradiated fuel are released in the exhaust stacks. Generation
of argon-41 is minimized by controlling air leakage into the
low-power critical assembly.
2.26.2 Basis for the Dose and Risk Assessment
2.26.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.26-1.
Table 2.26-1. Radionuclides released to air during 1986 from
Knolls Atomic Power Lab-Knolls.
Nuclide Release Rate (Ci/y)
Co-60
1-131
Kr-85
Kr-85m
Kr-87
Kr-88
Pu-238
Sb-125
Sn-113
Sr-90
U-234
U-235
U-236
U-238
Xe-131m
Xe-133
Xe-135
l.OE-6
3.7E-6
7.9E-1
4.1E-3
5.8E-3
1.2E-2
1.3E-7
2.8E-5
1.3E-6
2.5E-5
3.3E-6
l.OE-7
6.6E-9
9.1E-10
5.7E-7
1.4E-3
1.3E-2
In modeling the site, all releases were assumed to be made
from a 10-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class Y for uranium-234) were assumed.
2.26.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Albany/CO, New York. The 0-80 km population distribution was
produced using the computer code SECPOP and 1980 Census Bureau
data. Nearby individuals were located 250 m from the assumed
release point. Food consumption rates appropriate to an urban
location were used.
2-115
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2.26.3 Results of the Dose and Risk Assessment
•
The major contributor to exposure is uranium-234 (79
percent). The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.26-2 through 2.26-4. Table 2.26-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.26-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.26-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.26-2.
Organ
Estimated radiation dose rates from the Knolls
Lab-Knolls.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Lungs
1.7E-3
3.1E-2
Table 2.26-3. Estimated fatal cancer risks from the Knolls
Lab-Knolls.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
3E-9
1E-6
Table 2.26-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Knolls
Atomic Power Lab-Knolls.
Totals
Number of Persons
1,200,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
- 1E+0
- 1E-1
- 1E-2
- 1E-3
- 1E-4
- 1E-5
1E-6
0
0
0
0
0
0
1,200,000
0
0
0
0
0
0
1E-6
1E-6
2-116
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2.27 AMES LABORATORY
2.27.1 Site Description
Until 1978, the Ames Laboratory, which is operated by Iowa
State University, was used 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, moderated and cooled by heavy water, and operated
continuously at 5,000 watts thermal. Operation of the Ames
Laboratory Research Reactor was terminated on December 1, 1977.
Decommissioning began January 3, 1978, and was completed on
October 31, 1981. A waste processing and disposal facility still
located at the site serves the campus reactor and research
laboratories.
Prior to its decommissioning, the major airborne releases
from the research reactor were tritium and argon-41. Tritium,
the major radionuclide released during the 1981 decommissioning
activities, was emitted from the 30-m reactor stack, which is 215
m from the nearest property boundary. Monitoring has indicated
that no airborne emissions from the research laboratories have
reached the main campus.
2.27.2 Basis for the Dose and Risk Assessment
2.27.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.27-1.
Table 2.27-1. Radionuclides released to air during 1986 from
Ames Laboratory.
Nuclide Release Rate (Ci/y)
H-3 7.6E-2
In modeling the site, all releases were assumed to be made
from a 10-m stack.
2.27.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from
Waterloo, Iowa. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data.
Nearby individuals were located 750 m from the assumed release
point. Rural food consumption rates were used.
2-117
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2.27.3 Results of the Dose and Risk Assessment
The major contributor to exposure is tritium (100 percent).
The predominant exposure pathways are ingestion and inhalation.
The results of the dose and risk assessment are presented in
Tables 2.27-2 through 2.27-4. Table 2.27-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.27-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.27-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.27-2,
Organ
Estimated radiation dose rates from the Ames
Laboratory.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Remainder
Gonads
Breast
Lungs
Red marrow
1.6E-5
1.3E-5
1.3E-5
1.3E-5
1.3E-5
2.3E-4
1.8E-4
1.8E-4
1.8E-4
1.8E-4
Table 2.27-3. Estimated fatal cancer risks from the Ames
Laboratory.
Nearby Individuals
Lifetime Fatal Cancer Risk
Regional (0-80 km) Population
Deaths/y
4E-10
9E-8
Table 2.27-4.
Risk Interval
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from the Ames
Laboratory.
Totals
Number of Persons
680,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
_
—
—
—
-
—
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
0
0
680,000
0
0
0
0
0
0
9E-8
9E-8
2-118
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2.28 ROCKETDYNE ROCKWELL
2.28.1 Site Description
Rockwell International operates two facilities, one near Los
Angeles and one near Santa Susana, Calafornia. These facilities
conduct research and development and also manufacture nuclear
reactor components. The Los Angeles facility performs uranium
fuel processing operations and conducts research involving gamma
radiation. The Santa Susana facility uses neutron radiography to
inspect nuclear reactor components. This facility also serves as
a materials handling laboratory and waste processing operation
for other DOE facilities.
Radionuclide emissions originate from the materials handling
laboratory and the waste processing facilities at the Santa
Susana site.
2.28.2 Basis for the Dose and Risk Assessment
2.28.2.1 Source Terms and Release Point Characterization
The total airborne releases, in Ci/y, from all sources
during 1986 are listed in Table 2.28-1.
Table 2.28-1. Radionuclides released to air during 1986 from
Rocketdyne Division, Rockwell International.
Nuclide Release Rate (Ci/y)
Sr-90 1.3E-5
In modeling the site, all releases were assumed to be made
from a 30-m stack. Default particle sizes (1.00 AMAD) and
solubility classes (Class D for strontium-90) were assumed.
2.28.2.2 Other Parameters Used in the Assessment
Meteorological data used in the assessment are from Burbank,
California. The 0-80 km population distribution was produced
using the computer code SECPOP and 1980 Census Bureau data.
Nearby individuals were located 250 m from the assumed release
point. Food consumption rates appropriate to an urban location
were used.
2-119
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2.28.3 Results of the Dose and Risk Assessment
The major contributor to exposure is strontium-90
(100 percent). The predominant exposure pathway is inhalation.
The results of the dose and risk assessment are presented in
Tables 2.28-2 through 2.28-4. Table 2.28-2 presents the doses
received by nearby individuals and the regional population.
Doses to organs accounting for 10 percent or more of the risk are
presented. Table 2.28-3 presents the estimated lifetime fatal
cancer risk to nearby individuals with maximum exposure, as well
as estimated deaths per year in the regional population. Table
2.28-4 presents the estimated distribution of fatal cancer risk
to the regional population.
Table 2.28-2. Estimated radiation dose rates from Rocketdyne
Division, Rockwell International.
Organ Nearby Individuals Regional Population
(mrem/y) (person-rem/y)
Red marrow 7.0E-6 1.4E-3
Endosteum 1.5E-5 3.2E-3
Table 2.28-3. Estimated fatal cancer risks from Rocketdyne
Division, Rockwell International.
Nearby Individuals Regional (0-80 km) Population
Lifetime Fatal Cancer Risk Deaths/y
2E-11 7E-8
2-120
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Table 2.28-4. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population from Rocketdyne
Division, Rockwell International.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
MB
-
-
-
-
-
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
0
0
8,800,000
0
0
0
0
0
0
7E-8
Totals 8,800,000 7E-8
2-121
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2.29 REFERENCES
Ba87
Ch88
De87
EIS86
Em87
EPA84
GU88
Ho87
Ho88
K188
La88
Baker, Samuel I., "Site Environmental Report for
Calendar Year 1986," Report 87/58, Fermi National
Accelerator Laboratory, Batavia, IL, May 1987.
Chew, Eddie W., and Mitchell, Russell, "1987
Environmental Monitoring Program Report for the Idaho
National Engineering Laboratory Site,"
DOE/ID-12082(87), Idaho Operations Office, DOE, Idaho
Falls, ID, May 1988.
Devlin, T.K., "1986 Environmental Monitoring Report,"
SAND87-8210.UC-11, Sandia National Laboratories,
Albuquerque, NM, April 1987.
U.S. Department of Energy, "Effluent Information
System, EPA Release Point Analysis Report for Calendar
Year 1986," Environmental Guidance Division.
Emerson, Marjorie Martz, et al., "Environmental
Surveillance at Los Alamos During 1986," LA-10992-ENV,
Los Alamos National Laboratory, Los Alamos, NM, April
1987.
U.S. Environmental Protection Agency, "Radionuclides:
Background Information Document for Final Rules,"
Volume II, EPA 520/1-84-022-2, Washington, DC, October
1984.
Gunderson, Thomas, et al., "Environmental Surveillance
at Los Alamos During 1987," LA-11306-ENV, Los Alamos
National Laboratory, Los Alamos, NM, May 1988.
Hoff, Diana L., Chew, Eddie W., and Rope, Susan K.,
"1986 Environmental Monitoring Program Report for the
Idaho National Engineering Laboratory Site,"
DOE/ID-12082(86), Idaho Operations Office, DOE, Idaho
Falls, ID, May 1987.
Holland, R.C., and Brekke, D.D., "Environmental
Monitoring at the Lawrence Livermore National
Laboratory Annual Report 1987," UCRL-50027-87, Lawrence
Livermore National Laboratory, Livermore, CA, April
1988.
Klein, Richard D., "1987 Pinellas Plant Environmental
Monitoring Report," GEPP-EM-1114, General Electric
Aerospace, April 1988.
Laseter, William A., and Langston, David C.,
"Environmental Monitoring Report for Pantex Plant
Covering 1987," MHSMP-88-19, Mason & Hanger-Silas Mason
Co., Inc., Amarillo, TX, April 1988.
2-122
-------
Mi87a Millard, G., et al., "1986 Environmental Monitoring
Report, Sandia National Laboratories," SAND87-0606,
Sandia National Laboratories, Albuquerque, NM, April
1987.
Mi87b Miltenberger, R.P., Royce, B.A., and Naidu, J.R., "1986
Environmental Monitoring Report, Brookhaven National
Laboratory," BNL 52088, Brookhaven National Laboratory,
Upton, NY, June 1987.
Mo84 Moore, E.B., "Control Technology for Radioactive
Emissions to the Atmosphere at U.S. Department of
Energy Facilities," PNL-4621 Final, Pacific Northwest
Laboratory, Richland, WA, October 1984.
Mo85 Moore, E.B., and Fullam, H.T., "Control Technology for
Radioactive Emissions to the Atmosphere at U.S.
Department of Energy Facilities: The Los Alamos Meson
Physics Facility," PNL-4621 Add. 1, Pacific Northwest
Laboratory, Richland, WA, March 1985.
Oa87a Oakes, T.W., et al., "Environmental Surveillance of the
U.S. Department of Energy Portsmouth Gaseous Diffusion
Plant and Surrounding Environs During 1986,"
ES/ESH-1/V4, Martin Marietta Energy Systems, April
1987.
Oa87b Oakes, T.W., et al., "Environmental Surveillance of the
U.S. Department of Energy Oak Ridge Reservation and
Surrounding Environs During 1986," ES/ESH-1/V1, Martin
Marietta Energy Systems, Oak Ridge, TN, April 1987.
Oa87c Oakes, T.W., et al., "Environmental Surveillance of the
U.S. Department of Energy Oak Ridge Reservation and
Surrounding Environs During 1986," ES/ESH-1/V4, Martin
Marietta Energy Systems, Oak Ridge, TN, April 1987.
PNL87 Pacific Northwest Laboratory, "Environmental Monitoring
at Hanford for 1986," Report PNL-6120, Richland, WA,
May 1987.
PNL88 Pacific Northwest Laboratory, "Environmental Monitoring
at Hanford for 1987," Report PNL-6464, Richland, WA,
May 1988.
RMI86 RMI, "Annual Environmental Monitoring Summary for RMI
Company Extrusion Plant, Ashtabula, Ohio, for 1986,"
prepared for U.S. DOE under contract No.
DE-AC05-760R01405, 1986.
2-123
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RMI89
R088
Sc87
Se88
SW87
TekSl
Th86
We87
Ze87
Letter and attachments dated August 2, 1989 from
Richard Mason, Director of Environmental Affairs, RMI
Company, to James Hardin, Environmental Standards
Branch, Criteria and Standards Division, Office of
Radiation Programs, U.S. Environmental Protection
Agency.
Rogers, J.G., et al., "Environmental Surveillance of
the U.S. Department of Energy Oak Ridge Reservation and
Surrounding Environs During 1987," ES/ESH-4/V1, Martin
Marietta Energy Systems, Inc., Oak Ridge, TN, April
1988.
Schleimer, Gary E., et al., "Annual Environmental
Monitoring Report of the Lawrence Berkeley Laboratory,
1986," LBL-23235, Lawrence Berkeley Laboratory,
Berkeley, CA, April 1987.
Setlock, George H., et al., "Annual Environmental
Monitoring Report, U.S. Department of Energy, Rocky
Flats Plant," RFP-ENV-87, Rockwell International,
Golden, CO, April 1987.
Swindall, E.R., et al., "Environmental Report for
Calendar Year 1986 on Radiological and Nonradiological
Parameters, Battelle," BCD 5186, Battelle Columbus
Division, Columbus, OH, May 1, 1987..
"Technical Support for the Evaluation and Control of
Emissions of Radioactive Materials to Ambient Air,"
Teknekron Research, Inc., May 7, 1981.
Thompson, James J., "Environmental Report for Lovelace
Inhalatzion Toxicology Research Institute for CY-1985,"
LITRI, Albuquerque, NM, April 1986.
Westinghouse Materials Company of Ohio, "Feed Materials
Production Center, Environmental Monitoring Annual
Report for 1986," FMPC-2076, Cincinnati, OH, April
1987.
Zeigler, Carroll C., et al., "Savannah River Plant
Environmental Report - Annual Report for 1986,"
DPSPU-87-30-1, Vols. I and II, E.I. du Pont de Nemours
& Co., Savannah River Plant, Aiken, SC, 1987.
2-124
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3. NRC-LICENSED AND NON-DOE FEDERAL FACILITIES
3.1 INTRODUCTION AND BACKGROUND
The Nuclear Regulatory Commission (NRC) and the Agreement
States issue licenses for the use of radionuclides. This chapter
deals with all of these licensed facilities that are not involved
in nuclear power generation and with Federal facilities other
than those owned by the Department of Energy (DOE). The
facilities that are part of the light-water uranium fuel cycle
are discussed in Chapter 4 of this report, and DOE facilities are
examined in Chapter 2. Facilities licensed only for the
possession of sealed sources are not considered, since sealed
sources do not release radionuclides to air.
NRC and Agreement State licensees are divided into by-
product, source material, and special nuclear material
categories. By-product licensees are further divided into
hospitals, radiopharmaceutical manufacturers, research
laboratories, sealed source manufacturers, and low-level waste
incinerators. Special nuclear material licensees are divided
into research reactors and non-light-water reactor fuel
fabricators.
Most non-DOE Federal facilities are included in the above
categories. For example, Veterans Administration hospitals are
included in the hospital category. Federal facilities not
included in any other category are discussed separately. Thus,
this source category is divided into nine sub-categories:
o Hospitals
o Radiopharmaceutical Manufacturers
o Research Laboratories
o Research Reactors
o Sealed Source Manufacturers
o Non-LWR Fuel Fabricators
o Source Material Licensees
o Low-Level Waste Incinerators
o Non-DOE Federal Facilities.
There are approximately 6,000 such facilities, and they are
found in all 50 states. The largest groups are the 3,680
licensed hospitals and the 1,500 research laboratories. The
smallest group is the four non-light-water reactor fuel
fabricators. These facilities emit radionuclides over a wide
spectrum, usually in small amounts. Typically, effluent controls
are activated charcoal filters to delay the release of iodine and
noble gases and high efficiency particulate air (HEPA) filters to
capture particulates. Controls and information pertaining to
each category are discussed separately.
The information presented in this chapter was obtained from
sources identified by a literature search and direct contact with
licensees and regulators. Whenever possible, current (1988) data
3-1
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were used in the assessment. To determine which facilities are
fir:17 to nave the highest levels of emissions, Radiation Safety
Officers at licensed facilities and staff at the NRC and
Agreement States were contacted. The facilities identified were
then contacted to obtain effluent release data and additional
site-specific information. Since it was not possible to survey
all 6,000 licensees, facilities with high or unusual emissions
may have been missed.
The raw data from a Conference of Radiation Control Program
Directors' (CRCPD) survey of waste production and effluents were
also used (CRC87). While this survey does not identify
specific facilities or their exact locations, it does provide
data on the number of facilities and emissions. Additional data
were obtained from the American Hospital Association (AHA86) and
from survey results presented in Cook (Co81) and Corbit (Co83).
Based on the emissions identified for each facility, the
radiation doses and risks to nearby individuals and to the
regional population were asssessed. The methodology discussed in
Volume I of this Background Information Document was used in all
of the assessments.
3.2 HOSPITALS
3.2.1 General Description
Over half of the hospitals in the United States handle
radiopharmaceuticals. Most use them for radionuclide imaging, in
which a compound labeled with a nuclide such as technetium-99m is
traced through the patient's body using an elaborate radiation
detection system. Hospitals also administer large, therapeutic
amounts of nuclides such as 1-131. Radiopharmaceuticals are
mostly in liquid form but can also be gaseous or solid.
Radiogases, such as Xe-133, are used for in-vivo lung
studies. The gas is inhaled by the patient, then exhaled into a
collection or ventilation system. The gas is either released
directly to air, charcoal filtered, or held for decay. Liquids
are stored and handled in fume hoods, which may have effluent
filters. They can be volatilized during administration to the
patient, which normally occurs in a room at negative pressure but
without effluent controls.
Data from the American Hospital Association indicate that
there are 3,680 hospitals in the United States that handle
diagnostic radiopharmaceuticals (AHA86). About a third of these
(1,371) also handle therapeutic amounts of these drugs. Two-
thirds of these hospitals are located in urban areas; the rest
are in rural locations. States with the largest number of such
hospitals are California (317), Texas (270), and New York (197).
3-2
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3.2.2 Basis for Risk Assessment
The doses and risks caused by release of radionuclides to
air were assessed by constructing one model facility to represent
typical hospitals and a second model facility to represent a very
large hospital with larger emissions.
3.2.2.1 Emissions
Effluent data from over 100 hospitals, obtained from the
CRCPD survey (CRC87), were used to construct the model
facilities. Nearly all hospitals reported releases of xenon-133;
the highest release was 31.4 Ci/y and the average was 1 Ci/y.
Eight hospitals reported releases of iodine-125; the highest
release, 0.039 Ci/y, is about four times the average value of
0.01 Ci/y. Six reported releases of iodine-131; these also
averaged 0.01 Ci/y. Other nuclides were reported by one or two
hospitals. The absence of reported radioiodine releases is
common, due to the lack of effluent monitoring at hospitals.
Facilities with no reported emissions were omitted from the
computation of the average release rates.
The average emissions for xenon-133, iodine-125, and
iodine-131 were used to construct the typical model facility.
These average values are consistent with the release rates
reported by Corbit (Co83) and SC&A (SCA84). The large model
hospital was created using the maximum release reported in the
CRCPD survey. The estimated emissions for the model hospitals
are shown in Table 3-1.
Table 3-1. Estimated emissions from model hospitals.
Facility Radionuclide Release Rate
(Ci/y)
Typical Hospital
Large Hospital
Xe-133
1-125
1-131
Xe-133
1-125
1-131
l.OE+0
l.OE-2
l.OE-2
3.1E+1
3.9E-2
2.4E-2
3.2.2.2 Site Characteristics
The model representing typical hospitals was assessed at two
different locations. To represent the doses and risks in urban
areas, an assessment was made using demographic and
meteorological data for Boston, MA. Data for Columbia, MO, were
used to estimate the doses and risks for rural areas. The two
3-3
-------
assessments used urban and rural food supply assumptions
respectively. In both assessments, the stack height was set at
1 meter and the nearest individuals were assumed to be 150 meters
downwind from the release point. The large model hospital was
again assessed using a 1 meter release height, an urban location,
and assuming the nearby individuals are 100 meters downwind.
Detailed information on the values input to the assessment
codes for these models is presented in Appendix A.
3.2.3 Results of the Dose and Risk Assessment
The results of the dose and risk assessment of the model
hospital facilities are presented in Tables 3-2 and 3-3. The
highest doses and risks are estimated for the large model
hospital. The highest doses to both nearby individuals and the
regional population are to the thyroid, 5.1 mrem/y and 12 person-
rem/year, respectively. These doses are caused by the iodine-125,
predominately via the ingestion pathway. The risks predicted for
the model large hospital indicated that nearby individuals have a
lifetime fatal cancer risk of approximately 2 in one million, and
that there will be 7E-5 deaths/year in the regional population.
The results for the model hospitals representing typical
urban and rural hospitals show lower doses and risks. For
the nearby individuals and the regional population at the model
urban hospital the highest doses are also to the thyroid,
0.2 mrem/y, and 1.4 person-rem/year, respectively. Iodine-125
and iodine-131 are the significant radionuclides, and inhalation
is the predominant pathway. The releases from the urban hospital
are estimated to result in lifetime fatal cancer risks to nearby
individuals much less than 1 in one million and to cause
approximately 1E-5 deaths/year in the regional population.
The highest doses received by nearby individuals
(28 mrem/y) and the regional populations (7.0 person-rem/year) at
the model rural hospital are also to the thyroid, due to
emissions of radioiodines. These doses are higher than those at
the urban hospital due to the greater significance of the
ingestion pathway.
The estimated distribution of the fatal cancer risk in the
exposed populations is presented in Table 3-4. An estimated 6E-2
deaths/year are caused by emissions from all hospitals. These
estimates were made by scaling the results obtained for the
typical urban and rural model hospitals by the number of urban
and rural hospitals, 2,467 and 1,213, respectively. The number
of persons at risk was constrained to the population of the
United States.
3-4
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Table 3-2. Estimated radiation dose rates from model hospitals.
Nearby Regional
Facility Organ Individuals Population
(mrem/y) (person-rem/y)
Urban Hospital
Rural Hospital
Large Hospital
Gonads
Breast
Thyroid
Remainder
Thyroid
Thyroid
1.2E-2
1.4E-2
2.0E-1
7.1E-3
2.8E+1
5.1E+0
4.6E-2
5.6E-2
1.4E+0
2.8E-2
7.0E+0
1.2E+1
Table 3-3. Estimated fatal cancer risks from model hospitals.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Urban Hospital 2E-7 1E-5
Rural Hospital 5E-6 2E-5
Large Hospital 2E-6 7E-5
Table 3-4. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all hospitals.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0 0 0
1E-2 to 1E-1 0 0
1E-3 to 1E-2 0 0
1E-4 to 1E-3 0 0
1E-5 to 1E-4 0 0
1E-6 to 1E-5 * *
< 1E-6 240,000,000 6E-2
Totals 240,000,000 6E-2
*Results from the large model hospital indicate there may be some
individuals at this risk level, but insufficient information is
available to quantify either the number of persons or the
deaths/year.
3-5
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3-2.4 Supplementary Control Options and Costs
Emissions from facilities in this segment of the NRC-
licensed source category do not result in exposures or risks high
enough to warrant a full evaluation of supplementary control
options and costs. Well-proven control technologies such as
charcoal for iodine or decay traps for noble gases could be
employed. Costs for any such system cannot be accurately
determined due to the number of facilities and the lack of
information on ventilation rates and on the extent of current
use of controls.
3.3 RADIOPHARMACEUTICAL MANUFACTURERS
3.3.1 General Description
Radiopharmaceutical suppliers, distributors, and nuclear
pharmacies number approximately 120 (Ce81). These are broken
down into 15 large firms, 70 small to medium-sized firms, and 35
nuclear pharmacy operators. The analysis focused on the large
firms that manufacture the radionuclides. These firms handle
large amounts of radionuclides in hot cells, which are equipped
with air cleaning systems (typically HEPA filters and charcoal).
The smaller firms change the chemical form of the nuclides, while
the pharmacies repackage the material into convenient amounts.
Information obtained on small firms and pharmacies suggests
that radionuclides are handled in fume hoods, which are equipped
with very efficient air cleaning filters. The most common
filters are charcoal beds, which trap radioiodines and noble
gases. Airborne effluents of these facilities are consequently
very much lower than those of the large manufacturers.
3.3.2 Basis for Risk Assessment
The assessment of radiopharmaceutical manufacturers is based
on the results obtained for four reference facilities. The
reference facilities are actual manufacturers that are among the
largest producers.
3.3.2.1 Emissions
Emissions data for three of the reference facilities were
obtained from the manufacturers themselves. The fourth facility
operates a nuclear reactor and is thus required to file effluent
reports with the NRC. The dose and risk assessments are based on
1987 effluent data. Emissions data were also available from the
CRCPD survey (1987) for seven unidentified facilities. These
data were used for comparative purposes only. Emissions for the
reference facilities are shown in Table 3-5.
3-6
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Table 3-5.
Effluent release rates (Ci/y) for radiopharmaceutical
manufacturers.
Radionuclide
Reference Facility
B C
D
P-32
S-35
1-125
1-131
H-3
C-14
Xe-135
Xe-135m
Xe-133
Xe-133m
Kr-88
Kr-87
Kr-85
Kr-85m
Kr-83m
Ar-41
1.6E-2
1.9E-2 1.6E-2
1.3E-2 2.0E-2
2.5E-3
- -
- -
-
- -
2.8E+0
- -
- -
- -
9.5E-1
-
-
— «-
^m
3.8E-1
-
-
9.8E+1
8.5E+0
-
-
-
-
-
-
-
-
-
~
^
—
2.5E+0
3.9E+0
—
—
8.1E+3
2.9E+3
1.4E+4
4 . 5E+2
1.7E+3
1.2E+2
1.7E+0
1.3E+3
4.6E+2
1.1E+3
3.3.2.2 Site Characteristics
Actual site data, where available, were used for the risk
assessments. Meteorological data were taken from the nearest
airports: Chicago, IL (A); Boston, MA (B&C); and Newburgh, NY
(D). Stack heights used were all 15 m. Distances to the nearby
individuals are 430 m, 200 m, 150 m, and 480 m. Food fractions
typical of urban areas were assumed in all cases except Reference
Facility D where rural food fractions were used.
3.3.3 Results of the Dose and Risk Assessment
The doses and risks estimated for the four reference
facilities are presented in Tables 3-6 and 3-7. The highest
estimated doses and risks are at Reference Facility D, where
nearby individuals and the regional population are predicted to
receive doses to the thyroid of 9.5E+1 mrem/y and 6.0E+2 person-
rem/year respectively. The lifetime fatal cancer risk to nearby
individuals is estimated to be 2E-4; the releases cause
2E-2 deaths/year in the regional population.
The total risk from radiopharmaceutical manufacturers is
estimated to be 2E-2 deaths/year. This is the sum of the
estimates for Reference Facilities A through C, multiplied by 5
and added to the estimate for Reference Facility D. The factor
of 5 is used to expand the three reference facilities to cover
all 15 actual facilities. Facility D is treated individually
because it is the only facility that operates a nuclear reactor.
Table 3-8 presents the collective risks, and number of people at
risk, as a function of individual risk level.
3-7
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Table 3-6. Estimated radiation dose rates from radiopharmaceu-
tical manufacturers.
Reference Organ
Facility
A Gonads
Thyroid
B Gonads
Breast
Thyroid
Remainder
C Gonads
Breast
Red Marrow
Lungs
Remainder
D Gonads
Breast
Thyroid
Remainder
Nearby
Individuals
(mrem/y)
8.9E-4
5.4E-2
4.4E-2
5.2E-2
7.3E-1
2.8E-2
7.1E-3
7.5E-3
7.9E-3
7.2E-3
7.9E-3
7.6E+0
7.5E+0
9.5E+1
5.7E+0
Regional
Population
(person-rem/y)
8.9E-3
2.4E+0
1.4E-2
1.7E-2
1.6E+0
1.3E-2
7.2E-1
9.9E-1
1.3E+0
7.7E-1
9.9E-1
7.4E+1
7.6E+1
6.0E+2
5.4E+1
Table 3-7.
Facility
Estimated fatal cancer risks from reference radio-
pharmaceutical manufacturers.
Nearby Individuals Regional (0-80 km)
Lifetime Fatal Population
Cancer Risk Deaths/y
A
B
C
D
2E-8
1E-6
2E-7
2E-4
7E-6
9E-6
4E-4
2E-2
3-8
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Table 3-8. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all radio-
pharmaceutical manufacturers.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
to
to
to
to
to
to
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
3,
140,
110,000,
0
0
0
0
100
000
000
0
0
0
0
2E-3
3E-3
2E-2
Totals 110,000,000 2E-2
3.3.4 Supplementary Control Options and Costs
Supplemental controls are examined for Reference Facility D,
which has the highest estimated doses and risks. The nuclides
contributing the most to dose are iodine-125 and iodine-131.
Control of these nuclides is typically by adsorption on activated
charcoal. However, Reference Facility D already employs this
control method.
Nevertheless, it is possible to increase the efficiency of
the existing charcoal adsorption system. Factors that influence
efficiency are the impregnant used, flow rate, humidity, and
temperature (Mo83). The first supplemental control examined is
drying the exhaust air before it enters the charcoal adsorbers.
Because the retention efficiency of charcoal is degraded by high
humidity conditions, drying the exhaust air will boost
efficiency.
The second option is chilling the charcoal beds. At lower
temperatures, iodine is retained on the charcoal for longer
periods. With a short half-life nuclide, such as iodine-131 (8
days), the activity decaying on the beds can be greatly
increased.
The cost of employing these enhancements is difficult to
determine, because they are dependent upon the configuration of
the existing system. If the original installation allowed for
the addition of these options at a later date, then their
installation would not be difficult. However, this is probably
not the case.
Lacking the data needed to perform an engineering study, the
cost of these modifications can only be estimated grossly. At
50 percent of the cost of a new system, this is estimated to be
$350,000 (DM80).
3-9
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The effectiveness of these modifications can only be
estimated. A reduction in radioiodine emissions of 99 percent
and noble gas emissions of 75 percent can be assumed. Such a
reduction would lower the calculated risks from this facility to
5E-3 deaths/year, reducing the predicted fatalities caused by
releases from all radiopharmaceutical manufacturers to
7E-3 deaths/year.
3.4 LABORATORIES
3.4.1 General Description
The NRC and Agreement States license approximately 1,500
laboratories that use radionuclides in unsealed forms. This
number is obtained by taking the total number of NRC-licensed
laboratories to be approximately 800 (NRC87) and adding it to a
previous count of 700 facilities licensed by the Agreement States
(Co83) . These laboratories are estimated to be 57 percent
academic and the remainder either government or private research
facilities. This estimate assumes that the number of academic
laboratories is a more stable figure and has remained relatively
unchanged from previous estimates (CeSl).
Academic laboratories generally encompass a large number of
sites in one area and use small amounts of a large number of
radionuclides. Twenty-nine radionuclides were identified in use
at various laboratories. Private and government laboratories use
millicurie to curie amounts of particular radioisotopes,
depending upon the actual procedures used. One of the more
important applications is the use of radioactively labeled
chemicals (i.e., radioiodine labeled proteins) to trace dynamic
processes.
The most pervasive form of effluent control is one or more
high efficiency particulate air (HEPA) filters in series
connected to a fume hood, hot cell, or glove box containing the
radioactive material. Often charcoal filters are used alone or
in series with HEPA filters to control the release of iodine and
noble gases. Exhaust alarms are typically set to sound if the
concentration at the release point reaches 10 percent of the
maximum permissible concentration (MPC) limit established by the
licensing authority. Quality assurance is maintained by periodic
wipe testing of the exhaust system either before the last filter,
if the filters are in a series, or at the point of release.
3.4.2 Basis for Risk Assessment
3.4.2.1 Emissions
Emissions data were gathered from 46 facilities. The
results from the CRCPD survey of effluents were also used. This
was a confidential survey, with the laboratories separated into
academic, private, and government facilities. The results from
3-10
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Corbit (Co83) were used, but only on a limited basis, because data
were separated by isotope and not by facility.
Approximately 41 percent of all laboratories have emissions
that are either zero or below the lower limits of detection of
their monitoring equipment. The majority of the laboratories
that do emit detectable quantities have exhaust concentrations
between 1 and 5 percent of the applicable MFC. The largest
emissions are estimated to be less than 10 percent of the MFC,
but for the purpose of this study were conservatively assumed to
be 10 percent of the MFC. Emissions are usually not monitored
continuously; instead, surveys are conducted monthly or bi-
monthly, and the emissions are estimated from these measurements.
A weighted average of all the information, omitting zero
responses, was used to estimate emissions for the model facility.
These are given in Table 3-9. The emission data were weighted by
segment composition (private/government = 43 percent,
academic = 57 percent) and sample size (primary = 45, CRCPD =
140, and Corbit = 44). The Corbit study (Co83) was given a
weight equivalent to one-half of its actual weight because it was
not separated into academic and private facilities. Finally, the
large number of nuclides was reduced by screening out those
nuclides making a negligible contribution to dose.
Table 3-9. Effluent release rates (Ci/y) for laboratories.
Radionuclide Model Facility Reference Facility A
H-3
C-14
S-35
Co-60
Kr-85
1-125
1-131
Xe-133
Cs-137
Pu-239
Am-241
1.1E+0
3.9E-3
4.7E-4 '
3.8E-5
1.8E-1
2.4E-3
5.1E-4
2.2E-1
-
3.7E-9
7.6E-10
—
—
2.1E-4
—
—
8.1E-3
—
1.5E-4
—
—
3.4.2.2 Site Characteristics
The model facility was placed in an urban area for purposes
of the risk assessment. Meteorological data were taken from an
actual airport. The release point was characterized as a 6 m
stack, 350 m from the closest resident. Facility A is an actual
laboratory with a 10 m release height. Meteorological data from
the nearest airport were used in the analysis. The closest
resident was in an urban area, 100 m from the stack.
3-11
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3.4.3 Results of the Dose and Risk Assessment
The results of the dose and risk assessment of the largest
and model facilities are presented in Tables 3-10 and 3-11. The
estimated organ doses are all below 1 mrem/y for nearby
individuals, and the maximum lifetime fatal cancer risk is
estimated to be 3E-7.
The estimated distribution of the fatal cancer risk in the
exposed population is presented in Table 3-12. The total
collective risk (deaths/year) from research laboratories is
obtained by scaling the model facility risk by 622, the estimated
number of laboratories that have non-zero emissions. The result
is an estimated 8E-3 deaths/year. The number of persons at risk
is constrained to the population of the United States.
Table 3-10. Estimated radiation dose rates from laboratories.
Nearby Regional
Facility Organ Individuals Population
(mrem/y) (person-rem/y)
Model Laboratory
Reference
Laboratory A
Gonads
Breast
Thyroid
Remainder
Gonads
Breast
Thyroid
Remainder
1.2E-2
1.3E-2
1.5E-1
8.8E-3
6.6E-3
6.0E-3
2.7E-2
5.0E-3
3.4E-2
3.5E-2
4.3E-1
3.1E-2
9.9E-2
9.0E-2
5.2E-1
7.6E-2
Table 3-11. Estimated fatal cancer risks from laboratories.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Model Laboratory 3E-7 1E-5
Reference Laboratory A 1E-7 3E-5
3-12
-------
Table 3-12 Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from all
laboratories.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
to
to
to
to
to
to
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
0
0
240,000,000
0
0
0
0
0
0
8E-3
Totals 240,000,000 8E-3
3.4.4 Supplementary Control Options and Costs
Emissions from facilities in this segment of the NRC-
licensed source category do not result in doses or risks high
enough to warrant a full evaluation of supplementary control
options and costs.
3.5 RESEARCH AND TEST REACTORS
3.5.1 General Description
There were 70 research and test reactors operating as of
December 1987 (NRC87). These reactors range in power level from
zero (three critical experiment facilities) to 10,000 kilowatts.
Most are located at universities and are used for teaching and
research. Of the many different designs and manufacturers,
the most common is General Atomies' TRIGA reactor.
There are two additional unlicensed reactors operated by the
U.S. Army in Maryland and New Mexico. They are discussed in
Section 3.10 of this chapter.
Most facilities ventilate the reactor building directly to
the atmosphere through tall stacks or roof vents. The larger
facilities employ particulate filters. Nearly all of the
facilities monitor their effluents.
3.5.2 Basis for Risk Assessment
Doses and risks resulting from test and research reactors
are evaluated on the basis of four actual reactors with the
largest emissions.
3-13
-------
3.5.2.1 Emissions
Emission data, shown in Table 3-13, were collected for the
four largest emitters identified by Corbit (Co83). These include
three university research reactors and one government research
reactor. Emissions data from Corbit were supplemented by
information presented in the facilities' annual operating reports
(e.g., MIT87). The principal nuclide emitted is argon-41. Tritium
is also emitted, although in lesser amounts.
Table 3-13. Effluent release rates (Ci/y) for research reactors.
Radionuclide
Facility H-3 Ar-41
Reference Reactor A 1.6E+1 2
Reference Reactor B 1.6E+2 I
Reference Reactor C - o
Reference Reactor D - ^
3.5.2.2 Site Characteristics
Actual site data were used for the four risk assessments.
Meteorological data were taken from airports near the four
facilities (Columbia, MO; Ft. Meade, MD; Boston, MA; Providence,
RI). The stack heights are 33 m, 33 m, 50 m, and 34 m,
respectively. Rural food supply assumptions were used for all
cases except Boston. The distances to the nearest individuals
are 750 m, 1,500 m, 750 m, and 1,500 m, respectively.
3.5.3 Results of the Dose and Risk Assessment
Doses and risks were calculated for each of the four
reference reactors. The results are presented in Tables 3-14 and
3-15 The highest exposures received by nearby individuals are
estimated to be 0.8 mrem/y to the gonads, and the individuals at
highest risk are estimated to have a lifetime fatal cancer risk
of 2E-5.
The fatal cancer risk estimated from these four reactors was
extrapolated to obtain the total collective risk (deaths/year)
from all research and test reactors. The extrapolation is based
on the ratio of the argon-41 released by the four largest emitters
(7 416 Ci/y) to the argon-41 released by all 70 research reactors
(12 557 Ci/y). This ratio, 0.59, was used to scale up the risk
from the four reactors to the total population risk of
4E-2 deaths/year from all research and test reactors. Table 3-16
presents the estimated collective risk, and the number of people
at risk, as a function of individual risk level.
3-14
-------
Table 3-14
Facility
Estimated radiation dose rates from research
reactors.
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Reference Reactor A
Reference Reactor B
Reference Reactor C
Reference Reactor D
Gonads
Breast
Red Marrow
Lungs
Remainder
Gonads
Breast •
Red Marrow
Lungs
Remainder
Gonads
Breast
Red Marrow
Lungs
Remainder
Gonads
Breast
Red Marrow
Lungs
Remainder
7.8E-1
7.0E-1
6.0E-1
6.0E-1
6.0E-1
2.6E-1
2.3E-1
2.0E-1
2.0E-1
2.0E-1
2.7E-1
2.4E-1
2.1E-1
2.1E-1
2.1E-1
3.6E-2
3.3E-2
2.8E-2
2.8E-2
2.8E-2
8 . 1E+0
7.3E+0
6.3E+0
6.2E+0
6.3E+0
7.3E+0
6.9E+0
6.2E+0
6.2E+0
6.8E+0
6.8E+1
6.1E+1
5.2E+1
5.2E+1
5.2E+1
4.4E-1
3.9E-1
3.4E-1
3.3E-1
3.4E-1
Table 3-15. Estimated fatal cancer risks from research reactors.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Reference Reactor A
Reference Reactor B
Reference Reactor C
Reference Reactor D
2E-5
5E-6
6E-6
7E-7
2E-3
2E-3
2E-2
1E-4
3-15
-------
Table 3-16. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from research and
and test reactors.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to
to
to
to
to
to
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
0
0
0
0
1,300
630,000
23,000,000
0
0
0
0
2E-4
2E-2
2E-2
Totals 24,000,000 4E-2
3.5.4 Supplementary Control Options and Costs
Emissions from facilities in this segment of the NRC-
licensed source category do not result in exposures or risks high
enough to warrant a full evaluation of supplementary control
options and costs.
3.6 SEALED SOURCE MANUFACTURERS
3.6.1 General Description
Sealed source manufacturers take radionuclides in an
unsealed form and put them into a permanently sealed container.
Two categories of sealed source manufacturers contribute to
airborne emissions. The first category consists of manufacturers
that produce sealed radiation sources other than tritium (such as
Am-241). There are eight known manufacturers of this type. An
additional six manufacturers of this type (e.g., The Nucleus, Oak
Ridge, TN) use only exempt quantities of radionuclides and
produce negligible emissions.
The other category of sealed source manufacturer seals
tritium gas into self-luminous lights. There are three known
firms that perform this type of work. All of these facilities
are located in industrial areas. They rely heavily on engineered
safeguards to prevent releases of radionuclides.
The radiation source manufacturers use high efficiency
particulate air.(HEPA) filters singly or in series to remove
radionuclides from their effluent streams. The lighting
manufacturers use desiccant columns, sometimes combined with
catalytic recombiners, to remove tritium from their effluents.
The only part of the process that results in emissions is the
loading of radionuclides into containers which are subsequently
sealed. All of the work is done in controlled areas, with
3-16
-------
radiation monitors in operation to detect any leaks. The sealed
containers are stored and shipped without emissions.
3.6.2 Basis for Risk Assessment
The doses and risks resulting from the operations of sealed
source manufacturers are assessed using the actual emissions and
site characteristics for the three manufacturers of self-luminous
lights (Reference Facilities A, B, and C) and a model facility to
represent the non-tritium source manufacturing facilities.
3.6.2.1 Emissions
The source term for the model radiation source facility is
based on the arithmetic average of the emissions from four
facilities that provided data. The model facility emits krypton-
85, cobalt-60, americium-241, iridium-192, and californium-252,
as shown in Table 3-17. The tritium lighting producers all
provided effluent data for 1984, so no model facility is needed.
Their emissions are also shown in Table 3-17. Since 1984,
Reference Facility C has installed a catalytic recombiner system;
therefore, current emissions are lower than the 1984 values.
Table 3-17. Effluent release rates (Ci/y) for sealed source
manufacturers.
Radionuclide
H-3
Co-60
Ni-63
Kr-85
Ir-192
Po-210
Am-241
Cf-252
Model
Facility
—
3.2E-7
-
2.4E-1
3.3E-6
-
1.4E-7
3.0E-9
Reference
Facility A
3.4E+2
-
8.0E-6
-
-
1.4E-4
6.1E-5
^
Reference
Facility B
1.5E+3
-
-
-
-
-
-
*™
Reference
Facility C
2.2E+3
-
-
-
-
-
-
^
3.6.2.2 Site Characteristics
The model facility was placed in an urban area. It was
assumed to have a 6 m stack, 250 m away from the nearest
resident. The tritium lighting manufacturers were assessed using
actual site data. Meteorology was taken from nearby airports
(Buffalo, NY; White Plains, NY; and Harrisburg, PA). Stack
heights were set at 10 m. Nearby individuals are located 7,500 m,
400 m, and 150m, respectively, from the facilities. The New York
sites were treated as urban sites; the Pennsylvania site, as
rural.
3-17
-------
3.6.3 Results of the Dose and Risk Assessment
Tables 3-18 and 3-19 show the results of the assessment for
the model radiation source facility and all of the tritium
lighting facilities. The highest estimated doses from
non-tritium sealed source manufacturers are estimated to be to
the endosteum and red marrow, both less than 1 mrem/y. The
lifetime risk to nearby individuals is 8E-10. For the tritium
lighting manufacturers, nearby individuals are estimated to
receive doses on the order of 6 mrem/y and to have a lifetime
fatal cancer risk of 2E-4.
To estimate the collective risk (deaths/year) from all
sealed source manufacturers, the risk from the model was
multiplied by 8 and added to the sum of the risks from the three
tritium lighting facilities. This yields the total risk from
this category of 2E-2 deaths/year. Table 3-20 presents this
collective risk, and the number of people at risk, as a function
of individual risk level.
Table 3-18,
Estimated radiation dose rates from sealed source
manufacturers.
Facility
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rera/y)
Model Facility
Reference Facility A
Reference Facility B
Reference Facility C
Red Marrow
Endosteum
Remainder
Gonads
Breast
Red Marrow
Lungs
Endosteum
Remainder
Gonads
Breast
Red Marrow
Lungs
Remainder
Gonads
Breast
Red Marrow
Lungs
Remainder
1.8E-4
2.2E-3
l.OE-4
1.4E-1
1.3E-1
1.7E-1
1.4E-1
5.8E-1
1.9E-1
5.6E-1
5.6E-1
5.5E-1
5.6E-1
6.0E-1
5.4E+0
5.4E+0
5.4E+0
5.5E+0
6.7E+0
1.3E-3
1.5E-2
7.1E-4
l.OE+0
l.OE+0
1.2E+0
1.1E+0
3.4E+0
1.4E+0
2.8E+1
2.8E+1
2.8E+1
2.8E+1
3.3E+1
9.2E+0
9.2E+0
9.1E+0
9.2E+0
1.1E+1
3-18
-------
Table 3-19. Estimated fatal cancer risks from sealed source
manufacturers.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Model Facility 8E-10 8E-8
Reference Facility A 4E-6 4E-4
Reference Facility B 2E-5 1E-2
Reference Facility C 2E-4 4E-3
Table 3-20. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from sealed
source manufacturers.
Risk Interval Number of Persons Deaths/y
1E
-------
because this technology is not widely applied. There are only a
handful of such installations, and each one is custom engineered.
Applying this supplemental control to Reference Facility C
would cost approximately $1.7 to $7.0 million. The effectiveness
of this system can only be estimated. Assuming a 99 percent
reduction in emissions from Reference Facility C, the risk from
this category would be reduced by half, to 1E-2 deaths/year.
3.7 NON-LWR FUEL FABRICATORS
3.7.1 General Description
Facilities in this category fabricate uranium fuel for
research reactors and naval propulsion reactors. Three
facilities making naval fuel were identified. One other facility
manufactures only research reactor fuel. The process is similar
to fabrication of power reactor fuel, where enriched U02 is
formed into pellets, which are stacked inside tubes, and then
bundled into fuel assemblies or cores. Fabrication procedures
for naval fuel are classified.
Effluents to air are controlled using HEPA filters and/or
gas scrubbers. The scrubbers are used to neutralize and remove
the nitrogen oxides formed during HNO3 pickling (chemical
milling) operations at some facilities.
3.7.2 Basis for Risk Assessment
The doses and risks associated with this segment of the NRC-
licensed source category are evaluated using actual emissions
data and site characteristics for three of the four facilities.
3.7.2.1 Emissions
Recent (1987) data were obtained from operating reports for
three facilities. The nuclides released that contribute the most
to dose are uranium-234 and uranium-235. Release quantities of
these and other isotopes are shown in Table 3-21.
3.7.2.2 Site Characteristics
Actual site and facility data were used in the risk
assessment. Meteorology data were taken from nearby airports
(Providence, RI; Knoxville, TN; and North San Diego, CA). Urban
food supply assumptions were used, except in the analysis of the
first facility which is in a rural area. The first facility
releases effluents from a roof vent and was treated as an area
source (525 m2). The second and third facilities release through
35 m and 6 m stacks, respectively. The distances to the nearest
residents are respectively 425 m, 350 m, and 750 m.
3-20
-------
Table 3-21. Effluent release rates (Ci/y) for non-LWR fuel
fabricators.
Facility
Radionuclide Naval A Naval B Research
U-234
U-235
U-236
U-238
Am-241
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Th-232
Ar-41
Co-60
Sr-90
Y-90
Cs-137
1-131
4.3E-5
1.2E-6
7.8E-8
2.1E-9
-
-
-
-
-
-
-
-
-
-
-
-
~
3.4E-3
8.1E-5
1.2E-6
5.7E-5
2.6E-8
4.2E-8
2.2E-8
2.0E-8
2.8E-6
2.9E-11
-
-
-
-
-
-
«•
3.3E-6
1.5E-5
-
3.6E-6
-
-
-
-
-
-
4.0E-8
1.2E+0
4.0E-5
4.8E-7
4.8E-7
1.4E-4
l.OE-6
3.7.3 Results of the Dose and Risk Assessment
Off-site dose and risk were calculated for the three
facilities from which release data were obtained. The results
are shown in Tables 3-22 and 3-23. None of these facilities are
estimated to cause nearby individuals doses greater than
1 mrem/y, and the lifetime fatal cancer risks to nearby individuals
are less than 1E-6.
The estimated distribution of the fatal cancer risk to the
regional populations from all non-LWR fuel fabricators is
presented in Table 3-24. The deaths/year from the naval fuel
fabricators were added and scaled up by 50 percent to account for
the other facility of this type. The risks from the single
research reactor fuel fabricator were then added. The result is
the total risk of 2E-4 deaths/year.
3.7.4 Supplemental Control Options and Costs
Emissions from facilities in this segment of the NRC-
licensed source category do not result in exposures or risks high
enough to warrant a full evaluation of supplementary control
options and costs. The well-proven technology of additional HEPA
filtration systems could be employed to reduce emissions further.
3-21
-------
Table 3-22. Estimated radiation dose rates from non-LWR fuel
fabricators.
Facility
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Naval Fuel A
Naval Fuel B
Research Fuel
Lungs
Lungs
Gonads
Lungs
Remainder
1.5E-1
4.2E-1
1.1E-2
1.1E-1
8.4E-3
2.5E-1
4.7E+0
5.1E-2
5.8E-1
4.1E-2
Table 3-23,
Facility
Estimated fatal cancer risks from non-LWR fuel
fabricators.
Nearby Individuals
Lifetime Fatal
Cancer Risk
Regional (0-80 km)
Population
Deaths/y
Naval
Naval
Fuel A
Fuel B
Research Fuel
2E-7
7E-7
4E-7
6E-6
1E-4
3E-5
Table 3-24. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all non-LWR
fuel fabricators.
Risk Interval
Number of Persons
Deaths/y
1E-1 to
1E-2 to
1E-3 to
1E-4 to
1E-5 to
1E-6 to
< IE-
Totals
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
6
0
0
0
0
0
0
8,200,000
8,200,000
0
0
0
0
0
0
2E-4
2E-4
3-22
-------
3.8 SOURCE MATERIAL LICENSEES
3.8.1 General Description
Source material licensees are companies that handle
relatively large amounts of thorium or uranium (non-enriched)
during the manufacture of a product. The NRC licenses 12
facilities for the use of thorium (Mo88). Nine of them are
currently using thorium. It is assumed that a similar number of
facilities are active in Agreement States. This assumption is
probably conservative because after contacting half of the
Agreement States, only one active license for the use of thorium
was located. Only four facilities in the United States hold
source material licenses for the processing of depleted uranium.
The processes used by these licensees are varied. The
facilities that emit thorium process low-thorium-content alloys
into wire for lighting purposes. Other uses of thorium include
scrap collection, glass creation, and lens coating. The depleted
uranium is universally extruded into projectiles. In all of
these processes, HEPA filters are used in series to reduce
effluent levels. During extrusion and machining, lubricants are
sprayed on the material to prevent particles from becoming
airborne. The lubricants are then collected and disposed of as
solid waste.
3.8.2 Basis for the Risk Assessment
A reference thorium facility and a reference uranium
facility were used to evaluate the doses and risks of source
material manufacturers.
3.8.2.1 Emissions
The emissions from source material licensees are split
between facilities that have no emissions and facilities that
emit approximately 3E-4 Ci/y of thorium or uranium. The thorium
facilities are modeled by an existing facility that emits at this
level. The uranium plants emit depleted uranium in the hundreds
of microcuries. These plants are likewise modeled by a reference
facility. Release rates are shown in Table 3-25.
Table 3-25. Effluent release rates for source material
licensees.
Radionuclide (Ci/y)
Facility U-234 U-235 U-238 Th-232
Uranium
Thorium
2.7E-4 7.0E-6 2.7E-4
3.0E-4
3-23
-------
3.8.2.2 Site Characteristics
The two reference facilities were assessed using actual site
and facility data. Meteorology data came from nearby airports
(Cleveland, OH, and Bristol, TN). Effluent release heights are
10 m and 6 m, respectively. The nearest residents are located
100 m and 200 m away from the respective facility. Both
facilities were assessed using urban food assumptions.
3-8.3 Results of the Dose and Risk Assessment
Tables 3-26 and 3-27 present the results of the dose and
risk estimates for nearby individuals and the regional population
for the reference facilities. Nearby individuals are estimated
to receive doses to the lungs or the endosteum on the order of
3 mrem/y and to have a lifetime fatal cancer risk of about
4E-6.
Table 3-28 presents the estimated distribution of the fatal
cancer risk to the regional populations from all source material
licensees. This estimate was obtained by scaling the results for
the reference facilities by the number of actual facilities. The
total collective risk is estimated to be 1E-3 deaths/year.
Table 3-26. Estimated radiation dose rates from source material
licensees.
Nearby Regional
Facility Organ Individuals Population
(mrem/y) (person-rem/y)
Uranium
Thorium
Lungs
Lungs
Endosteum
2.7E+0
2.6E+0
4.1E+0
3.4E+0
l.OE+1
1.6E+1
Table 3-27. Estimated fatal cancer risks from source material
licensees.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Uranium 4E-6 8E-5
Thorium 3E-6 1E-4
3-24
-------
Table 3-28. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all source
material licensees.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
to
to
to
to
to
to
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
0
0
0
0
0
0
24,000,000
0
0
0
0
0
0
1E-3
Totals 24,000,000 1E-3
3.8.4 Supplemental Control Options and Costs
Emissions from facilities in this segment of the NRC-
licensed source category do not result in exposures or risks high
enough to warrant an evaluation of supplementary control options
and costs. The well-proven technology of additional HEPA
filtration systems could be employed to reduce emissions further.
3.9 LOW-LEVEL WASTE INCINERATORS
3.9.1 General Description
Airborne effluents from low-level waste handling and
disposal arise primarily from waste incineration. The practice
of evaporating disposal site liquids has ceased, so this is no
longer a source of releases to air. Incineration is done mainly
by large research laboratories and hospitals. About 100 such
incinerators are operating in the United States.
The older incinerators usually release directly to the
atmosphere. The newer ones are designed with sophisticated
effluent control systems, including afterburners, venturi
scrubbers, and gas scrubbers (e.g., NaOH and water). Since the
newer units have much higher capacities (e.g., 1,000 Ib/hr), they
are replacing the older units.
3.9.2 Basis for the Risk Assessment
The dose and risk assessment is based on a large reference
facility to obtain doses and risks to nearby individuals and a
model facility with average emissions to obtain collective doses
and risks.
3-25
-------
3.9.2.1 Emissions
Effluent data were obtained from the CRCPD survey (1987) for
35 incinerators. Nearly all reported releases of tritium and
carbon-14. Nine or fewer facilities reported releases of sulfur-35
chromium-51, iodine-125, and phosphorus-32. A model facility was '
created using the average releases of these nuclides. An actual
facility reporting the largest releases of the above nuclides was
also modeled. Table 3-29 presents the source terms used in the
assessment.
3.9.2.2 Site Characteristics
The model and large incinerator were both placed at a
suburban site for the risk assessment. They both have a stack
height of 35 m and a thermal release rate of 2.2E+5 cal/second.
The nearest resident is located 300 m away. Both assessments
used meteorological data from a nearby airport.
Table 3-29. Effluent release rates (Ci/y) for low-level waste
disposal facilities.
Radionuclide
H-3
C-14
P-32
S-35
Cr-51
Se-75
1-125
Model
Facility
l.OE-1
5.0E-2
7.0E-2
l.OE-1
l.OE-2
-
1.5E-2
Reference
Facility
"1.3E+0
1.5E+0
1.4E-1
8.7E-1
5.0E-2
l.OE-3
9.0E-2
3.9.3 Results of the Dose and Risk Assessment
Assessments for the model incinerator and the large
reference facility indicate that nearby individuals receive doses
less than 1 mrem/y and have lifetime fatal cancer risks of less
than 1E-6. The results are shown in Tables 3-30 and 3-31.
Table 3-32 presents the estimated distribution of the fatal
cancer risk to the regional populations from all low-level waste
disposal facilities. This estimate was obtained by scaling up
the risks from the model facility by a factor of 100. This gives
a risk of 1E-3 deaths/year from all incinerators.
3.9.4 Supplemental Control Options and Costs
Emissions from low-level waste disposal facilities do not
result in exposures or risks high enough to warrant an evaluation.
of supplementary control options and costs.
3-26
-------
Table 3-30,
Facility
Estimated radiation dose rates from low-level waste
disposal facilities.
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Model Facility
Reference Facility
Red Marrow
Lungs
Endosteum
Remainder
Gonads
Breast
Thyroid
Remainder
8.0E-4
4.6E-4
9.4E-4
2.5E-4
1.2E-2
1.4E-2
1.1E-1
8.1E-3
6.7E-2
2.3E-2
7.7E-2
2.5E-2
6.0E-1
7.5E-1
1.1E+1
5.5E-1
Table 3-31,
Facility
Estimated fatal cancer risks from low-level waste
disposal facilities.
Nearby Individuals
Lifetime Fatal
Cancer Risk
Regional (0-80 km)
Population
Deaths/y
Model Facility
Reference Facility
1E-8
3E-7
1E-5
2E-4
Table 3-32.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all low-level
waste disposal facilities.
Risk Interval
Totals
Number of Persons
240,000,000
Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
0
0
0
0
0
0
240,000,000
0
0
0
0
0
0
1E-3
1E-3
3-27
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3.10 NON-DOE FEDERAL FACILITIES
3.10.1 General Description
This category includes Department of Defense (DOD)
facilities. Other non-DOE federal facilities, such as Veterans
Administration hospitals and NASA research laboratories, are
included in the evaluations presented in Sections 3.2, 3.4, and
3.5. Federal facilities operated by the DOE are discussed in
Chapter 2.
This category is made up of two groups of DOD facilities.
The first and largest group consists of nuclear shipyards and
naval bases. The second consists of DOD research reactors.
There are 13 active shipyards and bases. Seven are on the east
coast, five are on the west coast, and one is in Hawaii. These
facilities refuel and service the Navy's nuclear fleet. Most of
the radioactive wastes are in solid form. According to the Navy,
there are no significant discharges of airborne radioactivity
(Ma88). Exhaust air from waste handling buildings is passed
through HEPA filters to control emissions.
The DOD operates two unlicensed research reactors, at
Aberdeen, MD, and White Sands, NM. Operations and effluent
control are essentially the same as for the research reactors
described in Section 3.5.
3.10.2 Basis for the Risk Assessment
A single model facility is used to estimate the doses and
risks from this segment of the NRC-licensed source category, as
the magnitudes of the releases from both the DOD reactors and the
shipyards are comparable.
3.10.2.1 Emissions
Effluent monitoring at DOD shipyards and bases reveals few
measurable nuclides (Ma88). However, the Navy has estimated
maximum releases, based on many years of monitoring data. These
releases are primarily noble gases and cobalt-60 (see
Table 3-33). Since the magnitude of the releases from DOD
research reactors (Co83) are comparable to the maximum releases
estimated by the Navy, the emissions for the single model
facility represent both types of actual DOD sites.
3.10.2.2 Site Characteristics
For purposes of the risk assessment, the model DOD facility
was placed at the site of an actual west coast shipyard.
Meteorological data came from that same shipyard. The release
height was assumed to be 15 m, and the distance to the nearest
residents is 1,500 m. Rural food supply assumptions were used.
3-28
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Table 3-33. Effluent release rates (Ci/y) for DOD facilities.
Radionuclide Model Facility
H-3 l.OE-3
C-14 l.OE-1
Co-60 l.OE-3
Kr-83m 2.0E-2
Kr-85m 2.4E-2
Kr-87 5.0E-2
Kr-88 2.0E-2
Xe-131m 5.0E-3
Xe-133m l.OE-2
Xe-133 2.1E-1
Xe-135 2.5E-1
Ar-41 4.1E-1
3.10.3 Results of the Dose and Risk Assessment
The doses and risks from the model facility are shown in
Tables 3-34 and 3-35. Table 3-36 presents the estimated
distribution of the fatal cancer risk to the regional populations
from all DOD facilities. This estimate was made by multiplying
the risks estimated for the model facility by a factor of 12.
This factor is obtained by considering the shipyards and bases
that are in proximity (e.g., Newport News and Norfolk, VA) as
single facilities. The collective population risk from all DOD
facilities is estimated to be 1E-3 deaths/year.
3.10.4 Supplementary Control Options and Costs
Emissions from facilities in this segment of the NRC-
licensed and non-DOE Federal source category do not result in
exposures or risks high enough to warrant an evaluation of
supplementary control options and costs.
Table 3-34. Estimated radiation dose rates from DOD facilities.
Nearby Regional
Facility Organ Individuals Population
(mrem/y) (person-rem/y)
Model Facility
Gonads
Breast
Red Marrow
Lungs
Remainder
1.1E-2
l.OE-2
8.9E-3
1.1E-2
9.0E-3
2.5E-1
2.4E-1
2.3E-1
2.3E-1
2.1E-1
3-29
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Table 3-35. Estimated fatal cancer risks from DOD facilities.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Model Facility 2E-7 8E-5
Table 3-36. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all DOD
facilities.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
0
0
0
0
0
0
64,000,000
0
0
0
0
0
0
1E-3
Totals 64,000,000 1E-3
3.11 SUMMARY OF THE COLLECTIVE RISKS FROM ALL FACILITIES
The population risks calculated for each of the nine sub-
categories were combined to obtain an estimate of the total
deaths/year resulting from emissions from all NRC-licensed
facilities. The results are presented in Table 3-37. Because
the regional population extends 80 km from each facility,
individuals are exposed to emissions from more than a single
facility. Thus, the combined regional population obtained by
summing the results of the individual estimates exceeds the total
population of the United States. The number of persons at risk
shown in Table 3-37 is therefore limited to 240 million persons,
the population of the United States. The total risk from this
category, 2E-1 deaths/year, was not adjusted to account for this
overlap, since virtually all the risk is incurred by individuals
living close to each facility.
The largest contributors to the collective risk are research
reactors and hospitals, estimated to cause 4E-2 and 6E-2
deaths/year, respectively. Although hospitals have relatively
low emissions, there are many of them. The next highest
contributors to collective risk are radiopharmaceutical
manufacturers, estimated to cause 2E-2 deaths/year, and research
laboratories (8E-3 deaths/year). Like hospitals, research
3-30
-------
laboratories have low emissions, but their large number results
in small risks to many persons.
Table 3-37. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from all NRC-
licensed facilities.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
5,
780,
239,000,
0
0
0
*
000
000
000
0
0
0
*
2E-3
3E-2
1E-1
Totals 240,000,000 2E-1
*Results indicate there may be some individuals at this risk
level, but insufficient information is available to quantify
either the number of persons or the deaths/year.
With respect to individual risk, the maximum value of 2E-4
lifetime fatal cancer risk is estimated for both a
radiopharmaceutical manufacturer and a sealed source
manufacturer. A research reactor and another sealed source
manufacturer account for the next highest individual risk,
estimated to be 2E-5.
These estimates of deaths per year in the regional
populations and maximum lifetime risks to nearby individuals must
be viewed with caution. Only a limited number of the 6,000
facilities in this category could be evaluated, and the
evaluations rest on unverified emissions data provided by the
facilities. While the methodology attempted to evaluate the
facilities with the greatest potential risk, the lack of
emissions data for so many of the facilities makes it impossible
to state with certainty that this goal was achieved. Thus, there
may be NRC-licensed and non-DOE Federal facilities causing
greater doses and risks than those that have been estimated in
this evaluation.
3-31
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3.12 REFERENCES
AHA86 American Hospital Association, "Annual Survey of
Hospitals," Chicago, IL, 1986.
Ce81 Centaur Associates, Inc., "An Economic Study of the
Radionuclides Industry," prepared for the U.S. Nuclear
Regulatory Commission, NUREG/CR-2048, Washington, D.C.,
1981.
Co81 Cook, J.R., "A Survey of Radioactive Effluent Releases
from Byproduct Material Facilities," U.S. Nuclear
Regulatory Commission, NUREG-0819, Washington, D.C., 1981.
Co83 Corbit, C.D., Herrington, W.N., Higby, D.P., Stout, L.A.,
Corley, J.P., "Background Information on Sources of Low-
Level Radionuclide Emissions to Air," Pacific Northwest
Laboratory, PNL-4670, Richland, WA, 1983.
CRC87 Council of Radiation Control Program Directors, Inc.,
"Compilation of State-by-State Low-Level Radioactive Waste
Information," U.S. Department of Energy, DOE/ID/12377,
Frankfort, KY, 1987.
DM80 Dames and Moore, "Airborne Radioactive Emission Control
Technology," prepared for the U.S. Environmental
Protection Agency, Office of Radiation Programs,
Washington, D.C., 1980
Ma88 Mangeno, J.J., Steele, J.M., Poletti, L.F., "Environmental
Monitoring and Disposal of Radioactive Wastes from U.S.
Naval Nuclear Powered Ships and Their Support Facilities
1987," Naval Nuclear Propulsion Program, NT-88-1,
Washington, D.C., 1988.
MIT87 M.I.T. Research Reactor Staff, "Annual Report to United
States Nuclear Regulatory Commission for the Period July
1, 1986 - June 30, 1987," Cambridge, MA, August 29, 1987.
Mo83 Moore, E., et al., "Control Technology for Radioactive
Emissions to the Atmosphere at U.S. Department of Energy
Facilities," Pacific Northwest Laboratory, PNL-4621,
Richland, WA, 1983.
Mo88 Moriarty, M., Personal Communication., U.S. Nuclear
Regulatory Commission, Washington, D.C., 1988.
NRC87 U.S. Nuclear Regulatory Commission, "Licensed Operating
Reactors: Status Summary Report," NUREG/0020, Washington,
D.C., 1987.
3-32
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SCA84 SC&A, Inc., "Impact of Proposed Clean Air Act Standards
for Radionuclides on Users of Radiopharmaceuticals,"
prepared for U.S. EPA, Office of Radiation Programs, under
Work Assignment #5, Contract #68-02-3853, with Jack
Faucett & Associates, October 1984.
3-33
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-------
4. URANIUM FUEL CYCLE FACILITIES
4.1 INTRODUCTION
The uranium fuel cycle includes uranium mills, uranium
hexafluoride conversion facilities, uranium enrichment
facilities, light-water reactor fuel fabricators, light-water
power reactors, and fuel reprocessing plants. With the exception
of the uranium enrichment facilities that are owned by the
Federal government and operated by contractors under the super-
vision of the Department of Energy (DOE), these facilities are
licensed by the Nuclear Regulatory Commission (NRC) or the
Agreement States. Releases of radioactive materials from these
facilities during normal operation are subject to the limits
established by 40 CFR 190. 40 CFR 190 limits the exposure to any
member of the general public from radionuclides released to air
or water to 25 mrem/y to the whole body or to any organ except
the thyroid, which is limited to 75 mrem/y. In addition, the
NRC requires releases of radioactive materials to be as low as
reasonably achievable (ALARA) below these regulatory limits.
As part of the current rulemaking, the EPA has performed a
dose and risk assessment of current airborne emissions from
uranium fuel-cycle facilities. The results of the dose and risk
assessment indicate that airborne emissions from operating
uranium mills cause greater doses and risks than those from the
uranium conversion, fuel fabrication, and light-water reactor
sectors of the fuel cycle.
4.1.1 Previous Evaluations
The potential public health impacts of the release of
radioactive materials into ambient air from the uranium fuel
cycle have been comprehensively evaluated. The EPA has prepared
a series of reports describing this evaluation. These reports
include:
U.S. Environmental Protection Agency, Environmental
Analysis of the Uranium Fuel Cycle - Part I - Fuel
Supply. EPA 520/9-73-003C, Office of Radiation
Programs, Washington, D.C., 1973;
U.S. Environmental Protection Agency, Environmental
Analysis of the Uranium Fuel Cvcle - Part II. Nuclear
Power Reactors. EPA 520/9-73-003C, Office of Radiation
Programs, Washington, D.C., 1973;
U.S. Environmental Protection Agency, A Radiological
Emissions Study at a Fuel Fabrication Facility. EPA
520/5-77-004, Office of Radiation Programs, Washington,
D.C., 1978;
4-1
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U.S. Environmental Protection Agency, Radiological
Impact Caused bv Emission of Radionuclides into Air in
the United States. EPA 520/7-79-006, Washington, D.C.,
1979;
U.S. Environmental Protection Agency, Final
Environmental Impact Statement for Remedial Action
Standards for Inactive Uranium Processing Sites, EPA
520/4-82-013-1, October 1982;
U.S. Environmental Protection Agency, Final
Environmental Impact Statement for Standards for the
Control of Byproduct Materials from Uranium Ore
Processing. EPA 520/1-83-008-1, September 1983;
U.S Environmental Protection Agency, Radionuclides.
Background Information Document for Final Rules. EPA
520/1-84-022, Office of Radiation Programs, October
1984; and
U.S. Environmental Protection Agency, Final Rule for
Radon-222 Emissions from Licensed Uranium Mill
Tailings. Background Information Document. EPA 520/1-
86-009, August 1986.
4.1.2 Scope of the Evaluation
The segments of the uranium fuel cycle addressed in this
chapter include:
1. Uranium mills and their associated tailings piles;
2. Uranium conversion facilities;
3. Fuel fabrication facilities; and
4. Nuclear power facilities.
Each of these categories is addressed in the following
sections, which include a general description of each facility's
characteristics, processes, emission controls, radionuclide
emissions, and predicted radiation dose equivalent rates and
health risks to nearby individuals and the populations within
80 kilometers of these facilities. In addition, for categories
with the highest exposures, supplementary control options and
costs are presented here.
The assessment of doses and risks shows that particulate
releases from operating uranium mills cause some members of the
general public to receive organ dose equivalents greater than
25 mrem/y; for nearby individuals, estimates of the dose
equivalent to the lungs and the endosteum are as high as 120 and
85 mrem/y, respectively. The nearby individuals at greatest risk
are estimated to have a lifetime fatal cancer risk of 2E-4. The
4-2
-------
basis for these estimates and the detailed results are presented
in the following sections.
The assessment of uranium mills addresses only particulate
emissions. Radon emissions from the tailings are addressed in
Chapter 9. The uranium enrichment plants are included in the
assessment of DOE facilities (see Chapter 2). As there are no
operable fuel reprocessing plants in the United States, and since
reprocessing is prohibited under current policies, this segment
of the uranium fuel cycle has not been evaluated. High-level
waste disposal facilities are addressed in Chapter 5.
4.2 URANIUM MILLS
4.2.1 General Description
4.2.1.1 Uranium Mill Operations in the United States
Uranium mills extract uranium from ores which contain only
0.01 to 0.3 percent ^03. Uranium mills, typically located near
uranium mines in the western United States, are usually in areas
of low population density. The product of the mills is shipped
to conversion plants, where it is converted to volatile uranium
hexafluoride (UFe) which is used as feed to uranium enrichment
plants.
As of December 1988, of 27 uranium mills in the United
States licensed by the NRC or Agreement States, 4 were operating,
8 were on standby, 14 were being decommissioned, and 1 had been
built but never operated. The 8 mills on standby could resume
operations, but the 14 mills that are being decommissioned will
never operate again. The status of each mill is presented in
Table 4-1. The status descriptions used in this document are not
necessarily the same as the license definitions. Umetco's Uravan
mill is listed as on standby; however, since the mill's tailings
impoundment is being reclaimed, the mill is considered to be
decommissioned for the purpose of this assessment.
The operating mills have a capacity of 9,600 tons of ore per
day. The number of operating mills is down considerably from
1981, when 21 mills were processing approximately 50,000 tons of
ore per day.
4.2.1.2 Process Description
The mined ore is stored on pads prior to processing.
Crushing and grinding and a chemical leaching process separate
the uranium from the ore. The uranium product is recovered from
the leach solution and then dried and packaged. The waste
product (mill tailings) is piped as a slurry to a surface
impoundment area (tailings pile).
4-3
-------
Table 4-1. Uranium mills licensed by the U.S. Nuclear Regulatory
Commission as of December 1988.
Licensee
Common Name
Location
Rated
Capacity^*)
(t ore/d) Status(fe) Process(c>
American Nuclear
Corp.
Anaconda
Atlas Minerals
Bear Creek
Uranium Co.
Bokum Resources
Chevron Resources Co.
Conoco-Pioneer
Cotter Corp.
Dawn Mining Co.
Exxon
Exxon Minerals
Homestake Mining Co.
BF American
Minerals Exploration
Pathfinder Mines «
Pathfinder Mines
Petrotomics
Plateau Resources
Quivira
Rio Algom
TVA
Umetco Minerals Corp.
Umetco Minerals Corp.
Umetco Minerals Corp.
UNC Mining & Milling
Western Nuclear Inc.
Western Nuclear Inc.
Gas Hills, WY
950
Bluewater
Moab
Bear Creek
Panna Maria
Canon
Dawn
Ray Point
Highland
Homestake
L-Bar
Sweetwater
Lucky Me
Shirley Basin
Petromics
Shootaring
Ambrosia
La Sal
Edgemont
Gas Hills
White Mesa
Uravan
Church Rock
Split Rock
Sherwood
Bluewater, NM 6000
Moab, UT 1400
Converse Co. WY 2000
Marquez, NM 2000
Panna Maria, TX 2500
Falls City, TX 3400
Canon City, CO 1200
Ford, WA 450
Ray Point, TX
Converse Co., WY 3200
Grants, NM 3400
Seboyeta, NM 1600
Sweetwater Co., WY 3000
Gas Hills, WY 2500
Shirley Basin, WY 1700
Shirley Basin, WY 1500
Shootaring Cnyn, UT 750
Ambrosia Lake, NM
La Sal, UT 750
Edgemont, SD
Gas Hills, WY 1400
Blanding, UT 2000
Uravan, CO 1300
Church Rock, NM 3000
Jeffrey City, WY 1700
Wellpinit, WA 2000
Status Codes:
1 - Facility Operating
2 - Facility on Standby
3 - Facility Decommissioned or
Being Decommissioned
4 - Facility Built, Never Operated
Data Sources:
(a) Tons of ore/day (Jo81).
(b) Personal communication with
Dale Smith, USNRC, Denver, Colorado
(c) From Ri81.
Process Codes:
1 - Acid leach
2 - Alkaline leach
3 - Solvent extraction
4 - Carbonate leach
5 - Eluex
6 - Caustic precipitation
7 - Column ion exchange
3
3
3
1,5
1,3
2,3
1,3
4
1
3
2
3
3
3
1
3
2
2
1
3
2
2
2
3*
3
1
2**
3
3
2
1,3
1.3
1,3
1,3
1,3
-
1,3
4,6
-
1,3
1,3
1,3
1,3
1,3
-
4,6
-
1,5
1,7
1,3
1,3
1,3
1.3
*
**
Decommissioning and long-term stabilization complete.
Per public comment by Umetco, the mill is being maintained on
standby although the tailings impoundment is being reclaimed.
4-4
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Radioactive materials released to the air during these
operations include natural uranium and thorium and their
respective decay products (e.g., radium, lead, radon). These
radionuclides, with the exception of radon, are released as
particulates.
4.2.1.2.1 Ore Storage
Ore is hauled from the mine in trucks. A minimum 10-day
supply of ore is kept on storage pads, which are several hectares
in area. The ore is transferred to the mill crushing unit via
front-end loaders or bulldozers. Although the ore is usually
moist upon receipt at the storage pad, it can become dry during
storage. The transfer operations, as well as wind erosion,
result in dust formation and release of radioactive material in
particulate form.
4.2.1.2.2 Milling
The process of extracting uranium from ore starts with
crushing and grinding. The ores are crushed dry, but water is
added during the grinding process. Some of the newer mills use a
one-step wet process called semi-autogenous grinding which
eliminates the dry ore crushing step.
The next step consists of leaching uranium out of the ore
and separating the uranium product from the leach solution.
There are two basic leaching processes: acid leaching for ores
with low lime content, and alkaline or carbonate leaching for
ores with high lime content. The leach solution is then
chemically treated to remove the uranium product. Most mills
that use the acid leaching process follow with solvent
extraction, a process where the uranium product is separated from
the solution by an organic solvent and is then separated from the
solvent by a stripping and precipitation operation. The mills
that use the alkaline or carbonate leaching process add a caustic
to the leach solution, resulting in the precipitation of sodium
diuranate. In both cases, the product is dried in large ovens
and packaged in 55-gallon drums.
The steps that generate significant radioactive emissions
are the dry operations: crushing, drying, and packaging. The
intermediate stages are carried out wet in enclosed vessels and
do not produce significant amounts of airborne emissions.
4.2.1.2.3 Tailings
After the uranium product is separated from the ore in the
leaching process, the residual ore is pumped as a slurry to a
tailings impoundment area. A tailings pile, typically about
100 hectares in area, is surrounded by an embankment of
impervious material. The liquid portion of the slurry is par-
tially recovered and recycled by some mills and is allowed to
evaporate at other mills. The solid tailings are made up of a
»
4-5
-------
sand fraction (particles from 38 to 200 mesh) and a slime fraction
(particles smaller than 200 mesh).
An active tailings pile contains wet and dry areas. The
slurry feed pipe is moved around the impoundment area to keep the
pile level; therefore, the pile has a pond area where the slurry
is fed while the rest of the pile is drying out.
As sections of the pile dry, the tailings become a source of
windblown dust. The slime component, the most likely to become
airborne because of its small particle size, contains uranium
concentrations twice as high as the sands (NRC79).
4.2.1.3 Existing Emission Controls
4.2.1.3.1 Ore Storage
Dust from ore storage pads can be controlled by the use of
windbreaks and water sprays. Windbreaks are concrete or wood
fences around the pile which reduce the amount of wind blowing
across the pile. This reduces the drying effect of the wind, as
well as reducing the tendency of the wind to pick up dust.
Ore piles with a moisture content of 4 percent or more do
not cause dust problems (NRC79). Spraying the pile increases the
moisture content of the ore. A tank truck with pumps and hoses
can be used for spraying.
4.2.1.3.2 Milling
Dust is controlled during the crushing process by placing
air exhaust hoods at the crusher, screens, and transfer points.
The exhaust air passes through a dust collector.before it is dis-
charged to the atmosphere through a roof vent. As indicated
earlier, if a semi-autogenous grinding process is used, then the
dry crushing step is eliminated and essentially no dust is
emitted.
The off-gas from the drying oven passes through a dust
separation system before discharge to the roof vent. Air exhaust
hoods are placed in the packaging area, and the exhaust is passed
through a dust collector before being vented.
The primary method of removing dust from the exhaust gas is
the wet scrubber. Wet scrubbers remove dust particles by impact-
ing them with water droplets. The most common type of wet
scrubber is the orifice scrubber, which has a removal efficiency
of 93.6 percent. Also common is the impingement scrubber, which
has a removal efficiency of 97.9 percent. The venturi scrubber,
used infrequently, has a removal efficiency of 99.5 percent but
requires more energy to operate than the other two scrubber
types. The removal efficiencies presented are those cited by the
NRC for these applications (NRC79).
4-6
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Baghouses are frequently used to remove dust from the
crushing and packaging area exhaust. The exhaust air is passed
through bag filters made of woven or felted material. Baghouses
have a rated removal efficiency of 99.9 percent. They are not
suitable for cleaning the dryer off-gas because of the high
temperature and moisture content.
4.2.1.3.3 Tailings
Control of dust from a tailings pile is similar to control
of dust from the ore storage pad. The tailings pile can be kept
wet by truck spraying or by discharging the slurry from multiple
discharge points instead of one point.
An alternative method of dust control for tailings surfaces
that are not being added to or disturbed is to put a chemical
stabilizer on the surface of the pile. Some stabilizers mix with
the tailings to form a crust. Other materials, such as asphalt
sprays, form a thin film on the pile surface. Both methods are
temporary and require annual maintenance.
4.2.2 Basis for the Dose and Risk Assessment of Uranium Mills
The following sections describe the basis for the site-
specific and model facilities used to assess the airborne re-
leases of radionuclides from uranium mills. Information on the
source term, meteorological, and demographic data assumed are
also presented. Detailed information on the parameters supplied
to the AIRDOS/DARTAB/RADRISK computer codes is presented in
Appendix A. Site-specific source term, meteorological, and
demographic data were supplied as input to the assessment codes
for the four operating mills and for six of the seven mills on
standby. Cotter Corporation's Canon City mine, which is on
standby, currently has no dry tailings piles and therefore was
not included in the assessment. A generic model mill was used
for the assessment of doses and risks from tailings piles of
mills that are either decommissioned or undergoing decommission-
ing. Outputs of the codes include estimates of: dose equivalents
to the most exposed individuals (mrem/y); lifetime fatal cancer
risk to the most exposed individuals; dose equivalents to the
regional (0-80 km) population (person-rem/year); and the number
of cancer deaths in the regional population per year of operation
(deaths/year).
4.2.2.1 Radionuclide Emissions
The magnitude of releases from uranium mills differs for
operating and shutdown facilities. Therefore, in addition to
measured process releases reported to the NRC, models were
developed to represent windblown particles from active tailings
and windblown releases from dry tailings piles where operations
have ceased and final stabilization has not yet occurred.
4-7
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4.2.2.1.1 Operational Experience and Projected
Future Emissions
The drying area and the crushing area are the major sources
of process releases at a typical plant. Ninety percent of the
uranium-234 and uranium-238 released come from the dryer area at
the end of the process. On the other hand, thorium-230 and
radium-226 emissions result primarily from operations, such as
crushing, that occur at the beginning of the process.
Although the number of operating uranium mills has decreased
sharply over the last decade, the demand for yellowcake has been
steadily increasing as more nuclear power plants have come on
line. Yellowcake from foreign sources has supplied an increasing
percentage of demand. However, the number of operating mills is
expected to stabilize or perhaps even increase slightly in the
near future. Radionuclide releases from uranium milling opera-
tions should be proportional to the quantity of uranium ore
processed.
4.2.2.1.2 Development of Source Term for Assessments
The source terms for operating uranium mills and mills on
standby include particulate radionuclides released to air from
process exhausts and those blown from the dry areas of the
tailings impoundments. The source terms used in the assessment
for operating mills, mills on standby, and a generic inactive
tailings impoundment, which was used to model decommissioned
mills, are presented in Table 4-2.
The source terms presented here for the operating facilities
differ from those presented in the draft document due to the use
of more current information concerning the total area of wet and dry
tailings and the concentration of radium-226 in the tailings at
each of the facilities. Also, source terms for mills on standby
are now presented wereas they'were not originally included in the
draft document.
The release rates (Ci/y) for process exhausts are based on
measurements of natural uranium, thorium-230, and radium-226.
These data were obtained for three of the four mills from the
semi-annual environmental monitoring reports submitted by the
mills to the Nuclear Regulatory Commission. Whereas Panna Maria
was not included in the original assessment due to an inability
to obtain measured process release rate data, the mill has now
been included using, information obtained from Chevron Resources
Company.
Tailings pile emissions are not measured by the mill
operators, since the size of the tailings impoundments makes
measurement of windblown releases impractical. Therefore, the
release rates (Ci/y) from the tailings presented in Table 4-2
were calculated using the methodology presented in NRC's
Regulatory Guide 3.59, and the areas of dried tailings and
4-8
-------
average radium concentrations shown in Table 4-3, using dusting
factors appropriate for the site meteorology and tailings pile
characteristics presented in EPA86.
The analysis includes consideration of the predominant
periods of tailings resuspension and dispersion during episodes
of high wind speed. No data were found showing particle size
distributions for process dusts. Particle size distributions for
tailings dusts show that approximately 30 percent of the
particles are in the respirable size range of 10 microns or less
(NRC80). Only the respirable fraction of the total dusts was
included in the assessment, and an activity median aerodynamic
diameter (AMAD) of 3.'0 microns, consistent with the data for
tailings dusts, was assumed. Data on lung clearance
classifications for windblown tailings could not be found.
Therefore, the default values recommended by the ICRP were used
for all radionuclides blown from the tailings.
Tailings pile release rates for the Canon City mill are not
shown in Table 4-2 since the site currently has no dried tailings
impoundments. Tailings release rates for Umetco Minerals
Corporations's Uravan mill are also not included. Although the
Uravan mill is on standby, the tailings impoundment is being
reclaimed. Thus, for the purposes of this assessment, the Uravan
mill is considered to be decommissioned and is therefore modeled
using the model inactive tailings impoundment.
The lung clearance classifications for uranium from process
exhausts are based on solubility studies of yellowcake in simu-
lated lung fluid (Co74, De79, De82, and Ka80). The classifica-
tions used for thorium, radium, lead, and polonium are the
default values recommended by the ICRP (ICRP66).
The NRC has calculated emissions from tailings piles from
several specific mills. These values range from 2.0E-4 to
2.7E-3 Ci/y for uranium-238/uranium-234, 3.3E-3 to 5.2E-2 Ci/y
for thorium-230, and 3.2E-3 to 5.5E-2 Ci/y for radium-226
(EPA79).
Annual radionuclide releases from tailings of the model
inactive mill, for which permanent stabilization has not been
performed, are also presented in Table 4-2. Methodology for
calculating these emissions was the same as that for calculating
emissions from tailings of active mills. The higher rate of
emissions for the inactive tailings pile is attributable to the
reduced moisture content of the inactive tailings and the
increased pile size.
4-9
-------
Table 4-2. Source terms for uranium milling.
Release Rate (Ci/y)
Radionuclide Lung Clearance AMAD Process ExhaustTailings
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
CHEVRON'S PANNA MARIA MILL(a)
Y
D
Y
D
Y
D
Y
W
D
W
3
3
3
3
3
3
0
0
0
0
0
0
1,
1,
3.0
3.0
3.0
3.0
1.9E-3
1.9E-3
1E-5
1E-5
1.9E-3
1.9E-3
9.6E-5
3.8E-6
3.8E-6
3.8E-6
HOMESTAKE'S HOMESTAKE MILL
Y
D
Y
D
Y
D
Y
W
D
W
3
3
3
3
3
3
3
3
0
0
0
0
0
0
0
0
3.0
3.0
1.7E-1
1.7E-1
8.3E-4
8.3E-4
1.7E-1
1.7E-1
4.3E-2
3.9E-2
3.9E-2
3.9E-2
MINERALS EXPLORATION'S SWEETWATER MILL(a)
Y
D
Y
D
Y
D
Y
W
D
W
3.0
3.0
3.0
3.0
3.0
3.0
3.0
3.0
3.0
3.0
-(d)
1.0E-4(C)
7.1E-7(C)
1.0E-4(C)
1.0E-3(C)
1.0E-3(C)
1.0E-3(C)
1.0E-3(C)
4.3E-3
3.0E-5
4.3E-3
4.3E-2
4.3E-2
4.3E-2
4.3E-2
(a) Source term added to those originally included in draft
document to reflect data obtained during comment period.
(b) Panna Maria currently has no dry tailings impoundments.
(c) Changes in source terms with respect to the draft document
reflect information on tailings areas and radium-226
concentrations obtained during comment period.
(d) Mill is currently on standy.
4-10
-------
Table 4-2. Source terms for uranium milling (continued).
Release Rate (Ci/y)
Radionuclide Lung Clearance AMAD Process Exhaust Tailings
PATHFINDER'S LUCKY MC MILL(a)
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
Y
D
Y
D
Y
D
Y
W
D
W
3.0
3.0
3.0
,0
0
,0
0
0
0
3
3,
3
3,
3,
3,
3.0
PATHFINDER'S SHIRLEY BASIN MILL
Y
D
Y
D
Y
D
Y
W
D
W
3
3
3
3
3.0
3
3
3
3
3.0
1.1E-2
1.1E-2
8.0E-5
8.0E-5
1.1E-2
1.1E-2
1.9E-4
5.9E-4
5.9E-4
5.9E-4
PLATEAU RESOURCES' SHOOTARING MILL(a)
Y
D
Y
D
Y
D
Y
W
D
W
3.0
3.0
3.0
3.0
0
0
0
0
0
3.0
1.1E-3
8.0E-6
1.1E-3
1.1E-2
1.1E-2
1.1E-2
1.1E-2
5.4E-3(C)
3.9E-5(°)
5.4E-3(°)
5.4E-2(C)
5.4E-2(C)
5.4E-2(C)
5.4E-2(C)
2.0E-4
1.4E-6
2.0E-4
2.0E-3
2.0E-3
2.0E-3
2.0E-3
(a) Source term added to those originally included in draft
document to reflect data obtained during comment period.
(b) Mill is currently on standby.
(c) Changes in source terms with respect to the draft document
reflect information on tailings areas and radium-226
concentrations obtained during comment period.
4-11
-------
Table 4-2. Source terms for uranium milling (continued).
Release Rate (Ci/y)
Radionuclide Lung Clearance AMAD Process Exhaust Tailings
QUIVIRA'S AMBROSIA LAKE MILL(a)
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
Po-210
U-238
U-238
U-235
U-235
U-234
U-234
Th-230
Ra-226
Pb-210
PO-210
Y
D
Y
D
Y
D
Y
W
D
W
3.0
3.0
3.0
3.0
3.0
3.0
3.0
0
0
-(b)
3,
3,
3.0
RIO ALGOM'S LA SAL MILL
Y
D
Y
D
Y
D
Y
W
D
W
3.0
3.0
,0
,0
,0
,0
,0
3
3
3
3
3
3.0
3.0
3.0
2.8E-2
2.8E-2
2.1E-4
2.1E-4
2.8E-2
2.8E-2
l.OE-4
2.8E-4
3.3E-4
3.3E-4
UMETCO'S WHITE MESA MILL
Y
D
Y
D
Y
D
Y
W
D
W
3.0
3.0
3.0
0
0
0
0
3
3
3
3
3.0
3.0
3.0
2.1E-2
2.1E-2
1.5E-4
1.5E-4
2.1E-2
2.1E-2
4.9E-4
4.8E-4
1.2E-3
1.2E-3
1.1E-3
7.5E-6
1.1E-3
1.1E-2
1.1E-2
1.1E-2
1.1E-2
1.4E-4(C)
1.4E-4(C)
1.4E-3(C)
1.4E-3(°)
1.4E-3(C)
1.4E-3(C)
(a) Source term added to those originally included in draft
document to reflect data obtained during comment period.
(b) Mill is currently on standby.
(c) Changes in source terms with respect to the draft document
reflect information on tailings areas and radium-226
concentrations obtained during comment period.
(d) La Sal currently has no dry tailings impoundments.
4-12
-------
Table 4-2. Source terms for uranium milling (continued).
Release Rate (Ci/y)
Radionuclide Lung Clearance AMAD Process ExhaustTailings
WESTERN NUCLEAR INC.'S SHERWOOD MILL(a)
U-238 Y 3.0 -(b) l.OE-3
U-238 D 3.0
U-235 Y 3.0 - 7.1E-6
U-235 D 3.0
U-234 Y 3.0 - l.OE-3
U-234 D 3.0
Th-230 Y 3.0 - l.OE-2
Ra-226 W 3.0 - l.OE-2
Pb-210 D 3.0 - l.OE-2
Po-210 W 3.0 - l.OE-2
MODEL INACTIVE TAILINGS PILE(C)
U-238 Y 3.0 8.0E-3
U-235 Y 3.0 5.8E-5
U-234 Y 3.0 8.0E-3
Th-230 Y 3.0 8.0E-2
Ra-226 W 3.0 8.OE-2
Pb-210 D 3.0 8.OE-2
Po-210 W 3.0 8.OE-2
(a) Source term added to those originally included in draft
document to reflect data obtained during comment period.
(b) Mill is currently on standby.
(c) After closure, prior to stabilization.
4-13
-------
Table 4-3
Mill
Areas of the tailings impoundments at uranium
mills and average radium-226 concentrations.(a)
Total Area Wet Area Dry Area Radium-226
(acres/ha) (acres/ha) (acres/ha) (pCi/g)
New Mexico
Ambrosia Lake
- Secondary
- Lined Ponds
Homestake
Texas
Panna Maria
121/49
280/113
210/85
160/65
13/5
162/66
140/57
160/65
108/44
118/47
70/28
0/0
237
22
300
198
La Sal
Shootaring
White Mesa
Washington
Sherwood
Wyoming
Lucky Me
- Piles 1,2, &
- Evap. Ponds
Shirley Basin
Sweetwater
Inactive Tailings
93/38
7/3
30/53
80/32
203/82
104/42
275/111
37/15
79/32
93/38
3/1
125/51
40/16
143/58
104/42
215/87
30/12
0/0
0/0
4/2
5/2
40/16
60/24
0/0
60/24
7/3
79/32
420
280
981
200
220
22
208
280
280
(a) The data in this table has changed with respect to the draft
document in response to information recieved during the
comment period.
4.2.2.2 Dispersion Parameters
In modeling the releases from the mills, both a stack source
and an area source were used to represent the process and
tailings releases respectively. A 12-meter stack with a
1.2-meter diameter and volumetric flow of 12.7 meters was used for
process exhausts. The total area (wet and dry) of the tailings
impoundments was used for the size of area sources.
Meteorological data from the nearest meteorological station
with joint frequency data in the form required by the assessment
codes were used for the active mills. For the inactive tailings,
generic meteorological data presented in NRC80 were used. The
sources of the meteorological data used for each assessment are
presented in Table 4-4.
4-14
-------
Table 4-4. Sources of meteorological data used in the assessment
of uranium milling.
Mill
Location
Meteorological
Station
New Mexico
Ambrosia Lake
Homestake
Texas
Panna Maria
Utah
La Sal
Shootaring
White Mesa
Washington
Sherwood
Wyoming
Lucky Me
Shirley Basin
Sweetwater
Inactive Tailings
Ambrosia Lake, NM Ambrosia Lake, NM
Grants, NM Ambrosia Lake, NM
Panna Maria, TX
La Sal, UT
Hanksville, UT
Blanding, UT
Wellpinit, WA
Riverton, WY
Casper, WY
Rawlings, WY
Generic (see text)
San Antonio, TX
Grand Junction, CO
Farmington, NM
Spokane, WA
Casper, WY
4.2.2.3 Demographic Data
The actual populations living within 5 km of the operating
mills were enumerated by sector segments during site visits made
to each mill in 1983 (PNL84). The data for Canon City, Ambrosia
Lake, Homestake, La Sal, and Sherwood were updated following site
visits by SC&A in 1989. These distributions, presented in Table
4-5, were used in conjunction with the population distributions
for 5 to 80 km generated by the computer code SECPOP from 1980
U.S. Census Bureau data. The population distribution for the
generic tailings pile was taken from NRC80.
Actual data on food production in the vicinity of these
mills were not obtained. Instead, generic food production rates
(urban/low productivity) representative of the areas where these
mills are located were used in the assessment.
4-15
-------
Table 4-5. Estimated populations living within 0 to 5 km of
active uranium milling facilities.(a)
Mill 0-0.5 0.5-1.0 1.0-2.0 2.0-3.0 3.0-4.0 4.0-5.0
km km km km km km
New Mexico
Ambrosia Lake* 0 0 0 0 o 0
Homestake* 0 0 187 104 42 57
Texas
Panna Maria 0 12 42 33 81 285
Utah
La Sal* 00 0 0 40 0
Shootaring 00000 171
White Mesa 000008
Washington
Sherwood* 00 0 0 32 17
Wyoming
Lucky Me 0 0 0 0 0 0
Shirley Basin 0 0 0 0 0 o
Sweetwater 000000
(a) The data source is PNL84 except where marked with an *.
These data were updated following site visits by SC&A in
1989.
4.2.3 Results of the Dose and Risk Assessments of Uranium Mills
The AIRDOS-EPA/DARTAB/RAPRISK assessment codes estimate the
50-year committed dose equivalents to organs from exposure via
air immersion, ground-surface, inhalation, and ingestion
pathways. Table 4-6 presents the results of the dose assessment
to nearby individuals and to the regional (0-80 km) populations
around uranium milling facilities. The organs listed in Table 4-6
are those where the dose is estimated to contribute 10 percent or
more of the total fatal cancer risk.
4-16
-------
Table 4-6. Estimated radiation dose rates from uranium mills.
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Mill
Organ Process
New Mexico
Ambrosia Lake Lungs
Endosteum
Red Marrow
Remainder
Homestake Lungs 8 . 7E+1
Endosteum 4.9E+1
Red Marrow
Remainder
Texas
Panna Maria Lungs 2.0E+0
Endosteum
Remainder
Utah
La Sal Lungs l.OE+0
Endosteum
Red Marrow
Shootaring
White Mesa
Washington
Sherwood
Wyoming
Lucky Me
Remainder
Lungs
Endosteum
Red Marrow
Remainder
Lungs 3.5E-1
Endosteum
Red Marrow
Remainder
Lungs
Endosteum
Red Marrow
Remainder
Lungs
Endosteum
Red Marrow
Remainder
Tailings
8.2E-2
2.8E-1
2.2E-2
7.0E-3
3.6E-1
1 . 1E+0
8.9E-2
9.
3.
2.
6.
1.
5.
4.
4.
1.
1.
2.
3.
1.
1.
4.
-
8E-2
1E-1
5E-2
2E-3
5E-3
OE-2
OE-3
-
2E-1
3E+0
1E-1
6E-2
7E-2
4E-1
1E-2
9E-3
Total
8,
2,
2,
7
8
5
8
2
1
9
3
2
6
3
5
4
4
1
1
2
3
1
1
4
,2E-2
.8E-1
.2E-2
,OE-3
.7E+1
.OE+1
.9E-2
.OE+0
NA
NA
.OE+0
-
.8E-2
.1E-1
.5E-2
.2E-3
.5E-1
.OE-2
.OE-3
-
.2E-1
.3E+0
.1E-1
.6E-2
.7E-2
.4E-1
.1E-2
.9E-3
Process Tailings
7
3
2
1
9 . 7E+1 4
6 . 7E+1 1
1
1 . 8E+0
1.4E+0
l.OE-1
9.7E-1
1 . 1E+0
9.8E-2
1
7
5
2
7.1E-1 3
7.7E-1 1
1
6.6E-2 9
1
1
8
8
1
9
7
6
.4E-1
.3E+0
.6E-1
.5E-1
.2E-1
.4E+0
.1E-1
-
.9E-2
.OE-2
.5E-3
.1E-3
.OE-2
.6E-1
.3E-2
.OE-3
.OE+0
.OE+1
.1E-1
.1E-1
.1E-1
.3E-1
.2E-2
.7E-2
Total
7.4E-1
3 . 3E+0
2.6E-1
1.5E-1
9.7E+1
6 . 8E+1
1.1E-1
1 . 8E+0
1 . 4E+0
l.OE-1
9.7E-1
1 . 1E+0
9.
1.
7.
5.
2.
7.
9.
1.
7.
1.
1.
8.
8.
1.
9.
7.
6.
8E-2
9E-2
OE-2
5E-3
1E-3
4E-1
3E-1
3E-2
5E-2
OE+0
OE+1
1E-1
1E-1
1E-1
3E-1
2E-2
7E-2
4-17
-------
Table 4-6,
Estimated radiation dose rates from uranium mills (continued).
Nearby Individuals
(mrem/y)
Regional Population
(person-rem/y)
Mill
Organ
Process Tailings Total ProcessTailingsTotal
Wyoming (cont.
Shirley Basin
Sweetwater
Inactive
Tailings
j.
Lungs
Endosteum
Red Marrow
Remainder
Lungs
Endosteum
Red Marrow
Remainder
Lungs
Endosteum
Red Marrow
Remainder
7.4E-2 2.0E-1
7.1E-1
5.6E-2
2.0E-2
2.7E-1
9.2E-1
7.3E-2
2.5E-2
9.8E+1
3 . 1E+2
2 . 5E+1
"
2.7E-1
7.1E-1
5.6E-2
2.0E-2
2.7E-1
9.2E-1
7.3E-2
2.5E-2
9.8E+1
3 . 1E+2
2 . 5E+2
-
3.9E-1 1.1E+0
8.6E-1 l.OE+1
7.9E-1
7.6E-2 7.5E-1
3.2E-1
2 . 4E+0
1.9E-1
1.7E-1
2 . 2E+0
1 . 6E+1
1 . 2E+0
1 . OE+0
1 . 5E+0
1 . 1E+1
7.9E-1
8.3E-1
3.2E-1
2 . 4E+0
1.9E-1
1.7E-1
2 . 2E+0
1 . 6E+1
1 . 2E+0
1 . OE+0
The lifetime fatal cancer risks to nearby individuals and
the estimated deaths per year in the regional populations are
shown in Table 4-7 for each mill. The estimated distribution of
the total fatal cancer risk from all mills and the number of
persons at each risk interval are presented in Table 4-8. The
values of fatal cancer risk distribution from the model inactive
tailings pile were multiplied by 15 to obtain an estimate of the
distribution from all decommissioned mills. The results for the
four operating mills and the seven mills on standby were added to
obtain the distribution from all mills.
The only significant pathways for dose and risk are
inhalation and ingestion. For nearby individuals, inhalation is
generally predominant; for regional populations, ingestion is
more important. For nearby individuals, the most significant
nuclides released from tailings piles are thorium-230 and
lead-210, while the most important plant emissions are
uranium-238 and uranium-234. For regional populations, the most
important nuclide released from tailings piles is lead-210, but
thorium-230, polonium-210, and radium-226 are also emitted in
significant quantities. Of nuclides emitted from process stacks
uranium-238 and uranium-234 contribute the most to population '
dose and risk with, in some cases, less important contributions
from lead-210 and thorium-230.
4-18
-------
Table 4-7. Estimated fatal cancer risks from uranium mills.
Nearby Individuals Regional (0-80 km)
Lifetime Fatal Population
Facility Cancer Risk Deaths/y
New Mexico
Ambrosia 2E-7 3E-5
Homestake 2E-4 2E-3
Texas
Panna Maria 3E-6 5E-5
Utah
La Sal 2E-6 3E-5
Shootaring 2E-7 7E-7
White Mesa 6E-7 2E-5
Washington
Sherwood 1E-6 8E-5
Wyoming
Lucky Me 1E-7 7E-6
Shirley Basin 6E-7 9E-5
Sweetwater 7E-7 2E-5
Model Inactive Tailings 2E-4 1E-4
Table 4-8. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from uranium mills.
Risk Interval Number of Persons Deaths/y
1E-1 to
1E-2 to
1E-3 to
1E-4 to
1E-5 to
1E-6 to
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
< 1E-6
6,
32,
2,200,
0
0
0
84
500
000
000
0
0
0
2E-4
1E-3
2E-3
2E-3
Totals
2,200,000 5E-3
4-19
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4.2.4 Supplementary Control Options and Costs
4.2.4.1 Controls for Process Releases
The NRC has evaluated additional controls for the process
operations that result in significant airborne emissions
(NRC80). Several well-proven control technologies can be
employed on the ore crushing and yellowcake drying and packaging
exhausts. Table 4-9 presents the predicted efficiencies and
costs of these technologies. The lifetime costs shown in the
last column of the table are based on 15 years of operation.
Table 4-9. Effluent controls for process emissions.
Costs (thousands of 1980 dollars)
Control Efficiency, % Capital Annual Lifetime
Ore Crushing Exhaust Dust-Removal Units
Orifice 94 55 14 325
Wet Impingement 97.9 138 16.8 390
Low-Energy Venturi Scrubber 99.5 205 32.8 695
Fabric Filter 99.9 387 33.2 885
Fabric Filter & HEPA 407 91.3 1775
Yellowcake Drying and Packaging Exhaust Dust-Removal Units
Vet Impingement
Low-Energy Venturi Scrubber
Medium-Energy Venturi Scrubber
High-Energy Venturi Scrubber
High-Energy Venturi & HEPA
97. '9
99.5
99.7
99.9
45.0
55.5
66.1
71.5
108.2
5.5
10.8
15.9
23.8
29.4
130
220
305
430
550
4.2.4.2 Controls for Windblown Particulates
The solid portion of a dry tailings pile, particularly the
slime, is a source of radioactive contamination. The slime
contains uranium concentrations twice as high as the sand and,
due to its small particle size, becomes easily airborne. Several
alternatives have been identified to control potential contami-
nated dust problems from dry tailings: (a) wetting the tailings;
b) leaching the tailings to remove residual radioactivity;
c) fixation/solidification of the tailings; (d) application of
stabilizers to the surface of the piles to form a crust; and
e) covering of the tailings either above or below the ground
surface. The method most commonly used at milling operations is
wetting of the dry tailings by sprinkler trucks.
This section presents estimated capital, operating, and
maintenance costs for each of the alternatives listed above. The
4-20
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following assumptions form the basis of the cost analysis: a) the
tailings are generated at a rate of 675 metric tons (MT) per day
or, assuming assuming a six day work week, 209,000 MT per year; b)
the tailings are discharged to a 30-hectare (ha) site which is
surrounded by embankments approximately 8 meters (m) in height
(the embankments occupy an additional 16 ha); and (c) the
tailings will be generated over a 15-year period.
4.2.4.2.1 Wetting of Tailings
Wetting of dry tailings is the most common method used to
control dust from tailings piles. Water is applied to the
tailings by sprinkling from tank trucks or by a stationary
sprinkling system. Tailings pond water is used to minimize
costs. Homestake Mines, Atlas Minerals, and M. K. Ferguson Mines
in New Mexico, Utah, and Colorado, respectively, use this method
of dust control.
4.2.4.2.1.1 Tank Truck Application
The costs for this alternative have been estimated for both
rental and purchase and are based on the following assumptions:
(a) 20 ha per day will be sprinkled with 0.3 cm of water in
8 hours; (b) each truck will travel 50 km per day;
(c) 18,925-liter trucks will be used; (d) each 30-ha site will be
sprinkled every day; (e) four trucks will be required for the
operation; and (f) the yearly escalation for all costs during the
life of the project is 5 percent.
The average yearly cost for conducting the wetting operation
with the use of rented trucks would be approximately $549,000.
The total cost for this alternative over 15 years of operation is
estimated to be $8.2 million. The purchase of four trucks would
cost approximately $300,000. The average yearly operating and
labor costs over the life of the project would be the same as
those for the rental option, approximately $318,000. The esti-
mated total cost for this alternative over 15 years of operation
is $5.1 million.
4.2.4.2.1.2 Stationary Sprinkling System
A stationary sprinkling system is currently in use at
Homestake Mill in Grants, New Mexico, to control dust from mill
tailings piles. This method has also been used at mining sites
in Wyoming. Maintenance and labor costs are less for a station-
ary system than for the tank truck alternative.
The cost estimate for this alternative is based on the
following assumptions: (a) high-density polyethylene (HOPE) pipe
(laid on top of the piles) would be used due to the
caustic/acidic nature of the tailings pond water; (b) standard
irrigation sprinkler heads, set approximately 9 m apart, would be
used; (c) an electric pump would be used to move the tailings pond
water through the distribution system; (d) 0.5 cm of tailings pond
4-21
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water would be used each day; (e) the system would be expanded by
two ha each year during the life of the project; and f) each
component of the system would be replaced every five years.
The total cost of a stationary sprinkling system over the
15-year life of the project is approximately $1.9 million. The
average yearly cost is estimated to be $126,000. Fifty nine
percent of this estimate is the labor cost associated with the
installation and operation of the system. Homestake Mines has
installed and operated its system with an in-house maintenance
staff, thereby reducing the cost of the sprinkling system
considerably.
4.2.4.2.2 Leaching of Tailings
None of the mines mentioned above uses this technique, and
no recent studies have been conducted to determine the
feasibility of leaching tailings. Laboratory tests have shown
that 98 percent of the nuclides could be leached from the
tailings with nitric acid. However, the residual radium
concentrations would be at least an order of magnitude greater
than that typically found in western U.S. soils. Therefore,
after acid leaching, dust suppression would still have to be
effected for the dry tailings.
The cost to construct and operate a nitric acid leaching
mill for 15 years is estimated to be $283 million. This estimate
is based on the cost contained in the NRC's Draft Generic
Environmental Impact Statement on Uranium Milling as updated
using the 1988 ENR Construction Cost Index (NRC79).
4.2.4.2.3 Solidification of Tailings
Solidification agents such as concrete or asphalt can be
added to the tailings to control dust at the piles. This
technique had not been used at any of the mines contacted during
preparation of this cost analysis.
Asphalt fixation would require the construction of a
facility to heat the asphalt, mix the asphalt with the tailings,
and dry the asphalt/tailings mixture. The capital cost for
construction of this facility is estimated to be $6.6 million.
Approximately 0.75 MT of asphalt would be required for each
metric ton of dry tailings. The current cost of asphalt is
$33/MT, resulting in an estimated average annual cost of
$7.5 million over 15 years of operation. The fuel requirements
for the wiped film evaporator (to evaporate water from the
asphalt/tailings mixture) will be about 50 MT of coal per day.
The average annual cost of coal is estimated to be $1.2 million.
The total estimated cost for this alternative is $138 million.
The cost for constructing and operating mixing equipment and
related facilities required for solidifying the tailings with
cement is estimated to be $1.8 million. One part cement to five
4-22
-------
parts tailings would be required to solidify the tailings. The
current cost of cement is $66/MT. The average yearly cost of
cement is estimated to be $4 million over 15 years of operation.
The total estimated cost for this alternative is $62 million.
4.2.4.2.4 Application of Stabilizers to Tailings Surfaces
Various chemicals are being used to stabilize the surface of
tailings piles. These stabilizers are sprayed on the surface of
the piles to form a cover. Studies have shown that these
stabilizers are temporary control measures which require
continued inspection and maintenance. Neilson, Inc., of Durango,
Colorado, is responsible for dust control at M. K. Ferguson
Mines. It has been using a polymer (Nelco 8803) and a latex
binder (CPB 12), manufactured by WEENDON of Moab, Utah. The
polymer has been found to have a short life span, whereas the
latex binder proved to be effective for more than a year.
The current cost of applying a latex binder to tailings
piles is about $l,650/ha. The average annual cost for this
alternative is estimated to be $2,280/ha. The total estimated
cost for this alternative, assuming that each 30 ha would be
treated annually, is $1.03 million. If the tailings can be
deposited such that 2 ha of tailings are added to the tailings
pile each year, the total cost can be reduced by approximately
$400,000.
4.2.4.2.5 Covering of Tailings
Tailings can be covered with natural or artificial covers
either above or below the ground surface. Natural cover
materials include native soil, gravel, and clay. Artificial
materials include asphalt and plastic. Asphalt and plastic are
less effective than clay in withstanding mechanical stresses and
resisting deterioration in sunlight.
The most effective dust control plan for dry tailings is
provided by a combination of natural and artificial cover
materials. A cap consisting of a synthetic liner overlain by
sand and native soil (planted with native grasses) will reduce
infiltration of rain water, control tailings dust, and require
minimal maintenance. In arid regions, a clay cap with riprap on
the surface would be very effective in eliminating exposure to
airborne tailings dust.
The cost estimates for this alternative are based on the
following assumptions: (a) embankment construction for the
above-ground surface alternative would be completed in the first
year of operation; (b) the excavation of the disposal site for the
below-ground surface alternative would be completed in the first
year of operation; (c) deposition of the tailings would begin in
the first year; (d) a 0.6-m clay cap would be constructed in
either alternative and the source of the clay would be 50 km from
the embankment/excavation; (e) tailings compaction and covering
4-23
-------
would be performed throughout the 15 years of operation; and
(f) interim dust control such as wetting or application of
stabilizers to the surface of the piles would not be required
because the tailings would be continuously covered with capping
materials. These cost estimates also include design and
construction management costs and a yearly escalation of
5 percent.
4.2.4.2.5.1 Above-ground Encapsulation
Site preparation for above-ground encapsulation requires
removal of the topsoil over a 46-ha area and the construction of
earthen dikes along the periphery of the disposal area. Removal
of the top soil (276,000 cubic meters) is estimated to cost
$1.08 million ($3.91/m3). The cost for construction of the
earthen embankments would be approximately $4.05 million. The
embankments would be approximately 8 m in height, 10 m wide at
the top, and 42 m wide at the bottom. During deposition, the
tailings would be compacted and covered at an average annual cost
of approximately $1.2 million. The cover would consist of 2.7 m
of fill material from the site and 0.6 m of clay. The total cost
of this alternative is estimated to be $23 million.
A plastic liner could be added to the capping system to
increase its effectiveness or substituted for clay in areas where
clay is not available at a reasonable cost. PVC, HOPE, or
Hypalon cover material could be used at estimated total costs of
$2.6 million, $3 million, and $4.3 million, respectively. These
estimates assume that 2 ha would be covered each year during
15 years of operation. Many manufacturers highly recommend HOPE
for this particular use due to its resistance to ultraviolet
light deterioration. HOPE has a life expectancy of at least
10 years for application as a cover material.
In arid regions, 0.5 m of riprap could be-used in place of
top soil and seeding. The estimated cost for placing 150,000 m3
of riprap on the 30-ha site over 15 years of operation is
$3.9 million.
Another alternative is to solidify the top 0.5 m of the
encapsulation site with cement. The total portland cement
requirement would be 30,000 m3 which would be mixed with the top
2.5 m of tailings during 15 years of operation. The estimated
total cost for this alternative is $6.9 million.
4.2.4.2.5.2 Below-Ground Encapsulation
The differences between the costs for above- and below-ground
encapsulation are that for the below-ground alternative,
embankments would not have to be constructed, a disposal site
would have to be excavated, and the tailings would have to be
transported to the disposal site. The average yearly and total
costs estimated for compacting and covering the tailings are the
same as for the above-ground alternative.
4-24
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Excavation costs including loading, hauling, and depositing
materials less than 1 km from the excavation site are estimated
to be $10 million. The average yearly cost to excavate and
transport the tailings to the disposal site (assuming the site is
1 km from the tailings pond) is estimated to be $345,000. The
total estimated cost for this alternative is $33 million.
4.2.4.2.6 Summary
A summary of the estimated costs for each of the alterna-
tives is presented in Table 4-10.
Table 4-10. Estimated costs for alternatives to control
windblown particulates from tailings piles.
Estimated Costs
(Dollars in Millions)
Alternative
Wetting Using Rented Trucks
Wetting Using Purchased Trucks
Wetting Using Stationary System
Acid Leaching
Solidification with Asphalt
Solidification with Cement
Application of Latex Binders
Above-Ground Encapsulation
Below-Ground Encapsulation
Per
Hectare
0.27
0.17
0.06
9.40
4.60
2.10
0.03
0.77
1.10
Total
8.2
5.1
1.9
283.0
138.0
62.0
1.0
23.0
33.0
The application of latex stabilizers to the tailings piles
is the most cost-effective method for controlling dust from the
piles. This method is currently in use and has proved effective
for up to one year per application.
The stationary sprinkling system is the second most cost-
effective alternative. When installed and operated by existing
maintenance personnel, this alternative is more cost-effective
than the application of latex stabilizers. The added advantage
is that evaporation of the tailings pond water, an operational
goal of each milling operation, would be substantially increased.
4.3 URANIUM CONVERSION FACILITIES
4.3.1 General Description
The uranium conversion facility purifies and converts
uranium oxide (yellowcake) to volatile uranium hexafluoride
(UF6), the chemical form in which uranium enters the enrichment
plant.
4-25
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4.3.1.1 Uranium Conversion Operations in the United States
Currently, two commercial uranium hexafluoride
production facilities are operating in the United States, the
Allied Chemical Corporation facility at Metropolis, Illinois, and
the Kerr-McGee Nuclear Corporation facility at Seguoyah,
Oklahoma. The Allied Corporation facility, a dry-process plant
in operation since 1968, has the capacity to produce about
12,600 MT of uranium per year in the form of UFs. The Kerr-McGee
facility is a wet process plant in operation since 1970, with a
capacity of about 9,100 MT per year (AEC74, Do88) .
4.3.1.2 Process Description
Two industrial processes are used for uranium hexafluoride
production, the dry hydrof luor method and the wet solvent extrac-
tion method. Each method produces roughly equal quantities of
uranium hexafluoride; however, the radioactive effluents from the
two processes differ substantially. The hydrofluor method
releases radioactivity primarily in the gaseous and solid states,
while the solvent extraction method releases most of its radio-
active wastes dissolved in liquid effluents.
4.3.1.2.1 Dry Hydrofluor Process
The hydrofluor process consists of reduction, hydrof luorina-
tion, and f luorination of the ore concentrates to produce crude
uranium hexafluoride. Fractional distillation is then used to
obtain purified UFs. Impurities are separated either as volatile
compounds or as a relatively concentrated and insoluble solid
waste that is dried and drummed for disposal.
4.3.1.2.2 Solvent Extraction Process
The solvent extraction process employs a wet chemical
solvent extraction step at the start of the process to purify the
uranium for subsequent reduction, hydrof luorination, and
f luorination steps. The wet solvent extraction method separates
impurities by extracting the uranium from the organic solvent,
leaving the impurities dissolved in an aqueous solution. The
raffinate (barren waste from the solvent extraction process) is
impounded in ponds at the plant site.
4.3.1.3 Existing Emission Controls
No irradiated material is handled by conversion facilities;
therefore, the radionuclides present are those that occur in
nature. These radionuclides include thorium, uranium, and their
respective decay products. Uranium is the major source of
radioactivity in the emissions. Possible chemical species of
uranium effluents include ^Og, U02, UF4, UFg, (Nl^^^Oy, and
UO2F2.
4-26
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4.3.1.3.1 Dry Hydrofluor Process
Uranium emissions are higher in the dry hydrofluor process
than in the solvent extraction process, since large amounts of
dust are produced in the initial sampling, pre-treatment , and
reaction stages. During the low temperature steps such as
sampling, mixing, and crushing, exhaust systems that vent to
baghouses are used to control emissions. During high temperature
process steps that may emit gaseous as well as particulate
effluents, a combination of metal filters and scrubbers is used.
4.3.1.3.2 Solvent Extraction Process
In the wet solvent extraction method, uranium is present as
dissolved uranyl nitrate, a chemical species that may also
appear in emissions. Thus, uranium may be released as both
soluble and insoluble aerosols. The discharge to the environment
is through low stacks and vents.
4.3.2 Basis for the Dose and Risk Assessment of Uranium
Conversion Facilities
4.3.2.1 Radionuclide Emissions
4.3.2.1.1 Operational Experience and Projected Future
Emissions
The radionuclide emission rates given in Table 4-11 are
derived from measurements of releases from vents and stacks as
reported in the semi-annual environmental monitoring reports
submitted by the facilities to the NRC. These values are
averaged over the period 1984 to 1987.
Table 4-11. Reported atmospheric radioactive emissions for
uranium conversion facilities (Ci/y) .
Metropolis (a) Metropolis (k) Sequoyah(a)
Radionuclide 1984 - 1987 1979 - 1982 1984 - 1987
Ra-226 1.0 E-5 6.7 E-4 5.0 E-3
Th-230 5.0 E-4 6.6 E-3 5.0 E-3
U-Natural 1.0 E-l 2.2 E-l 5.0 E-2
(a)From semi-annual environmental monitoring reports, 1984
through 1987.
(b)From NRC84.
4-27
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Table 4-11 also includes measured data for Metropolis that
were obtained from 1979 to 1982. These values, in combination with
the 1984 to 1987 values, show the trend toward lower emission
rates for all radionuclides.
It is anticipated that the existing uranium conversion
plants will be able to accommodate future uranium demand by
nuclear power plants. The radionuclide emissions are propor-
tional to the quantity of uranium produced and thus should remain
relatively constant.
4.3.2.1.2 Source Terms Used in the Assessment
The annual atmospheric radioactive emissions assumed for
each conversion facility are presented in Table 4-12. These
values are averages of the measured releases for each facility
for 1984 through 1987.
4.3.2.2 Site Characteristics Used in the Assessment
The plant parameters used in the assessment are specific to
each site (NRC84, NRC85b). Each stack height is an average of
all release points for that plant. In calculating the average,
the data were weighted by the ventilation rate of each release
point. Detailed information on the parameters supplied to the
AIRDOS/DARTAB/RADRISK computer codes is presented in Appendix A.
The ingestion pathway food source data assume fractions
representative of an urban/low productivity site.
4.3.3 Results of the Dose and Risk Assessment of Uranium
Conversion Facilities
The estimated annual radiation dose equivalents and fatal
cancer risks from the uranium conversion facilities are presented
in Tables 4-13 and 4-14.
The annual radiation dose equivalents from both the dry and
wet conversion processes result primarily from exposure to
uranium-234 and uranium-238 (51 percent and 46 percent for the
dry process, respectively; 39 and 35 percent for the wet process,
respectively). In the wet process, there is also about a
22 percent contribution from thorium-230. Inhalation is the
dominant exposure pathway in each case.
4.3.3.1 Doses and Risks to the Nearby Individual
Doses and fatal cancer risks to the nearby individuals are
presented in Tables 4-13 and 4-14, respectively. The nearby
individuals are located 500 meters from the release
point for both facilities. The organs listed in Table 4-13 are
those where the dose is estimated to contribute 10 percent or
more of the total fatal cancer risk. For the reference dry
process facility, the maximum organ dose equivalents to the nearby
4-28
-------
Table 4-12. Atmospheric radioactive emissions assumed for reference
dry and wet process uranium conversion facilities.
Facility
Radionuclide
Emissions
(Ci/y)
Solubility Class (%)
D tf Y
Reference
Allied Corp,
Metropolis, IL
Sequoyah Fuels
Sequoyah, OK
U-Natural
Th-230
Ra-226
U-Natural
Th-230
Ra-226
0.10000
0.00050
0.00001
0.050
0.005
0.005
56
0
0
65
0
0
30
0
100
5
0
100
14
100
0
30
100
0
NRC84
NRC85b
(a) Solubility classes D, W, and Y refer to the retention of inhaled
radionuclides in the lungs; representative half-times for reten-
tion are less than 10 days for class D, 10-100 days for class W,
and greater than 100 days for class Y.
(b) Particle size 3.4 urn.
(c)Particle size (urn) X (Average: 1980-1984)
9.3
9.7
5.5
6.5
13.5
55.3
Data taken from NUREG-1157 (NRC85b).
4.
2.
1.
0.
0.
0.
2
1
3
69
39
00
to
to
to
to
to
to
10.
4.
2.
1.
0.
0.
2
2
1
3
69
39
Table 4-13.
Process
Radiation dose equivalent rates from atmospheric
radioactive emissions from reference uranium
conversion facilities.
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Dry
Wet
Lungs
Endosteum
Remainder
Lungs
Endosteum
Remainder
1.4E+1
8.3E+0
—
2.5E+1
1.4E+1
«•
2.1E+1
5.7E+1
4.9E+0
1.9E+1
3.3E+1
2 . OE+0
4-29
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Table 4-14. Fatal cancer risks due to atmospheric radioactive
emissions from reference uranium conversion
facilities.
Nearby Individuals Regional (0-80 km)
Lifetime Fatal Population
Process Cancer Risk Deaths/y
3E-5 8E-4
Wet 4E-5 6E-4
individuals are 14 mrem/y to the lungs and 8 mrem/y to the
endosteum. For the reference wet process facility, the maximum
organ dose equivalents are 25 mrem/y to the lungs and 13
mrem/y to the endosteum.
The estimated lifetime risk of fatal cancer to the nearby
individuals is estimated to be 3E-5 for the reference dry process
facility and 4E-5 for the reference wet process facility.
4.3.3.2 Doses and Risks to the Regional Population
Doses and fatal cancer risks to the regional population due
to atmospheric releases of radionuclides from uranium conversion
facilities are also summarized in Tables 4-13 and 4-14,
respectively. Here also, the organs listed in Table 4-13 are
those where the dose is estimated to contribute 10 percent or
more of the total fatal cancer risk. For the reference dry
process facility, maximum organ dose equivalents are 21 person-
rem/year to the lungs and 57 person-rem/year to the endosteum.
For the reference wet process facility, the maximum organ dose
equivalents are 19 person-rem/year to the lungs and 33 person-
rem/year to the endosteum.
The lifetime risks to the regional population are estimated
to be 8E-4 and 6E-4 fatal cancers per year of operation for the
reference dry and wet process facilities, respectively.
4.3.3.3 Projection of Fatal Cancers Per Year and the Risk
Distribution for the Uranium Conversion Segment of the
Uranium Fuel Cycle
Based on the results for the reference dry process and wet
process uranium conversion facilities, the total risk from all
uranium conversion facilities is estimated to be 1E-3 fatal
cancers per year of operation. This estimate is based on the
assumption of continuing operation of one dry process facility
and one wet process facility.
The estimated distribution of the estimated lifetime total
cancer risk projected for the uranium conversion segment of the
4-30
-------
uranium fuel cycle is presented in Table 4-15. This distribution
is based on the estimated 500,000 persons around the dry process
facility and 430,000 persons around the wet process facility.
Table 4-15. Estimated distribution of lifetime fatal cancer-
risks projected for uranium conversion facilities.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< l.OE-6
0
0
0
0
90
9,900
920,000
0
0
0
0
2E-5
3E-4
1E-3
Totals 930,000 1E-3
4.3.4 Supplementary Control Options and Costs
Well-proven particulate control technologies such as fabric
filters and scrubbers can be added to the existing control
systems at uranium hexafluoride conversion plants to reduce
emissions. The selection of additional controls must take
into account the presence of moisture and corrosive contaminants
(particularly fluorine) in some of the exhaust lines.
A previous study has estimated the cost of providing
additional fabric filters for both the wet and dry process plants
(TEK81). The estimated capital costs of the systems (in 1979
dollars) are approximately $2.1 million and $4.5 million for the
wet and dry plant, respectively. The total annual costs
(operating and maintenance) for the wet and dry process plants
are approximately $0.6 million and $1.3 million, respectively.
4.4 FUEL FABRICATION FACILITIES
4.4.1 General Description
Light water reactor (LWR) fuels are fabricated from uranium
that has been enriched in uranium-235 at a gaseous diffusion plant.
There natural uranium in the form of UFe has been processed to
increase the uranium-235 content from 0.7 percent up to 2 to 4
percent by weight. The enriched uranium hexafluoride product is
shipped to LWR fuel fabrication plants where it is converted to
solid uranium dioxide pellets and inserted into zirconium alloy
(Zircaloy) tubes. The tubes are fabricated into fuel assemblies
which are shipped to nuclear power plants.
4-31
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4.4.1.1 Fuel Fabrication Facilities in the United States
Table 4-16 presents a list of the seven licensed uranium
fuel fabrication facilities in the United States which fabricate
commercial LWR fuel. Of the seven, only five had active operating
licenses as of January 1, 1988. Of those five facilities, two
use enriched uranium hexafluoride to produce completed fuel
assemblies and two use uranium dioxide. The other facility
converts UF6 to UO2 and recovers uranium from scrap materials
generated in the various processes of the plant.
4.4.1.2 Process Description
The processing technology used for uranium fuel fabrication
consists of three basic operations: (1) chemical conversion of
UF6 to U02; (2) mechanical processing including pellet
production and fuel-element fabrication; and (3) recovery of
uranium from scrap and off-specification material. The most
significant potential environmental impacts result from convert-
ing UFg to UO2 and from the chemical operations involved in scrap
recovery.
4.4.1.2.1 Chemical Conversion of UFg to UO2
Two methods are currently used in UFs conversion and UO2
powder production: the ammonium diuranate (ADU) wet process and
the direct-conversion (DC) dry process.
The ADU process converts UF6 to (NH4)2U207 which is then
calcined to U02 powder. The UFg which is received from the
enrichment facility is vaporized and transferred to the reaction
vessels. The UF6 is hydrolyzed with water and neutralized with
NH4OH at a pH of 8 to 9 to form a slurry of ADU in an aqueous
solution of ammonium fluoride and ammonium hydroxide. The ADU is
recovered in a centrifuge and a clarifier and is subsequently
dried and calcined to form UO2 powder.
The DC process hydrolyzes the UF6 and reduces the uranium
directly to UO2. Cylinders of UF6 are placed in steam-heated
cabinets to vaporize the contained UF6. The UFg gas enters a
first reactor containing a bed of UO2F2 particles which is
fluidized by steam. The gas reacts with the steam on the hot, wet
surface of the particles to form a coating of UO2F2. The
reaction is:
UF6 + 2 H20 —> U02F2 + 4 HF
The particles of UO2F2, which are approximately 120 urn in
diameter, overflow to a product hopper. After the desired amount
is accumulated, the batch is transferred to the next vessel where
the bed is fluidized by steam and ammonia. Here it is reduced to
UO2. A high percentage of the UO2F2 is converted to UO2 in the
second reactor, but the product goes into a third reactor where,
by the same process, the reaction is carried to completion.
4-32
-------
Table 4-16. Light water reactor commercial fuel fabrication facilities licensed by the Nuclear
Licensee
Advanced
Nuclear
Fuels
Babcock &
Wilcox -
CNFP
Babcock &
». Vilcox
t
u>
CO
Combustion
Engineering
Combustion
Engineering
General
Electric
Facility
Location
Richland,
VA
Lynchburg ,
VA
Apollo,
FA
Windsor ,
CT
Hematite,
MO
Wilmington,
NC
Process Used
to convert
Operations UFg to U02
LEU(*) Conversion Dry & Wet
(UF6 to U02),
Fabrication & Scrap
Recovery; Commercial
LWR Fuel
LED Fabrication;
Commercial LWR Fuel
Authorized decontam- Wet
inat ion ; pending
Nuclear Reactor
Service Operations
LEU Fabrication;
Commercial LWR Fuel
LEU Conversion Dry
(UF6 to U02) &
Scrap Recovery
LEU Conversion Dry & Wet
(UF6 to U02) &
Final Product
Complete fuel
assemblies
Use U02 powder
to produce fuel
assemblies
U02 powder
Use UOo powder
to produce fuel
assemblies
U02 powder
Complete fuel
assemblies
1980
Operating
Capacity
(t/yr)
650
(250)
250
(150)
150
1,500
Active
Operating
License
as of
June 1987
No
Yes
No
Yes
Yes
Yes
Westinghouse Columbia,
Electric SC
Fabrication;
Commercial LWR Fuel
LEU Conversion
(UF6 to U02),
Fabrication & Scrap
Recovery; Commercial
LWR Fuel
Dry & Wet Complete fuel 750
assemblies
Total 3,300
Yes
(a)Low enrichment uranium.
-------
The gaseous effluent from each of the three converter
vessels (reactors) passes through a sintered nickel filter in the
top of each vessel before going to the gaseous effluent treatment
system where HF and particulates are removed from the off-gas
stream.
4.4.1.2.2 Mechanical Processing
Mechanical processing involves (l) pretreatment of UO2
powder by comminution, compaction, and granulation to the desired
size distribution; (2) pelletizing; (3) sintering the pellets
under a reducing atmosphere; (4) grinding to final dimensions;
(5) washing and drying the pellets; (6) loading the pellets into
Zircaloy tubes, fitting with end caps, and welding the end cap to
form fuel rods; and (7) assembling fuel rods to form finished
fuel elements.
4.4.1.2.3 Scrap Recovery Operations
A scrap recovery operation is important to the profitable
operation of a fuel fabrication plant. This system recycles the
scrap materials generated in the various processes of the plant
to recover the value of the scrap.
4.4.1.3 Existing Emission Controls
Emission control technology differs for ADU and DC
facilities. In either kind of facility, both process off-gases
and ventilation air are treated.
In the ADU facility, process gas passes through wet (water)
scrubbers (90 percent removal of entrained solids) and HEPA
filters before release to the atmosphere. Ventilation off-gases
go through roughing filters and HEPA filters before release to
the atmosphere.
In the DC facility, process gas passes through sintered
nickel filters, with trapped solids returned to the process; off-
gases continue to KOH scrubbers (for HF removal), then are
diluted for release to the atmosphere. Ventilation off-gases
pass through roughing filters and HEPA filters and are released.
4-4.2 Basis for the Dose and Risk Assessment of Fuel Fabrication
Facilities
4.4.2.1 Radionuclide Emissions
4.4.2.1.1 Operational Experience and Projected Future
Emissions
Table 4-17 presents reported uranium effluents from 1983
through 19'87 for each of the fuel fabrication facilities with
current operating licenses. The data in Table 4-17 show that the
Westinghouse and General Electric facilities have releases 10 to
4-34
-------
•t*.
I
u>
(JI
Table 4-17,
Licensee
Location
License No.
Docket No.
Light water reactor commercial fuel fabrication facilities reported annual
uranium effluent releases for 1983 through 1987 in /*Ci/y.
Year
U-234
U-235
U-236
U-238
Total
Babcock and Wilcox-CNFP
Lynchburg , VA
SNM-116
70-1201
Combustion Engineering
Windsor, CT
SNM-1067
70-1100
Combustion Engineering
Hematite, MO
SNM-33
70-36
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
4.7 E+0
5.6 E+O
4.6 E+O
5.7 E+O
3.9 E+O
NA(a)
NA
NA
NA
NA
NA
NA
NA
NA
NA
2.1 E-l
2.5 E-l
2.1 E-l
2.5 E-l
1.7 E-l
NA
NA
NA
NA
NA
NA
NA
NA
NA
NA
2.1 E-2
2.3 E-2
2.1 E-2
2.6 E-2
1.7 E-2
NA
NA
NA
NA
NA
NA
NA
NA
NA
NA
1.1 E+O
1.3 E+O
1.1 E+O
1.3 E+O
9.1 E-l
NA
NA
NA
NA
NA
NA
NA
NA
NA
NA
6.0 E+O
7.2 E+O
5.9 E+O
7.3 E+O
5.0 E+O
3.9 E+l
2.7 E+l
4.9 E+l
5.5 E+l
4.7 E+l
2.1 E+2
4.2 E+l
7.3 E+l
6.7 E+2
2.8 E+2
(a)Not available; only total curies of uranium released reported to the NRC,
-------
Table 4-17
Light water reactor commercial fuel fabrication facilities reported annual
uranium effluent releases for 1983 through 1987 in /*Ci/y (continued) .
Licensee
Location
License No.
Docket No.
Year
U-234
U-235
U-236
U-238
Total
General Electric
Wilmington, NC
SNM-1097
70-1113
Westinghouse Electric
Columbia, SC
SNM-1107
70-1151
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
3.1
4.0
4.1
1.2
1.6
1.2
1.5
1.2
1.1
1.0
E+2
E+2
E+2
E+3
E+2
E+3
E+3
E+3
E+3
E+3
2
2
2
7
1
5
1
7
5
5
.0
.6
.7
.1
.0
.3
.2
.2
.3
.6
E+l
E+l
E+l
E+l
E+l
E+l
E+2
E+l
E+l
E+l
4.5 E+2
5.7 E+0
5.7 E+0
1.6 E+l
2.0 E+0
NR(b)
NR
NR
NR
NR
1.3
1.7
1.5
3.5
5.6
2.5
3.2
3.1
3.4
3.1
E+2
E+2
E+2
E+2
E+l
E+2
E+2
E+2
E+2
E+2
4.6
6.0
5.9
1.6
2.3
1.5
1.9
1.6
1.5
1.4
E+2
E+2
E+2
E+3
E+2(a)
E+3
E+3
E+3
E-;-3
E+3
(a)Second half of 1987 is not available but is assumed to be same as first half.
(b)NR denotes not reported. Values are small and not included in total.
-------
100 times those of the Babcock and Wilcox and Combustion Engi-
neering facilities. This is expected because the Westinghouse
and General Electric plants start with uranium hexafluoride while
the other two facilities begin the fuel fabrication process with
U02.
The operating capacity of the existing commercial facilities
in 1980 was about 3,300 tons/year. If planned facility
expansions take place, the existing industry should be able to
meet demands as high as 4,600 tons/year in the immediate future.
Radionuclide emissions would be expected to remain proportional
to this production rate.
4.4.2.1.2 Source Term Used in the Assessment
The atmospheric radioactive emissions assumed to be released
each year by the reference fuel fabrication facility are
presented in Table 4-18. These values, with the exception of
uranium-236, represent the geometric mean of the reported
effluent releases for the Westinghouse fuel fabrication facility
for 1983 through 1987. The value for uranium-236 is based on
release data for 1983 through 1987 as reported in the semi-annual
environmental monitoring reports submitted to the NRC by the
General Electric facility at Wilmington, North Carolina.
Table 4-18. Atmospheric radioactive emissions assumptions for
reference fuel fabrication facility.
Radionuclide
U-234
U-235
U-236
U-238
Emissions
(Ci/y)
1.
6.
1.
3.
2
7
6
0
E-3
E-5
E-5
E-4
4.4.2.2 Site Characteristics Used in the Assessment
The Westinghouse plant at Columbia, South Carolina, was used
as the basis for the reference fuel fabrication facility. This
is appropriate since all phases of fuel fabrication (i.e., both
ADU and DC conversion of UFg to UC>2, mechanical fabrication of
fuel assemblies, and scrap recovery) take place at this site.
4-37
-------
The release point and climatological and demographic data
supplied to the AIRDOS/DARTAB/RADRISK computer codes are listed
in Appendix A. The climatological data are based on measurements
taken at the U.S. Weather Bureau Station at Columbia Metropolitan
Airport (NRC85a). Sets of hourly meteorological data obtained
from the airport for 1984 through 1986 were used to develop wind
frequency distributions for stability classes A through F. The
demographic data represent the 1986 population estimates within
80 kilometers of the Westinghouse plant.
The ingestion pathway food source data assume fractions
representative of an urban/low productivity site.
4.4.3. Results of the Dose and Risk Assessment for the Reference
Fuel Fabrication Facility
The estimated annual radiation dose equivalent and fatal
cancer risks from the reference facility are presented in Tables
4-19 and 4-20. The predominant exposure pathway is inhalation.
The annual radiation dose is primarily from uranium-234 and
uranium-238 (78 percent and 17 percent, respectively), for
both nearby individuals and the regional population.
4.4.3.1 Doses and Risks to the Nearby Individuals
Estimates of the annual dose equivalent and fatal cancer
risk to the nearby individuals due to the atmospheric emissions
of radionuclides from the reference fuel fabrication facility are
presented in Tables 4-19 and 4-20, respectively. The nearby
individuals are located 500 meters from the release point. Lung
is the only organ listed in Table 4-19, since it is the only
organ for which the dose is estimated to contribute 10 percent or
more of the total fatal cancer risk. The highest organ dose
equivalent to the nearby individual is 2.2 mrem/y, to the
lungs.
The lifetime risk of fatal cancer to nearby individuals from
the reference fuel fabrication facility is estimated to be 4E-6.
4.4.3.2 Doses and Risks to the Regional Population
Estimates of the annual dose equivalent and fatal cancer
risk to the regional population due to atmospheric emissions of
radionuclides from the reference fuel fabrication facility are
also presented in Tables 4-19 and 4-20. Here also, lung is the
only organ listed in Table 4-19, since it is the only organ for
which the dose is estimated to contribute 10 percent or more of
the total fatal cancer risk. The maximum organ annual dose
equivalent rate from the reference facility is 3.5 person-
rem/year, to the lungs.
4-38
-------
Table 4-19. Radiation dose equivalent rates from atmospheric
radioactive emissions from model fuel fabrication
facility.
Nearby Regional
Individuals Population
Organ (mrem/y) (person-rem/y)
Lungs 2.2E+0 3.5E+0
Table 4-20. Fatal cancer risks due to atmospheric radioactive
emissions from reference fuel fabrication facility.
Nearby Individuals Regional (0-80 km)
Lifetime Fatal Population
Cancer Risk Deaths/y
4E-6 8E-5
The incremental risk of fatal cancers in the regional
population is estimated to be 8E-5 per year of operation for the
reference facility.
4.4.3.3 Estimated Distribution of Lifetime Fatal Cancer Risks
Projected for Fuel Fabrication Facilities
Based on the evaluation of the reference fuel fabrication
facility, the total number of fatal cancers per year from all
fuel fabricators is estimated to be approximately 4E-4. This
estimate is based on the assumption of five operating fuel
fabrication facilities.
The estimated distribution of the lifetime fatal cancer risk
projected for all fuel fabricators is presented in Table 4-21.
This distribution was based on the assumption of 3,900,000
persons around five active fuel fabrication facilities. The
distribution does not account for any overlap in the populations
exposed to radionuclides released from multiple facilities.
4.4.4 Supplementary Control Options and Costs
Because the predicted dose equivalents and resultant health
risks to the nearby individuals and regional populations from
atmospheric emissions of radionuclides from the reference fuel
fabrication facility are low, no supplementary control options
are evaluated.
4-39
-------
Table 4-21. Estimated distribution of lifetime fatal cancer
risks projected for all fuel fabrication facilities.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< l.OE-6
0
0
0
0
0
50
3,900,000
0
0
0
0
0
3E-6
4E-4
Totals 3,900,000 4E-4
4.5 NUCLEAR POWER FACILITIES
4.5.1 General Description
4.5.1.1 Nuclear Power Generation in the United States
As of December 1986, there were 100 operable nuclear power
reactors in the United States, with a total generating capacity
of 85,177 MWe. With only one exception (a high-temperature gas-
cooled reactor), all of these nuclear power reactors are either
boiling water reactors (BWR) or pressurized water reactors (PWR).
Pressurized water reactors comprise approximately two-thirds of
the light-water generating capacity. It is assumed this two-to-
one PWR-BWR ratio will continue through the year 2000.
Table 4-22 presents a list of the commercial nuclear power
reactors in the United States (DOE87). A recent update of
nuclear power in the United States provided in Nuclear News
(2/88) indicated 102 operable commercial nuclear power reactors.
4.5.1.2 Process Description
A light-water-cooled nuclear power station generates
electricity using the same basic principles as a conventional
fossil-fueled (oil or coal) power station except that the source
of heat used to produce steam is provided by nuclear fission
instead of combustion.
In a boiling water reactor, the coolant boils as it passes
through the reactor. The resulting steam is passed through a
turbine and a condenser. The condensed steam is then pumped back
into the reactor. The energy removed from the steam by the
turbine is transformed into electricity by a generator.
The process is the same in a pressurized water reactor
except that the reactor coolant water is pressurized to prevent
4-40
-------
Table 4-22.
State/Site
U.S. nuclear power generating units operable as of
December 31, 1986 (DOE87).
Utility
Unit Name
Type
Alabama
Decatur
Decatur
Decatur
Decatur
Decatur
Arizona
Wintersburg
Wintersburg
Arkansas
Runnellville
Runnellville
California
Avila Beach
Avila Beach
Clay Station
San Clemente
San Clemente
San Clemente
Colorado
Platteville
Connecticut
Haddam Neck
Waterford
Waterford
Waterford
Florida
Florida City
Florida City
Ft. Pierce
Ft. Pierce
Red Level
Georgia
Baxley
Baxley
Tennessee Valley
Authority
Tennessee Valley
Authority
Tennessee Valley
Authority
Alabama Power
Alabama Power
Arizona Public Service
Arizona Public Service
Arkansas P & L
Arkansas P & L
Pacific Gas & Electric
Pacific Gas & Electric
Sacramento Municipal
Utility District
Southern Calif. Edison
Southern Calif. Edison
Southern Calif. Edison
Public Service Co.
of Colorado
Connecticut Yankee
Atomic Power
Northeast Utilities
Northeast Utilities
Northeast Utilities
Florida P & L
Florida P & L
Florida P & L
Florida P & L
Florida Power Corp.
Georgia Power
Georgia Power
Browns Ferry 1
Browns Ferry 2
Browns Ferry 3
Joseph M. Farley 1
Joseph M. Farley 2
Palo Verde 1
Palo Verde 2
Arkansas Nuclear 1
Arkansas Nuclear 2
Diablo Canyon 1
Diablo Canyon 2
Rancho Seco
San Onofre 1
San Onofre 2
San Onofre 3
Fort St. Vrain
Haddam Neck
(Connecticut Yankee)
Millstone 1
Millstone 2
Millstone 3
Turkey Point 3
Turkey Point 4
St. Lucie 1
St. Lucie 2
Crystal River 3
Hatch 1
Hatch 2
BWR
BWR
BWR
PWR
BWR
PWR
PWR
PWR
PWR
PWR
PWR
PWR
PWR
PWR
PWR
HTGR
PWR
BWR
PWR
PWR
PWR
PWR
PWR
PWR
PWR
BWR
BWR
4-41
-------
Table 4-22.
State/Site
U.S. nuclear power generating units operable as of
December 31, 1986 (continued) (DOE87).
Utility
Unit Name
Type
Illinois
Byron
Cordova
Cordova
Morris
Morris
Seneca
Seneca
Zion
Zion
Iowa
Palo
Kansas
Burlington
Louisiana
St Francisville
Taft
Maine
Vicasset
Maryland
Lusby
Lusby
Massachusetts
Plymouth
Rowe
Michigan
Bridgman
Bridgman
Charlevoix
Newport
South Haven
Minnesota
Monticello
Red Wing
Red Wing
Mississippi
Port Gibson
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Iowa Electric L & P
Kansas City P & L
Gulf State Utilities
Louisiana P & L
& Kansas G & E
Maine Yankee Atomic
Power
Baltimore G & E
Baltimore G & E
Boston Edison
Yankee Atomic Electric
Indiana & Michigan Elec .
Indiana & Michigan Elec.
Consumers Power
Detroit Edison
Consumers Power
Northern States Power
Northern States Power
Northern States Power
Mississippi P & L
Byron 1
Quad-Cities 1
Quad-Cities 2
Dresden 2
Dresden 3
LaSalle 1
LaSalle 2
Zion 1
Zion 2
Duane Arnold
Wolf Creek
River Bend 1
Waterford 3
Maine Yankee
Calvert Cliffs 1
Calvert Cliffs 2
Pilgrim 1
Yankee Rowe 1
Donald C . Cook 1
Donald C. Cook 2
Big Rock Point
Fermi 2
Palisades
Monticello
Prairie Island 1
Prairie Island 2
Grand Gulf 1
FWR
BVR
BVR
BVR
BVR
BVR
BVR
PVR
PVR
BVR
PVR
BVR
PVR
PVR
PVR
PVR
BVR
PVR
PVR
PVR
BVR
BVR
PVR
BVR
PVR
PVR
BVR
4-42
-------
Table 4-22,
State/Site
U.S. nuclear power generating units operable as of
December 31, 1986 (continued) (DOE87).
Utility
Unit Name
Type
Missouri
Felton
Nebraska
Brownsville
Fort Calhoun
New Jersey
Forked River
Salem
Salem
Salem
New York
Buchanan
Buchanan
Rochester
Oswego
Scriba
North Carolina
Coweas Ford Dam
Coweas Ford Dam
Southport
Southport
Ohio
Oak Harbor
North Perry
Oregon
Prescott
Pennsylvania
Berwick
Berwick
Middletown
Lancaster
Lancaster
Pottstown
Shippingport
Union Electric
Nebraska Public Power
Omaha Public Power Dist.
Jersey Central P & L
Public Service E & G
& Philadelphia Electric
Public Service E & G
& Philadelphia Electric
Public Service E & G
Consolidated Edison
Power Authority of
the State of New York
Rochester Gas & Elec.
Niagara Mohawk Power
Power Authority of
the State of New York
Duke Power
Duke Power
Carolina P & L
Carolina P & L
Cleveland Elec. Ilium.
Cleveland Elec. Ilium.
Portland General Elec.
Pennsylvania P & E
Pennsylvania P & E
Metropolitan Edison
Philadelphia Electric
& Public Service E & G
Philadelphia Electric
& Public Service E & G
Philadelphia Electric
Duquesne Light
Callaway 1
Cooper
Fort Calhoun 1
Oyster Creek 1
Salem 1
Salem 2
Hope Creek 1
Indian Point 2
Indian Point 3
Robert E. Ginna
Nine Mile Point 1
James A. Fitzpatrick
McGuire 1
McGuire 2
Brunswick 1
Brunswick 2
Davis -Besse 1
Perry 1
Trojan
Susquehanna 1
Susquehanna 2
Three Mile Island 1
Peach Bottom 2
Peach Bottom 3
Limerick 1
Beaver Valley 1
PWR
BVR
PVR
BVR
PVR
PVR
BVR
PVR
PVR
PVR
BVR
BVR
PVR
PVR
BVR
BVR
PVR
BVR
PVR
PVR
PVR
PVR
BVR
BVR
BVR
PVR
4-43
-------
Table 4-22.
State/Site
U.S. nuclear power generating units operable as of
December 31, 1986 (continued) (DOE87).
Utility
Unit Name
Type
South Carolina
Clover
Clover
Hartsvllle
Jenkinsville
Seneca
Seneca
Seneca
Tennessee
Daisy
Daisy
Vermont
Vernon
Virginia
Surry
Surry
Mineral
Mineral
Washington
Richland
Wisconsin
Carlton
Genoa
Two Creeks
Two Creeks
North Carolina Electric
Membership Corp.
North Carolina
Municipal Power
Carolina F & L
South Carolina E & G
Duke Power
Duke Power
Duke Power
Tennessee Valley
Authority
Tennessee Valley
Authority
Vermont Yankee Nuclear
Power
Virginia Power Co.
Virginia Power Co.
Virginia Power Co.
Virginia Power Co.
Washington Public Power
Supply System
Wisconsin Public Service
Dairy Land Power Corp.
Wisconsin Elec. Power
Wisconsin Elec. Power
Catawba 1
Catawba 3
H. B. Robinson 2
Summer 1
Oconee 1
Oconee 2
Oconee 3
Sequoyah 1
Sequoyah 2
Vermont Yankee
Surry 1
Surry 2
North Anna 1
North Anna 2
WNP 2
Kewaunee
La Cross e
Point Beach 1
Point Beach 2
PWR
PWR
PWR
PWR
PWR
PWR
FWR
PWR
PWR
BWR
PWR
PWR
PWR
PWR
BWR
PWR
BWR
FWR
PWR
4-44
-------
boiling. Energy is transferred through a heat exchanger (steam
generator) to a secondary system where the water does boil.
Reactor coolant water is kept at high pressures by maintaining a
closed system and electrically heating water in a tank called the
pressurizer. After passage through the steam generator, the
water is returned to the reactor. Secondary steam turns the
turbine, is cooled in the condenser, and is pumped back into the
steam generator.
During the fission process, radioactive fission products are
produced and accumulate within the nuclear fuel. In addition,
neutrons produced during fission interact within the fuel and
coolant to produce radioactive activation products. A reactor
may experience periodic fuel failure or defects which result in
the leakage of some of the fission and activation products out of
the fuel and into the coolant. Accordingly, a typical light
water reactor will experience build-up of radioactive fission and
activation products within the coolant. For both PWRs and BWRs,
the radioactive contaminants that accumulate within the coolant
are the source of radioactive emissions from the facility.
4.5.1.2.1 Boiling Water Reactors
For BWRs, the primary sources of routine gaseous emissions
are from the off-gas treatment system and the building
ventilation system exhaust.
The off-gas treatment system collects noncondensable gases
and vapors which are exhausted at the condenser via the mechani-
cal vacuum pump and air ejectors. The off-gases are processed
through a series of delay systems and filters to remove airborne
radioactive particulates and halogens and delay the release of
gases, thereby allowing only small quantities of the longer-lived
radioactive noble gases to be released.
Building ventilation systems are also a source of airborne
radioactive emissions from BWRs. Airborne releases from the
reactor building are due to primary coolant leakage. Releases
from the turbine building are due to steam leakage. Releases
from the auxiliary building are due to leakage from the liquid
waste treatment system. Releases from the fuel handling
facilities are associated with evaporation from the fuel pool.
4.5.1.2.2 Pressurized Water Reactors
In PWRs, there are four primary sources of radioactive
emissions:
1. Discharges from the gaseous waste management system;
2. Discharges associated with the exhaust of noncon-
densable gases at the main condenser;
4-45
-------
3. Discharges from the steam generator to blowdown
exhaust; and
4. Radioactive gaseous discharges from the building
ventilation exhaust, including the reactor building,
reactor auxiliary building, fuel handling building,
and turbine building.
The exhaust may pass through separate or combined exhaust
points and typically passes through high efficiency particulate
air (HEPA) filters and charcoal filters prior to discharge.
The gaseous waste management system collects fission
products, mainly noble gases that accumulate in the primary
coolant. A small portion of the primary coolant flow is
continually diverted to the primary coolant purification, volume,
and chemical control system to remove contaminants and adjust the
chemistry and volume. During this process, noncondensable gases
are stripped and routed to the gaseous waste management system
which typically consists of a series of gas storage tanks where
they are held long enough to allow short-lived radioactive gases
to decay, thereby leaving relatively small quantities of longer-
lived radionuclides to be released to the atmosphere.
The second source of radioactive emissions is at the main
condenser, where noncondensable gases are stripped from the
secondary system and exhausted to enhance the efficiency of
energy conversion. The noncondensable gases may include small
quantities of fission and activation products which can enter the
secondary coolant system via primary coolant to secondary coolant
leakage at the steam generators.
A third possible source of radioactive emissions is the
exhaust of noncondensed vapors and gases associated with steam
generator blowdown. A portion of the reservoir of secondary side
water in the steam generators is routinely let down to the steam
generator blowdown treatment system to help maintain the chemical
purity of the secondary side coolant, thereby helping to reduce
secondary side corrosion. Some treatment processes result in the
generation of water vapor and noncondensable gases which, follow-
ing filtration, are discharged to the environment.
The last category of radioactive emissions is the exhaust of
airborne radioactive materials via the building ventilation
exhaust. Leakage of primary and secondary coolant, steam leak-
age, evaporation from the fuel pool, and leakage from various
liquid processing systems result in the accumulation of airborne
radionuclides which are discharged via the building ventilation
system exhaust.
4.5.1.3 Existing Emission Controls
A number of effluent and process controls are employed at an
LWR to reduce radionuclide emissions to the atmosphere. Some of
4-46
-------
the controls operate directly on the emissions prior to release,
while the others indirectly reduce emissions by limiting the
amount of radioactive materials that leak from process systems.
4.5.1.3.1 BWR Emission Controls
HEPA and charcoal filters are routinely used to remove
particulate and radioiodine emissions from the various building
ventilation exhausts. In addition, all BWRs employ a main
condenser off-gas treatment system to filter and hold up airborne
radionuclides vented by the mechanical vacuum pumps and the air
ejection system. The off-gas treatment system typically consists
of a delay line followed by cryogenically cooled charcoal delay
systems. These systems increase the holdup times for noble gases.
Other indirect methods are also used to help reduce atmos-
pheric emissions. Some of these systems include the following
techniques:
1. Venting the gaseous emissions from the
[mechanical] vacuum pump to the condenser
virtually eliminates this source of radioiodine
emission;
2. The steam generator blowdown flash tank is vented
to the condenser or the blowdown is cooled,
thereby precluding a vapor flash; and
3. Special provisions are taken to control steam
leakage from steam line valves.
BWRs also employ turbine gland sealing systems which help to
reduce the steam leakage from the turbine.
4.5.1.3.2 PWR Emission Controls
For PWRs, controls applied at the point of release include
HEPA and charcoal filtration units. The HEPA filters are de-
signed and tested to ensure 99.97 percent efficiency for
particulate emissions. Charcoal filter efficiency for
radioiodines varies depending on the depth of the charcoal
filters, whether provisions exist to control the relative
humidity of the discharge air, and numerous other factors.
Efficiency for iodine removal on charcoal adsorbers ranges from a
decontamination factor (ratio of the amount of radioactive
material initially present to the amount remaining after
processing) of 10 to 1,000, the typical value being 100 (Mo84).
In addition to filtration systems, PWRs employ gas decay
tanks to collect and store noble gases which are stripped from
the primary coolant via the chemical and volume control system.
The holdup time provided by the gas decay tanks depends on the
number and volume of each tank and the storage pressure. Typi-
cally, storage times are on the order of 60 to 90 days, which
results in the decay of all but the long-lived noble gases.
4-47
-------
Delay systems based on charcoal adsorption are also used,
but to a lesser degree. In addition, some delay systems use a
nitrogen cover gas which is continuously recycled. This results
in virtually unlimited holdup of gaseous radionuclides that
enter the system.
PWRs also employ internal containment cleanup systems which
recycle the containment atmosphere and remove airborne
particulates and radioiodines prior to venting the gas.
Other indirect methods are also used to help reduce atmos-
pheric emissions. These systems include the three techniques
described for BWR emissions (Section 4.5.1.3.1).
4.5.2 Basis for the Dose and Risk Assessment of Power Reactor
Facilities
4.5.2.1 Radionuclide Emissions
4.5.2.1.1 Operational Experience and Projected Future
Emissions
Tables 4-23 and 4-24 present the geometric mean and standard
deviation for releases of selected radionuclides during 1981
through 1985 for BWRs and PWRs respectively. For BWRs, the
annual emissions for each radionuclide have been decreasing with
time. The emission rate for PWRs has remained stable for
tritium, iodine-131, and xenon-133 and has decreased for
cesium-137.
The future of nuclear power in the United States is uncer-
tain. The principal factors affecting the longer term future of
nuclear power are the demand for electricity, interest rates, the
price of oil, public attitude, and the regulatory climate. The
probable range of nuclear capacity by the year 2000 is projected
to be from 100 to 110 plants.
4.5.2.1.2 Source Terms Used in the Assessment
Tables 4-25 and 4-26 present the source terms assumed for
the model BWRs and PWRs respectively. These source terms are
based on the respective geometric means for concentrations of
tritium, iodine-131, krypton-85m, krypton-85, krypton-87,
krypton-88, xenon-131m, xenon-133, xenon-135m, xenon-135,
xenon-138, and cesium-137 in reported airborne releases for 1985.
These radionuclides were chosen since they contribute the
majority of the dose. The source terms for the remaining radio-
nuclides were calculated based on their ratio to either
iodine-131, xenon-133, or cesium-137 as obtained by examining
these ratios for nuclear power plants that have release rates
close to the geometric mean values.
4-48
-------
Table 4-23. Geometric mean and standard deviation by year for selected radionuclides for
boiling water reactors in the United States for 1981 through 1985 in /iCi/y.
vo
Year
1981
1982
1983
1984
1985
Year
1981
1982
1983
1984
1985
H-3
Geometric
Mean Std. Dev.
18
19
14
13
12
3.7
3.1
5.3
2.5
3.8
Kr-87
Geometric
Mean Std. Dev.
451
265
240
182
57
11
30
60
25
15
1-131
Geometric
Mean
4.0E-2
4.0E-2
3.4E-2
1.3E-2
1.1E-2
Std. Dev.
7.5
8.2
9.4
12
6.8
Kr-88
Mean
353
410
195
203
51
Geometric
Mean
661
502
461
453
77
Std. Dev.
12
28
74
20
15
Mean
28
3.
51
127
29
Kr-85m
Geometric
Std. Dev.
9.0
10
65
22
8.3
Xe-131m
Geometric
Std. Dev.
41
2 61
24
20
23
Kr-85
Geometric
Mean Std. Dev.
6.8 36
3.0 150
13.0 112
14.0 14
2.9 39
Xe-133m
Geometric
Mean Std. Dev.
69 7.3
46 4.3
34 79
34 18
30 12
-------
Table 4-23. Geometric mean and standard deviation by year for selected radionuclides for
boiling water reactors in the United States for 1981 through 1985 in
(continued).
Xe-133
Xe-135m
Xe-135
Geometric
Year
1981
1982
1983
* 1984
g 1985
Mean
1180
1980
1390
1400
633
Std. Dev.
15
8.2
29
14
14
Xe-l38
Geometric
Year
1981
1982
1983
1984
1985
Mean
1330
1320
825
195
70
Std. Dev.
12
9.5
13
22
150
Geometric Geometric
Mean Std. Dev. Mean Std. Dev.
421 10 711 18
502 8.4 1650 7.0
417 12 1250 6.4
122 12 617 16
57 22 377 8.5
Cs-137
Geometric
Mean Std. Dev.
9.8E-4 7.1
8.0E-4 4.7
4.6E-4 4.9
3.5E-4 9.5
1.6E-4 15
-------
Table 4-24. Geometric mean and standard deviation by year for selected radlonuclides for
pressurized water reactors In the United States for 1981 through 1985 In ^Cl/y.
H-3
1-131
Kr-85m
Kr-85
I
01
Year
1981
1982
1983
1984
1985
Year
1981
1982
1983
1984
1985
Geometric
Mean Std. Dev.
11 7.1
13 6.0
22 7.4
24 6.9
15 5.0
Kr-87
Geometric
Mean Std. Dev.
5.2E-1 46
6.6E-1 48
7.0E-1 39
2.2E-1 34
1.8E-1 31
Geometric
Mean
5.7E-3
4.5E-3
5.6E-3
5.7E-3
3.1E-3
Std. Dev.
9.6
11
10
11
7.2
Kr-88
Geometric
Mean
1.2
2.7
1.3
1.2
0.6
Geometric
Mean
4.9E-1
7.1E-1
6.2E-1
8.0E-1
5.7E-1
Std. Dev.
54
31
54
16
20
Std. Dev.
17
15
33
10
26
Xe-131m
Geometric
Mean Std. Dev.
6.6
5.4
5.1
1.4
2.3
12
6.6
18
392
30
Geometric
Mean
6.0
10
23
6.6
5.6
Std. Dev.
14
7.9
310
11
13
Xe-133m
Geometric
Mean
4.7
8.1
4.4
6.8
4.7
Std. Dev.
7.7
8.2
11
12
15
-------
Table 4-24. Geometric mean and standard deviation by year for selected radionuclides for
pressurized water reactors in the United States for 1981 through 1985
in pCi/y (continued).
Xe-133
Xe-135m
Xe-135
-------
Table 4-25.
Atmospheric radioactive emissions assumed for model
boiling water reactor.
Annual
Emissions Reference
Radionuclide (/*Ci/y) Radionuclide
Ratio
Reference
Plant
H-3
1-131
1-132
1-133
1-134
1-135
1.2E+1
1.1E-2
2.9E-2
7.5E-2
2.1E-2
2.0E-1
H-3
1-131
1-131
1-131
1-131
1.00
2.71
7.02
1.96
18.5
LaSalle 1 & 2
LaSalle 1 & 2
LaSalle 1 & 2
LaSalle 1 fie 2
LaSalle 1 & 2
Kr-85m
Kr-85
Kr.-87
Kr-88
Xe-131m
Xe-133m
Xe-133
Xe-135m
Xe-135
Xe-138
N-13
Ar-41
Cr-51
Mn-54
Co-58
Co-60
Zn-65
Sr-89
Sr-90
Nb-95
Zr-95
Cs-137
Ba-140
La-140
5.1E+1
2.9E+0
5.7E+1
7.7E+1
2.9E+1
2.9E+1
6.3E+2
5.7E+1
3.8E+2
7.0E+1
7.2E+0
4.3E+1
1.6E-3
2.2E-4
1.1E-4
1.8E-3
1.2E-4
6.9E-3
3.1E-4
3.8E-6
3.8E-6
1.6E-4
9.8E-3
9.8E-3
Kr-85m"
Kr-85*
Kr-87*
Kr-88*
Xe-131m'
Xe-lSSm"
Xe-133*
Xe-1351
Xe-138*
Cs-137
Cs-137
Cs-137
Cs-137
Cs-137
Cs-137
Cs-137
CS-137
CS-137
Cs-137
Cs-137
Cs-137'
Cs-137
Cs-137
4.66E+4
2.80E+5
10.0
1.39
0.72
11.90
0.75
44.30
2.02
2.43E-2
2.43E-2
1.00
63.20
63.20
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
J.A
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
Fitzpatrick
J.A. Fitzpatrick
*Geometric mean calculated from 1985 reported atmospheric
radioactive emissions for U.S. boiling water reactors.
4-53
-------
Table 4-26.
Atmospheric radioactive emissions assumed for model
pressurized water reactor.
Annual
Emissions Reference
Radionuclide (/iCi/y) Radionuclide
Ratio
Reference
Plant
H-3
1.5E+1
H-3:
1-131
1-132
1-133
1-135
3.1E-3
1.8E-6
2.5E-4
9.2E-7
I-1311
1-131
1-131
1-131
1.00
8.70E-4
0.04
3.00E-4
Arkansas One 1
Arkansas One 1
Arkansas One 1
Arkansas One 2
Kr-85m
Kr-85
Kr-87
Kr-88
Xe-131m
Xe-133m
Xe-133
Xe-135m
Xe-135
Xe-138
6.4E-1
5.6E+0
1.8E-1
5.7E-1
2.3E+0
4.7E+0
l.OE+3
5.9E-1
3.5E+1
8.0E-1
Kr-85m"
Kr-85*
Kr-87*
Kr-88*
Xe-131m*
Xe-133m*
Xe-133*
Xe-135m*
Xe-135*
Xe-138*
Ar-41
Mn-54
Co-58
Fe-59
Co-60
Zn-65
Sr-89
Sr-90
Cs-137
Ba-140
La-140
5.9E-2
1.2E-4
2.3E-6
3.4E-4
7.6E-5
2.2E-4
1.5E-2
2.7E-2
5.7E-5
2.3E-4
1.2E-4
Cs-137
Cs-137
Cs-137
Cs-137
Cs-137
CS-137
Cs-137
Cs-137
CS-137'
Cs-137
Cs-137
l.OOE+3
2.07
0.04
6.01
1.33
3.92
2.61E+2
4.81E+2
1.00
4.11
2.10
Crystal
Crystal
Crystal
Crystal
Crystal
Crystal
Crystal
Crystal
Crystal
Turkey
Turkey
River
River
River
River
River
River
River
River
River
Point 3
Point 3
*Geometric mean calculated from 1985 reported atmospheric
radioactive emissions for U.S. pressurized water reactors.
4-54
-------
4.5.2.2 Other Parameters Used in the Assessment
Sets of joint frequency data from on-site meteorological
stations for a representative group of U.S. nuclear power plant
sites were obtained and compared. The meteorological data for
Limerick were used for the assessment.
A review of the population distribution in the vicinity of
nuclear power plants reveals a wide variation in average popula-
tion densities. The data for 91 plants show the population
density varies between 19 to 2,099 persons per square mile
(NRC82). Table 4-27 presents the minimum, maximum, and 90th
percentile.
Table 4-27,
Distance
(miles)
Minimum, maximum, median, and 90th percentile
population densities for nuclear power reactor
sites in the United States.
Persons/Square Mile
Minimum
Maximum
Median
90%
0-5
5-10
10-20
20-30
30-50
0
2
0
2
0
790
700
730
2,000
2,500
40
80
90
110
110
190
260
380
490
660
Source: NRC82
Limerick, with a density of about 900 persons per square
mile, was selected as the reference site. The population
distribution used in the assessment was generated using the
SECPOP code. To assess the potential risk to nearby individuals,
doses and risks were evaluated at 750 m in the predominant wind
direction.
Food fractions representative of a rural location were used
in assessing both the model BWR and PWR. Details of the inputs
to the assessment code are given in Appendix A.
4-55
-------
4.5.3 Results of the Dose and Risk Assessment of Power Reactor
Facilities
4.5.3.1 Results for Model Power Reactor Facilities
The estimated annual radiation dose and fatal cancer risks
from the model BWR and PWR facilities are presented in Tables
4-28 and 4-29.
Table 4-28. Dose rates from model light water reactors.
Facility
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Model BWR
Model PWR
Gonads
Breast
Red Bone Marrow
Lungs
Remainder
Red Bone Marrow
Breast
Gonads
Endosteum
Remainder
2.5E-1
2.4E-1
1.9E-1
1.9E-1
1.9E-1
3.2E-1
1.1E-1
9.5E-2
6.8E-1
7.3E-2
4.9E+0
4.8E+0
3.7E+0
3.8E+0
3.7E+0
6.0E+0
1.8E+0
1.5E+0
1.3E+1
1.2E+0
Table 4-29. Fatal cancer risks for model light water reactors.
Source
Nearby Individuals
Lifetime Fatal
Cancer Risk
Regional (0-80 km)
Population
Deaths/y
Model BWR
Model PWR
5E-6
3E-6
1E-3
7E-4
4.5.3.1.1 Doses and Risks to the Nearby Individuals
Estimates of the annual dose and fatal cancer risk to the
nearby individuals due to atmospheric emissions of radionuclides
from the model BWR are presented in Tables 4-28 and 4-29,
respectively. The organ receiving the maximum dose is the
thyroid, but this contributes less than 10 percent of the risk.
All organ doses are predicted to be below 1 mrem/y. The
predominant exposure pathway for the model BWR is air immersion.
4-56
-------
Approximately 32 percent of the dose results from exposure to
kryptons, 30 percent from exposure to xenons, and 10 percent
from exposure to argon-41. The lifetime risk of fatal cancer due
to the estimated radionuclide exposures from the model BWR is
5E-6.
Estimates of the annual dose and fatal cancer risk to nearby
individuals due to atmospheric emissions from the model PWR are
also summarized in Tables 4-28 and 4-29. The organs receiving
the maximum dose are red bone marrow and the breast. All organ
doses are below 1 mrem/y. The predominant exposure pathways
are air immersion and inhalation. Xenon isotopes contribute
74 percent of the dose, and strontium-90 contributes 14 percent.
The lifetime risk of fatal cancer due to the estimated
radionuclide exposures from the model PWR is 3E-6.
4.5.3.1.2 Doses and Risks to the Regional Population
Estimates of the collective dose rate and fatal cancer risk
to the regional population due to atmospheric releases of
radionuclides from the model BWR are presented in Tables 4-28 and
4-29, respectively.
All organ doses are predicted to be below 6 person-rem/year.
The most important population pathway for the model BWR is air
immersion, with some contribution from exposure to ground
surface. The most important nuclides are the xenons (39 percent)
and the kryptons (32 percent). The incremental risk to the
regional population is estimated to be 1E-3 fatal cancers per
year of operation.
For the model PWR, the estimates of collective dose and
fatal cancer risks to the regional population are also summarized
in Tables 4-28 and 4-29. All organ doses are estimated to be
less than 1 person-rem/year. Air immersion is the most important
population pathway for the model PWR, with contributions from
ingestion and inhalation. The most important nuclides are
xenon-133 (64 percent) and strontium-90 (26 percent). The
incremental risk to the regional population is 1E-4 fatal cancers
per year of operation.
4.5.3.2 Projection of Fatal Cancers per Year and the Risk
Distribution for the Power Reactor Segment of the
Uranium Fuel Cycle
Based on the results of the calculations of the model BWR
and PWR facilities, the total risk from all power reactors in the
United States is estimated to be 9E-2 fatal cancers per year.
This estimate is based on the assumption of 63 PWRs and 37 BWRs.
The estimated distribution of the lifetime fatal cancer risk
projected for all power reactors is presented in Table 4-30.
4-57
-------
The distribution does not account for overlap in the
populations exposed to radionuclides released from more than a
single reactor and may understate the risk to some individuals
residing near multiple reactors.
Table 4-30. Estimated distribution of lifetime fatal cancer
risks projected for all power reactors.
Risk Interval Number of Persons Deaths/y
1E-01 to 1E+00 0 0
1E-02 to 1E-01 0 0
1E-03 to 1E-02 0 0
1E-04 to 1E-03 0 0
IE-OS to 1E-04 0 0
1E-06 to 1E-05 * *
< l.OE-06 240,000,000 9E-2
Totals 240,000,000 9E-2
*The results of the assessments of the model facilities indicate
that there might be persons in this risk interval, but without
site-specific assessments, the EPA cannot quantify the number.
4.5.3.3 Doses Reported by Power Reactor Operators
Power reactor operators are required to calculate and report
the estimated doses to the "maximally exposed individual" residing
near the site. Table 4-31 presents the exposures reported by
operators to the NRC in recent years. Since the operators do not
use a consistent methodology in making their estimates, the last
column of Table 4-31 provides an estimate of the doses in terms
of the ICRP's effective dose equivalent. Five reactors have reported
doses of 1 mrem/y or greater during the period examined (1984-
1987). The highest estimated doses are below 5 mrem/y,
consistent with the ALARA objectives of 10 CFR 50, Appendix I.
4.5.4 Supplementary Control Options and Costs
Emissions from the light-water reactor segment of the
uranium fuel cycle do not result in doses or risks high enough to
warrant a full evaluation of supplementary control options and
costs. The well-proven control technologies such as additional
decay tanks for noble gases and additional charcoal adsorbers for
radioiodines can be employed. Costs for such systems can be
developed only on a reactor-specific basis due to the unique
designs of these facilities. A rough figure of $5 million per
reactor can be estimated.
4-58
-------
Table 4-31. Doses to maximally exposed individuals in mrem/y.
I
ui
vo
Facility
VOGTLE 1
OYSTER CREEK
CATAVBA
HADDAM NECK
MCGUIRE 1
VATERFORD
COOPER
LA CROSSE
Docfcet Year
50-424 1987
1988
50-219 1986
1985
1987
50-413 1986
1985
1987
50-213 1984
1985
1987
1986
50-369 1985
1986
1987
50-382 1987
1986
1985
50-298 1985
1986
1987
50-409 1986
1987
Whole
Body
2
1
4
1
1
2
8
8
1
1
6
3
1
1
0
6
0
0
5
4
1
4
2
.8E+0
.8E+0
.3E+0
.4E+0
.7E-1
.2E+0
.8E-1
.9E-1
.5E+0
.OE+0
.6E-1
.9E-1
.8E+0
.5E-1
.OE+0
.6E-1
.OE+0
.OE+0
.7E-1
.OE-1
.8E-2
.7E-1
.OE-1
Thyroid
9
2
8
8
1
0
0
6
2
1
7
8
2
0
8
1
5
3
6
5
9
0
0
.9E-3
.5E-3
.1E-1
.8E+0
.7E-1
.OE+0
.OE+0
.7E-1
.8E-1
.4E-1
.3E-2
.7E-2
.6E+0
.OE+0
.1E-2
.4E+0
.5E+0
.1E+0
.OE-1
.6E-1
.7E-2
.OE+0
.OE+0
Bone
5.0E-7
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
5.6E-3
O.OE+0
O.OE+0
6.4E-1
4.3E-1
2.9E-2
O.OE+0
O.OE+0
Liver
9.8E-3
2.5E-3
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
6.7E-1
O.OE+0
O.OE+0
5.6E-1
3.9E-1
1.8E-2
O.OE+0
O.OE+0
Lung
9.9E-3
2.5E-3
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
6.6E-1
O.OE+0
O.OE+0
5.6E-1
3.9E-1
1.8E-2
O.OE+0
O.OE+0
Skin
O.OE+0
O.OE+0
4 . 5E+0
1 . 5E+0
1.7E-1
0 . OE+0
O.OE+0
2 . 5E+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
4.1E-1
2. OE-1
O.OE+0
O.OE+0
O.OE+0
9.4E-1
7.4E-1
4.0E-2
0 . OE+0
O.OE+0
GI-
Tract
9.8E-3
2.5E-3
0 . OE+0
O.OE+0
O.OE+0
3 . 3E+0
2 . 2E+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
6.6E-1
O.OE+0
O.OE+0
5.5E-1
3.9E-1
1.9E-2
O.OE+0
O.OE+0
Kidney
9.8E-3
2.5E-3
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
6.6E-1
O.OE+0
O.OE+0
5.6E-1
3.9E-1
1.9E-2
O.OE+0
O.OE+0
-------
Table 4-31. Doses to maximally exposed individuals in mrem/y (continued).
Facility
PALO VERDE
PILGRIM
Docket
50-528
50-293
Year
1987
1988
1985
1985
1986
Whole
Body
3.8E-1
2.1E-1
1.6E-2
4.9E-1
2.7E-2
Thyroid
5.2E-1
3.4E-1
2.1E-2
1.8E-1
6.4E-2
Bone
2.5E-1
1.6E-1
1.6E-2
6. OE-2
7.2E-2
Liver
3.8E-1
2.1E-1
1.6E-2
4.9E-2
2.8E-2
Lung
3.8E-1
2.1E-1
1.6E-2
4.8E-2
3.2E-2
0
0
0
8
3
Skin
.OE+0
.OE+0
.OE+0
.3E-2
.4E-2
GI-
Tract
3.8E-1
2.1E-1
1.6E-2
4.9E-2
2.8E-2
Kidney
3.8E-1
2.1E-1
1.6E-2
5. OE-2
2.9E-2
en
o
RANCHO SECO
GRAND GULF
YANKEE -ROWE
CRYSTAL RIVER
RIVER BEND
OCONEE
PEACH BOTTOM
50-312
50-416
50-29
50-302
50-458
50 287
50-278
1985
1987
1985
1986
1987
1984
1986
1987
1985
1985
1986
1987
1985
1986
1986
1985
1987
1.
3.
9.
6.
0.
0.
2.
2.
2.
2.
1.
3.
1.
8.
1.
4.
1.
7E-1
4E-1
OE-2
8E-2
OE+0
OE+0
1E.-1
OE-1
2E-2
OE-1
7E-1
9E-2
5E-1
7E-2
2E-1
1E-2
5E-2
1.7E-1
9.4E-1
0 . OE+0
O.OE+0
0 . OE+0
3. OE-2
3.8E-3
2.7E-2
3.1E-1
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
9.7E-1
7. OE-1
1 . 2E+0
1.3E-1
1.7E-1
0 . OE+0
0 . OE+0
O.OE+0
2. OE+0
7.2E-1
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
1.7E-1
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
1.7E-1
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
4.6E-1
1 . 1E+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
5.5E-1
5.8E-1
0 . OE+0
3.9E-1
3.2E-1
O.OE+0
9.1E-1
2.5E-1
2.2E-1
2.1E-1
4.3E-2
1.7E-1
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
3.9E-1
0 . OE+0
0 . OE+0
7.7E-1
1.2E+0
1.4E-1
1.7E-1
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
-------
I
o\
Table 4-31. Doses to maximally exposed individuals in mrem/y (continued).
Bone Liver Lung Skin
Facility
Docket Year
Whole
Body Thyroid
GI-
Tract Kidney
ST. LUC IE 1
KEWAUNEE
FARLEY
MILLSTONE 1
WOLF CREEK
TROJAN
COOK
FT ST VRAIN
SEQUOYAH
50-335 1986
1985
1987
50-305 1986
50-348 1985
1986
1987
50-245 1986
1987
1985
50-482 1988
1987
50-344 1985
50-315 1985
1987
1986
50-267 1987
1986
1985
50-327 1985
1986
1987
1.1E-2
1.3E-2
2.3E-3
1.2E-1
1.3E-1
1.2E-1
8.1E-2
2.2E-1
8.3E-2
7.0E-3
8.2E-2
6.5E-2
6.9E-2
5.7E-2
2.4E-2
2.0E-2
1.9E-1
4.3E-3
7.3E-5
1.9E-1
2.0E-3
0 . OE+0
5
4
7
1
1
9
5
7
1
7
0
0
0
1
1
2
0
0
0
5
0
0
.8E+0
.2E+0
.6E-1
.3E-2
.8E-1
.OE-2
.4E-2
.OE-4
.5E-3
.OE-4
.OE+0
.OE+0
.OE+0
.9E+0
.3E+0
.7E-1
.OE+0
.OE+0
.OE+0
.4E-2
.OE+0
.OE+0
1
1
2
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
.5E-2
.1E-2
.OE-3
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
1.9E-2
1.8E-2
3.2E-3
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
5. OE-4
4. OE-3
8.6E-4
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
1.4E-1
3.0E-1
2.7E-1
3.3E-1
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
1.7E-1
1.8E-1
1.5E-1
5.6E-2
0 . OE+0
O.OE+0
O.OE+0
4.4E-1
2. OE-3
0 . OE+0
1.3E-3
4.5E-3
9.6E-4
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
2.4E-2
2.8E-2
4.8E-3
5. OE-3
9.7E-4
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
-------
o\
to
Table 4-31. Doses to maximally exposed individuals in mrem/y (continued).
Whole GI-
Facility Docket Year Body Thyroid Bone Liver Lung Skin Tract Kidney
HB ROBINSON
HATCH
SUSQUEHANNA
MONTICELLO
DRESDEN
ST.LUCIE 2
ZION
BRUNSWICK
WASHINGTON
TURKEY POINT 3
50-261
50-321
50-388
50-263
50-249
50-389
50-295
50-324
50-397
50-250
1987
1986
1987
1986
1985
1985
1987
1986
1987
1985
1986
1984
1985
1987
1986
1984
1985
1987
1987
1985
1986
1986
1987
6
1
1
4
6
1
1
6
0
0
0
2
6
2
2
9
4
4
2
4
4
4
8
.BE -2
.6E-2
.3E-1
.OE-3
.5E-4
.4E-1
.1E-2
.9E-3
.OE+0
.OE+0
.OE+0
.OE-2
.2E-3
.8E-3
.IE- 3
.2E-2
.4E-2
.7E-4
.8E-2
.2E-2
.1E-2
.2E-3
.7E-3
1.1E-1
3.5E-1
2.6E-1
2.9E-1
9.3E-2
l.OE-1
O.OE+0
0 . OE+0
2 . 6E+0
1 . 3E+0
1 . 2E+0
9.7E-1
2.4E+0
1 . 1E+0
8.9E-1
2.9E-2
7.8E-3
O.OE+0
9.3E-2
O.OE+0
O.OE+0
2.5E-2
2.0E-1
6.5E-2
5.4E-3
1.1E-1
4. OE-3
6.7E-4
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
3.2E-3
2.7E-3
2.4E-3
O.OE+0
O.OE+0
0 . OE+0
3.3E-2
O.OE+0
0 . OE+0
1.9E-3
1.2E-3
6.8E-2
1.8E-2
1.8E-1
7.7E-3
7.9E-4
0 . OE+0
9.8E-2
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
9.2E-3
4. OE-3
3.3E-3
O.OE+0
0 . OE+0
O.OE+0
2.8E-2
O.OE+0
O.OE+0
5.8E-3
9.7E-3
7. OE-2
1.5E-2
2. OE-2
1.9E-3
5.1E-4
0 . OE+0
0 . OE+0
7.3E-3
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
1.6E-3
7.7E-4
5.0E-4
O.OE+0
0 . OE+0
1.6E-2
2.8E-2
O.OE+0
O.OE+0
3.8E-1
8.4E-3
1.8E-1
1.5E-2
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
2. OE-2
O.OE+0
0 . OE+0
0 . OE+0
4. OE-2
O.OE+0
0 . OE+0
O.OE+0
4.7E-1
4.1E-1
4.4E-3
6.5E-2
0 . OE+0
O.OE+0
2.0E-4
8.6E-5
6.8E-2
1.5E-2
6.2E-2
4.9E-3
2.2E-3
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
1.9E-3
9.0E-4
6.0E-4
O.OE+0
O.OE+0
O.OE+0
2.8E-2
O.OE+0
O.OE+0
3.6E-3
8.3E-3
6.8E-2
1.7E-2
8.1E-3
5.1E-3
5.2E-4
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
2.6E-3
1.2E-3
9.1E-4
0 . OE+0
O.OE+0
0 . OE+0
2.8E-2
O.OE+0
O.OE+0
2.7E-3
6.8E-3
-------
Table 4-31. Doses to maximally exposed individuals in mrem/y (continued).
Facility
HARRIS
SALEM
NMPNS
NORTH ANNA
BROWNS FERRY
CALLAWAY
PRAIRIE ISLAND
ARKANSAS 1
BEAVER VALLEY
Docket Year
50-400 1987
50-311 1987
1986
1985
50-220 1987
1986
1985
50-338 1985
1986
1987
50-296 1985
1986
1987
50-483 1986
1987
1985
50-282 1985
1987
50-313 1986
1987
50-334 1986
1987
Whole
Body
2
4
2
1
2
1
2
0
0
0
6
0
0
3
1
6
0
0
6
4
2
1
.2E-2
.7E-2
.8E-2
.6E-2
.4E-2
.3E-2
.5E-4
.OE+0
.OE+0
.OE+0
.OE-2
.OE+0
.OE+0
.4E-2
.6E-2
.9E-3
.OE+0
.OE+0
.OE-3
.4E-3
.3E-2
.4E-3
Thyroid
2.2E-2
0 . OE+0
0 . OE+0
0 . OE+0
7.3E-1
4.8E-1
9.8E-3
1 . 3E+0
8.0E-1
4.4E-1
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
8.3E-1
5.4E-3
9.2E-2
1.7E-3
Bone
2.
0.
0.
0.
0.
0.
0.
0.
0.
0.
3.
1.
0.
0.
0.
0.
0.
0.
4.
7.
7.
3.
2E-2
OE+0
OE+0
OE+0
OE+0
OE+0
OE+0
OE+0
OE+0
OE+0
7E-2
OE-2
OE+0
OE+0
OE+0
OE+0
OE+0
OE+0
5E-3
IE -4
1E-3
5E-6
Liver
2.2E-2
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
7.5E-3
4.8E-4
2.4E-2
1.4E-3
Lung
2.2E-2
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
4.3E-3
4.3E-3
2.5E-2
1.4E-3
Skin
5. OE-2
1.1E-1
6.4E-2
3.5E-2
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
l.OE-1
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
GI-
Tract
2.2E-2
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
7.6E-3
O.OE+0
0 . OE+0
O.OE+0
5.6E-1
6.6E-2
4.5E-3
4.3E-3
2.3E-2
1.4E-3
Kidney
2.2E-2
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
8.7E-3
4.6E-3
2.3E-2
1.4E-3
-------
Table 4-31. Doses to maximally exposed Individuals In mrem/y (continued).
Whole GI-
Facillty Docket Year Body Thyroid Bone Liver Lung Skin Tract Kidney
LIMERICK
QUAD CITIES 1
QUAD CITIES 2
MILLSTONE 2
*-
1
2 CALVERT
TURKEY POINT 4
3 MILE ISLAND
MILLSTONE 3
PALISADES
YANKEE -ROVE
MCGUIRE 2
50-352
50-254
50-265
50-336
50-317
50-251
50-289
50-423
50-255
50-29
50-370
1987
1986
1985
1987
1985
1987
1985
1987
1986
1987
1987
1986
1986
1987
1987
1986
1987
1985
1986
1985
1987
1986
2
7
2
2
2
2
1
1
1
0
8
3
1
2
1
5
0
0
0
0
3
0
.2E-4
.9E-4
.OE-2
.5E-3
.OE-2
.1E-3
.5E-2
.3E-2
.OE-2
.OE+0
.8E-3
.8E-3
.9E-2
.8E-3
.7E-2
.2E-4
.OE+0
.OE+0
.OE+0
.OE+0
.6E-3
.OE+0
0 . OE+0
O.OE+0
1.6E-1
1.2E-1
1.3E-1
9.9E-2
3.8E-2
4. OE-2
4.3E-2
4.4E-1
2.2E-1
3.2E-2
0 . OE+0
O.OE+0
1.4E-2
l.OE-1
O.OE+0
l.OE-1
7.3E-3
0 . OE+0
0 . OE+0
4.3E-1
2.1E-1
4.5E-2
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
1.3E-3
1.3E-3
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
2.0E-1
0 . OE+0
O.OE+0
6.9E-2
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
9.8E-3
5.1E-3
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
8.3E-3
3.8E-3
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
5.7E-4
1.5E-3
4.6E-2
8.8E-3
4.3E-2
4.5E-3
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
1.2E-4
1.9E-4
4.6E-2
8.0E-3
O.OE+0
7.1E-4
O.OE+0
O.OE+0
O.OE+0
O.OE+0
l.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
8.3E-3
3.6E-3
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
6.8E-3
2.5E-3
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
-------
Ul
Table 4-31. Doses to maximally exposed individuals in mrem/y (continued).
Whole
Facility Docket Year Body Thyroid Bone Liver Lung Skin
GI-
Tract Kidney
DAVIS -BESSE
VERMONT YANKEE
SAN ONOFRE
SURRY
ARKANSAS 2
INDIAN FT
BYRON 1
KEVAUNEE
SAN ONOFRE 1
CLINTON
50-346
50-271
50-361
50-281
50-368
50-286
50-454
50-305
50-206
50-461
1987
1985
1986
1987
1985
1985
1986
1987
1987
1986
1986
1987
1986
1985
1985
1987
1987
1985
1987
1988
1
8
6
0
0
0
0
0
0
0
1
2
4
7
1
3
8
0
0
2
2
.2E-2
.IE- 3
.4E-4
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.7E-3
.3E-3
.9E-4
.8E-4
.5E-3
.1E-4
.1E-5
.OE+0
.OE+0
.1E-4
.1E-5
4. OE-2
5.6E-2
6.4E-4
4.2E-1
0 . OE+0
4.1E-1
1.4E-1
4.9E-2
3.6E-1
3.5E-2
3.6E-2
7. OE-3
6.2E-2
2.9E-2
2.2E-2
3.1E-2
2.2E-2
1.6E-2
1.4E-2
2.6E-4
8.1E-3
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
9.2E-4
1.8E-4
2. OE-4
4.9E-4
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
6.4E-5
3.0E-5
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
2.3E-3
2.9E-3
6. OE-4
8.5E-4
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
2.1E-4
3.5E-5
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
2.4E-3
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
1.6E-3
2.8E-3
4.1E-4
7.5E-4
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
2.3E-4
l.OE-5
3. OE-2
2.9E-3
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
2.9E-3
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
1.6E-3
2.8E-3
4.2E-4
7.5E-4
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
2.1E-4
1.1E-5
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
2. OE-3
2.9E-3
6.3E-4
8.3E-4
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
2.1E-4
3.9E-5
-------
Table 4-31. Doses to maximally exposed individuals in mrem/y (continued).
Whole GI-
Facility Docket Year Body Thyroid Bone Liver Lung Skin Tract Kidney
VIRGIL SUMMER 50-395
DIABLO CANYON 1 50-275
DIABLO CANYON 2 50-323
1986
1984
1987
1987
1986
1985
1986
1987
1985
5
1
1
0
0
0
0
0
0
.1E-4
.5E-4
.1E-6
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
.OE+0
0
0
0
4
3
1
4
2
4
.OE+0
.OE+0
.OE+0
.7E-3
.5E-3
.4E-3
.3E-3
.9E-3
.1E-5
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
0 . OE+0
O.OE+0
0 . OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
O.OE+0
0 . OE+0
0 . OE+0
0 . OE+0
O.OE+0
0 . OE+0
I
o\
o\
-------
4.6 SUMMARY
Estimates of dose rates and fatal cancer risks resulting
from atmospheric emissions of radionuclides from the uranium fuel
cycle facilities evaluated in this study are summarized in
Table 4-32.
Table 4-32. Summary of fatal cancer risks from atmospheric
radioactive emissions from uranium fuel cycle
facilities.
Highest Individual Regional (0-80 km)
Lifetime Fatal Population
Facility Cancer Risk Deaths/y
Uranium Mills
Ambrosia Lake
Homestake
La Sal
Lucky Me
Panna Maria
Sherwood
Shirley Basin
Shootaring
Sweetwater
White Mesa
Model Inactive Tailings
2E-7
2E-4
2E-6
1E-7
3E-6
1E-6
6E-7
2E-7
7E-7
6E-7
2E-4
3E-5
2E-3
3E-5
7E-6
5E-5
8E-5
9E-5
7E-7
2E-5
2E-5
1E-4
Uranium Conversion
Dry 3E-5 8E-4
Wet 4E-5 6E-4
Fuel Fabrication 4E-6 8E-5
Nuclear Power Reactors
Pressurized
Water Reactors 3E-6 7E-4
Boiling Water
Reactors 5E-6 1E-3
Where actual facilities are assessed, estimates for nearby
individuals and for regional populations reflect the actual
demography of the site. Where model facilities were used, the
estimates for nearby individuals were made at 500 meters in the
predominant wind direction, and the estimates for the regional
population were made using a reference site. The estimates of
organ dose equivalent rates to the nearby individuals for all
facilities are below 75 mrem/y, except for the Homestake uranium
mill and the model inactive tailings which have an estimated lung
dose equivalent of 87 mrem/y and 98 mrem/y, respectively. The
4-67
-------
doses for the Homestake mill will be lower when the new effluent
control system for the yellowcake processing area is installed
(Fa88).
A summary of the estimated distribution of lifetime fatal
cancer risks from uranium fuel cycle facilities is presented in
Table 4-33. The cumulative risk estimates have been computed by
aggregating the estimated distributions, constrained to the U.S.
population, for each of the individual fuel cycle facilities. The
total number of incremental cancer deaths per year attributed to
uranium fuel cycle facilities is estimated to be 9E-2. The
total number of people estimated to incur an incremental risk of
l.OE-3 to l.OE-4 from these facilities is 84, while 6,600 people
are predicted to incur an incremental risk of l.OE-4 to l.OE-5,
42,000 people are predicted to incur an incremental risk of
l.OE-5 to l.OE-6, and 240,000,000 people are predicted to incur
an incremental risk of less than l.OE-6.
Table 4-33. Estimated distribution of lifetime fatal cancer
risks for uranium fuel cycle facilities.*
Risk Number of Persons Deaths/y
1E-01 to 1E+00 0 0
1E-02 to 1E-01 0 0
1E-03 to 1E-02 0 0
1E-04 to 1E-03 84 2E-4
IE-OS to 1E-04 6,600 1E-3
1E-06 to IE-OS 42,000 2E-3
< l.OE-06 240,000,000 9E-2
Totals 240,000,000 9E-2
*Computed as the aggregate of the estimated distributions for
each of the individual fuel cycle segments multiplied by the
number of facilities of that type. The number of facilities
of each type is as follows:
Uranium Mills - Active 4
- Standby 7
- Inactive 15
Uranium Conversion - Dry l
- Wet l
Fuel Fabrication 5
Pressurized Water Reactors 63
Boiling Water Reactors 37
4-68
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4.7 REFERENCES
AEC73 U.S. Atomic Energy Commission, "Proposed Rule Making
Action: Numerical Guides for Design Objectives and
Limiting Conditions for Operation to meet the Criterion
"As Low As Practicable1 for Radioactive Material in
Light-Water-Cooled Nuclear Power Reactor Effluents,"
WASH-1258, July 1973.
AEC74 U.S. Atomic Energy Commission, Fuels and Materials
Directorate of Licensing, "Environmental Survey of the
Uranium Fuel Cycle," April 1984.
Co74 Cooke, N. and Holt, F.B., "The Solubility of Some Uranium
Compounds in Simulated Lung Fluid," Health Physics,
Vol. 27, No. 1, 1974.
De79 Dennis, N.A., "Dissolution Rates of Yellowcake in
Simulated Lung Fluids," Master's Thesis, University of
Pittsburgh, Department of Radiation Health, 1979.
De82 Dennis, N.A. and Blauer, H.M., "Dissolution Fractions and
Half-Times of Single Source Yellowcake in Simulated Lung
Fluids," Health Physics, Vol. 42, No. 4, April 1982.
Do88 Dolezal, W., formerly of Kerr-McGee Nuclear Corporation,
Sequoyah, Oklahoma, personal communication with D. Goldin,
SC&A, Inc., September 1988.
DOE87 U.S. Department of Energy, Energy Information
Administration, Commercial Nuclear Power 1987.
Prospects for the United States and the World.
DOE/EIA-0438(87), July 1987.
EPA79 U.S. 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.
EPA84a U.S. Environmental Protection Agency, Radionuclides:
Background Information Document for Final Rules
(Vol. I), EPA 520/1-84-022-1, October 1984.
EPA84b U.S. Environmental Protection Agency, Radionuclides;
Background Information Document for Final Rules
(Vol. II), EPA 520/1-84-022-2, October 1984.
EPA86 U.S. Environmental Protection Agency, Final Rule for
Radon-222 Emissions from Licensed Uranium Mill
Tailings. Background Information Document. EPA
520/1-8-009, August 1986.
4-69
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Fa88 Parrel, R., Radiation Safety Officer, Homestake Mills,
personal communication with SC&A, Inc. personnel,
December 1988.
Ha83 Hannery, K., Towards Intrinsically Safe Light Water
Reactorsf Oak Ridge Associated Universities Report,
ORAU/IEA-83-2(M), 1983.
ICRP66 "Deposition and Retention Models for Internal Dosimetry of
the Human Respiratory Tract," Task Group on Lung Dynamics,
Health Physics, Vol. 12, 1966.
Jo81
Ka80
Mo 8 4
NRC74
NRC79
NRC80
NRC82
Jones, J.Q., "Uranium Production," Uranium Industry
Seminar Proceedings, October 21-22, 1981, Grand Junction,
Colorado, Department of Energy, GAO-108(81), 1981.
Kallkwarf, D.R., "Solubility Classification of Airborne
Uranium Products Collected at the Parameter of the Allied
Chemical Plant, Metropolis, Illinois," NUREG/CR-1316,
U.S. Nuclear Regulatory Commission, Washington, D.C., 1980.
Moore, E.B., "Control Technology for Radioactive Emissions
to the Atmosphere at U.S. Department of Energy Facilities,"
PNL-4621 Final, Pacific Northwest Laboratory, Richland,
Washington, October 1984.
U.S. Nuclear Regulatory Commission, Environmental
Statement Related to Operation of Shirley Basin Uranium
Millf December 1974
U.S. Nuclear Regulatory Commission, "Generic
Environmental Impact Statement on Uranium Milling"
(Draft), Vol.11, NUREG-0511, April 1979.
U.S. Nuclear Regulatory Commission, Final Generic
Environmental Impact Statement on Uranium Milling.
September 1980.
U.S. Nuclear Regulatory Commission, Technical Guidance
for Siting Criteria Developmentf NUREG/CR-2239,
November 1982.
NRC84 U.S. Nuclear Regulatory Commission, "Environmental
Impact Appraisal for the Renewal of Source Material
License No. SUB-526," NUREG-1071, May 1984.
NRC85a U.S. Nuclear Regulatory Commission, "Environmental
Assessment for Renewal of Special Nuclear Material
License No. SNM-1107," NUREG-1118, May 1985.
4-70
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NRC85b U.S. Nuclear Regulatory Commission, "Environmental
Assessment for Renewal of Source Material
License No. SUB-1010," NUREG-1157, August 1985.
PNL84 Pacific Northwest Laboratory, "Estimated Population Near
Uranium Tailings," PNL-4959, WC-70, Richland, WA, January
1984.
Ri81 Rives, F.B. and Taormina, "Worldwide U^OQ Producer
Profiles," Nuclear Assurance Corporation, Grand
Junction, Colorado, 1981.
TEK81 Teknekron Research, Inc., "Technical Support for the
Evaluation and Control of Radioactive Materials to Ambient
Air," prepared for the U.S. Environmental Protection
Agency, Office of Radiation Programs, Washington, D.C.,
May 1981.
4-71
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-------
5. HIGH-LEVEL WASTE DISPOSAL FACILITIES
The Nuclear Waste Policy Act of 1982 (the Act) provides that
spent nuclear fuel and transuranic high-level radioactive wastes
be disposed of in deep geologic repositories (NWP83). The term
"high-level wastes" is used throughout this chapter to include
all the materials covered by the Act. High-level waste
repositories, whether for civilian or defense waste, will be
operated by the Department of Energy (DOE) and licensed by the
Nuclear Regulatory Commission (NRC). The Act also directed the
Secretary of Energy to investigate the need for, and the
feasibility of, monitored retrievable storage for high-level
wastes. DOE is also developing a repository for disposal of
radioactive waste from national defense programs.
The High-Leyel Waste Disposal Facility source category
includes facilities designed to handle the interim or ultimate
disposal of high-level radioactive wastes, as defined by 40 CFR
191. No such facility is in operation in the United States.
Therefore, this assessment evaluates the risks from the two
currently planned facilities. These are the Waste Isolation
Pilot Plant (currently under construction in Carlsbad, New
Mexico) and a geologic repository at Yucca Mountain, Nevada.
Both facilities are subject to the standards established by 40
CFR 191.
v>
A monitored retrievable storage (MRS) facility, which would
be subject to the standards established by 40 CRF 191, is also
being planned. However, the MRS facility has not been included
in this assessment since the, facility is not to be used as a
final disposal site.
5.1 DESCRIPTION OF THE HIGH-LEVEL WASTE DISPOSAL FACILITIES
5.1.1 General Description
High-level wastes comprise those materials that the
Environmental Protection Agency (EPA) has regulated under
40 CFR 191. These include:
1. used nuclear fuel when there is no intent to reprocess;
2. liquid wastes resulting from the operation of the
first solvent extraction cycle (or equivalent) in a
facility for reprocessing spent nuclear fuel, the
concentrated wastes from subsequent extraction cycles
(or equivalent), and solids into which such liquids
have been converted; and
3. wastes containing more than 100 nanocuries per gram of
transuranic elements with half-lives greater than 20
years.
5-1
-------
In 1978, the NRC gave projected values for production of
spent fuel at light-water power reactors (NRC78). These values
are given in Table 5-1.
Table 5-1. Projected generation of spent fuel.
Year MTHM(cum)(a)
1980
1985
1990
1995
2000
7,200
18,000
33,000
59,000
95,000
(a) MTHM = metric tons of heavy metal
Source: NRC78
The projected amount of high-level and waste to be disposed
of by placement in a geologic repository shortly after the turn
of the century is about 70,000 metric tons of uranium (MTU), or
equivalent. Of this, about 62,000 MTU would be spent fuel from
civilian reactors, and 8,000 MTU-equivalents would be defense
waste (including waste from West Valley, New York). The
difference between the 95,000 MTU of spent fuel shown in Table 5-
1 and 62,000 MTU to be placed in the repository would be
accounted for by at-reactor storage and interim storage not at
the reactor (DOES5).
This chapter is limited to evaluation of the air emissions
from facilities specifically used for handling, storage, and
final disposal of high-level wastes. Emissions from such
materials at reactors or at DOE facilities, such as Hanford, the
Savannah River Plant, or the Idaho National Engineering
Laboratory (INEL), are included in the assessments of Uranium
Fuel Cycle Facilities and DOE Facilities (see Chapters 4 and 2,
respectively).
5.1.2 Facility and Process Descriptions
The following subsections describe the operations that
result in the release of radioactive materials to the atmosphere
at each of the two facilities.
5.1.2.1 The Waste Isolation Pilot Plant
The Waste Isolation Pilot Plant (WIPP) is for the disposal
of defense radioactive waste, primarily transuranic wastes, in a
mined geologic repository in salt. Transuranic wastes are
5-2
-------
designated as contact-handling (CH) and remote-handling (RH). At
the facility, packaged waste containers are inspected,
decontaminated, and prepared for underground disposal (DOE86a).
Most operations at WIPP are done in the waste-handling (w-h)
building, which has separate areas for the receipt, inventory,
inspection, and transfer of CH and RH transuranic (TRU) wastes.
Air exhausted to the atmosphere from this building is filtered
through HEPA filters.
Contact-handling TRU waste shipping containers on rail cars
and trucks will enter the w-h building through airlocks. After
inspection for contamination, acceptable packages will be moved
to the CH-waste inventory and preparation room and transported
underground. Contaminated or damaged containers will be
decontaminated, overpacked or repaired, and sent to the inventory
and preparation room to be transported underground.
Remote-handling TRU waste shielded shipping casks on rail
cars and trucks will be unloaded, inspected, and decontaminated
if necessary. Each cask will then be moved to the cask
preparation and decontamination area for necessary treatment and
then to the cask unloading room. The RH-waste canisters will be
unloaded from the casks into the hot cell. Any contaminated or
damaged canister will be inserted into an overpack. The
canisters will be moved from the hot cell into a facility
transfer cask for transfer to the underground disposal area.
Both packaged CH- and RH-wastes are emplaced in holes in the
bedded-salt underground mine matrix. Disposal area ventilation
air is routed through the disposal exhaust shaft to the disposal
exhaust filtration building. This exhaust is not filtered except
when monitors indicate radioactive material releases. Then, air
flow volumes are approximately halved and diverted through HEPA
filters.
5.1.2.2 Yucca Mountain Geologic Repository
The function of a repository is the permanent isolation of
high-level radioactive waste. The Yucca Mountain site will
contain a mined repository for the geologic disposal of spent
fuel and processed defense high-level waste in accordance with
the provisions of the Nuclear Waste Policy Act of 1982. The Act
provides a limit of the equivalent of 70,000 MTU for this first
repository.
Both unconsolidated and consolidated spent fuel will be
handled at the repository. Phase 1 provides for the disposal
into the mine of about 400 MTU per year of spent, unconsolidated
fuel. The unconsolidated fuel will be packaged at the
repository. In Phase 2, facility capacity will be increased to
3,000 MTU per year, and the facility will receive wastes other
than spent fuel.
5-3
-------
The main surface components of the facility are two waste
handling buildings and a waste treatment building (for waste
generated onsite). There is also an access portal to the ramp
leading to the mine itself. Surface facilities will occupy about
0.6 square kilometers. Most operations take place in hot cells
in the waste handling building. Emissions from these cells are
discharged through multiple-stage HEPA filters.
The underground repository is a mined area in the tuff
matrix of the site. It will occupy about 1,500 acres (6 square
kilometers) at a depth of more than 230 meters. Conventional
mining room-and-pillar construction will be adequate for the
repository.
Radioactive waste will be shipped to the repository in
federally licensed transport casks. In the earliest (Stage 1)
operations, about 1,000 truck shipments and 500 rail shipments of
spent fuel assemblies, amounting to a total of 400 MTU, would be
received each year. In the second phase, receipts would increase
to 3,000 MTU per year (DOE86b).
In the first phase, only unconsolidated spent fuel will be
emplaced in the repository. In the second phase, spent fuel will
also be consolidated and repackaged for burial. This
consolidation is essentially the same operation as that performed
at the Monitored Retrieval Storage facility .
5.1.3 Emission Controls
The primary emission control for all these facilities is the
waste package. The waste is contained in massive steel
canisters, which are welded to be leak-free. A secondary
emission control for all the facilities is HEPA filters, which
are fitted to the cells in which operations that could release
radionuclides to the atmosphere take place. HEPA filters are
also provided for the underground area ventilation stack of the
Yucca Mountain mine.
5.2 BASIS OF THE EXPOSURE AND RISK EVALUATION
5.2.1 Emissions
As none of the high-level waste disposal facilities is in
operation, source terms must be based on engineering estimates.
The Agency has reviewed the estimates made by DOE and has found
that they are conservative. Thus, the DOE's engineering
estimates of source terms are used in this evaluation. The
estimated emissions for each facility are given in Table 5-2.
5-4
-------
Table 5-2. Emissions from normal operations at HLW disposal
facilities.
Radionuclide
Release Rates
(Ci/y)
WIPP
H-3
C-14
Kr-85
1-129
Pu-238
Pu-239
Pu-240
Pu-241
Am-241
Cm-244
Yucca
2.8E+2
1.1E+1
1.4E+4
2.8E-2
6.6E-8
4.6E-8
l.OE-8
2.8E-6
1.6E-7
2.4E-8
Source: DOE86a
5.2.1.1 Waste Isolation Pilot Plant
Emissions to air from normal operations arise from
radioactive contamination of the surface of received containers
and from containers found damaged or defective on receipt.
Calculations of emissions are based on the design maximum annual
throughput of 34,000 drums and 2,200 boxes of CH TRU waste and
250 canisters of RH TRU waste: The emission values given in
Table 5-2 were obtained from DOE documents (DOE80, DOE86a). The
HEPA filter decontamination value appears to be too high, but
this is counterbalanced by the very conservative assumptions as
to the extent of surface contamination and of defective packages.
It is assumed that all the contact-handling packages have
surface contamination at the maximum level permitted by the Waste
Acceptance Criteria and that 100 drums and 10 boxes per year are
defective or damaged. It is estimated that 0.1 percent of the
surface radioactivity of contaminated CH TRU packages is
resuspended and becomes airborne in the waste-handling (w-h)
building, and that a further 0.1 percent becomes resuspended in
the underground disposal area. It is also estimated that
1 percent of the content of defective packages is spilled in the
w-h building and that 0.1 percent of this spilled amount becomes
airborne. The material that becomes airborne in the w-h building
is discharged to the atmosphere through two stages of HEPA
filtration, with an estimated decontamination factor of 106. The
exhaust air from the underground area bypasses filtration except
when monitors indicate a high radioactivity level.
5-5
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It is assumed that all the remote-handling packages have
surface contamination at the maximum level permitted by the Waste
Acceptance Criteria and that one package per year is defective
or damaged. It is estimated that 0.1 percent of the surface
radioactivity of contaminated RH TRU packages is resuspended and
becomes airborne in the w-h building and that a further
0.1 percent becomes resuspended in the underground disposal area.
It is further estimated that 0.1 percent of the content of
damaged or defective packages becomes airborne in the hot cell.
The airborne activity in the w-h building is discharged to the
atmosphere through two stages of HEPA filtration.
The only pathways for direct emission to the air after
closure would be from volcanic action or a hit by a meteorite
(Sm82). The mine is placed very deep in a non-volcanic area.
Only a meteorite so large that its occurrence is extremely
improbable could penetrate to this depth. The post-closure
emission rate to air is therefore assumed to be zero.
5.2.1.2 Yucca Mountain Geologic Repository
Any emissions to air from normal operations would arise
primarily from handling spent fuel assemblies. The fraction of
failed fuel rods is estimated at 0.02 percent (Wo83). In
addition, during consolidation, there is some damage to fuel rods
that have become bound to the assembly spacers. The fraction of
fuel rods damaged in this way is estimated at about 0.3 percent.
Only volatile nuclides are projected to be emitted, because all
these releases would occur only in filtered hot cells.
In estimating annual emissions, a processing rate of 3,000
MTU of 10-year-old spent fuel per year is assumed. The release
fractions of 0.3 for krypton-85 and 0.1 for iodine-129 given in
Regulatory Guide 1.25 (NRC72) have been used, and release
fractions of 0.1 for tritium and carbon-14 have been assumed.
5.2.2 Other Parameters Used in the Assessment
5.2.2.1 Dispersion
5.2.2.1.1 Discharge Height and Location
Useful information on stack characteristics is available
only for the Waste Isolation Pilot Plant, the only facility whose
design is sufficiently advanced. For the other facility, the
characteristics of the WIPP waste-handling building stack have
been used since operations in there are very similar to the
operations at WIPP. Again for lack of information, it is assumed
that the discharge points are at the center of the site. In the
WIPP analysis, a correction was made for the momentum of the air
leaving the stacks. In the Yucca Mountain analysis, for
conservatism, no corrections were made for plume rise or
buoyancy. Radioactive wastes from WIPP are discharged through
two stacks, one for the waste-handling building and one for
storage exhaust. The WIPP stacks are described in Table 5-3.
5-6
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Table 5-3. WIPP discharge stacks.
Stack
Height Diameter Flow Rate Velocity Filtration
(m) (m) (m3/s) (m/s)
Waste handing 10
Building
Storage 7.3
Exhaust
Sources: DOE86a; Ch88
2.1
3.1
42.6
198.6
11.9
27.2
Continuous
Only when
airborne
activity
in area
5.2.2.1.2 Meteorology
Meteorological data from nearby airports or nuclear
facilities were used. For the WIPP (Los Medanos) site, which is
approximately 25 miles east of Carlsbad, New Mexico,
meteorological data from the Carlsbad airport were used. For the
Yucca Mountain site, meteorology for the Nevada Test Site (NTS),
which is immediately adjacent to Yucca Mountain, was used.
5.2.2.2 Population Distribution
The computer code SECPOP was used to develop the population
distributions for the circular area 80 kilometers in radius
around each discharge point.
At the WIPP site, there are only a few people closer than
20,000 meters. The location of the nearest individual is
800 meters from the source. The Yucca Mountain site is located
on and immediately adjacent to the southwestern corner of the
Nevada Test Site (NTS), about 137 kilometers (85 miles) northwest
of Las Vegas, Nevada. The Federal Government controls all of the
site land. About 33 percent is on the NTS, 40 percent on Nellis
Air Force Range (NAFR), and about 25 percent on Bureau of Land
Management (BLM) land. None of the land is presently used. The
NAFR land is in an area used only for overflight. The nearest
grazing lease on the BLM land is about 5 kilometers west of
the site. An estimated 4,800 persons live within 80 kilometers
of the proposed site, with the nearest individuals approximately
25 kilometers away. The nearest highly populated area is Las
Vegas.
5.3 RESULTS OF THE DOSE AND RISK ASSESSMENT
5.3.1 Exposures and Risks to Nearby Individuals and to Regional
Population
The locations of individuals receiving the highest dose at
the two facilities were 800 meters south-southwest at the WIPP
5-7
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and 70,000 meters south at the Yucca Mountain facility. There
are residences closer than 70,000 meters at the Yucca site, but
they are not in a downwind direction. Doses to the selected
individuals and to regional populations are presented in Table
5-4. The organs that contribute the most to risk are identified,
and the dose to each of these organs is given.
Risks to the nearby individual and fatal cancers projected
within a radius of 80 kilometers from each of the facilities are
presented in Table 5-5.
Table 5-4
Facility
Estimated radiation dose rates from high-level
waste disposal facilities.
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Yucca Mountain
Geologic Repository
Waste Isolation
Pilot Plant
Thyroid
Remainder
Red Marrow
Breast
Gonads
Endosteum
Remainder
Red Marrow
Lungs
3.7E-2
2.6E-3
4.0E-3
2.7E-3
1.6E-3
7.6E-4
3.4E-5
6.2E-5
6.0E-5
1.8E-1
1.1E-2
1.8E-2
1.2E-2
6.7E-3
4.6E-4
2.1E-5
3.7E-5
3.0E-5
Table 5-5. Estimated fatal cancer risks from high-level
waste disposal facilities.
Facility
Nearby Individuals
Lifetime Fatal
Cancer Risk
Regional (0-80 km)
Population
Deaths/y
Waste Isolation Pilot
Plant
Yucca Mountain Geologic
Repository
3E-10
7E-8
2E-9
4E-6
5.3.1.1 Waste Isolation Pilot Plant
The most important pathway for dose to the selected
individual is inhalation, which accounts for over 99 percent of the
dose. The most important nuclide is americium-241 (51 percent of
5-8
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total dose); next are plutonium-238 (20 percent), plutonium-239
(14 percent), plutonium-241 (6 percent), curium-244 (5 percent),
and plutonium-240 (3 percent).
The pathway contributing most to population dose is also
inhalation (87 percent). Ingestion contributes 13 percent. Air
immersion and exposure to ground surface are not significant.
Americium-241 contributes 57 percent of the population dose; next
come plutonium-238 (14 percent), plutonium-239 (11 percent),
plutonium-241 (11 percent), curium-244 (5 percent), and
plutonium-240 (2 percent).
5.3.1.2 Yucca Mountain
The most important pathway for dose to the selected
individual is ingestion, which accounts for 87 percent of the
dose. The inhalation pathway accounts for 10 percent of the dose,
3 percent comes from immersion, and <1 percent from ground
surface exposure. The most important nuclides are carbon-14
(55 percent of the total dose) and tritium (31 percent); next is
iodine-129 (8 percent) and then krypton-85 (6 percent).
The pathway contributing most to population dose is
ingestion (93 percent). Inhalation contributes 5 percent, air
immersion 2 percent, and exposure to ground surface, <1 percent.
Carbon-14 contributes 59 percent of the population dose; next is
tritium, 29 percent, and then iodine-129 with 9 percent and
krypton-85, 3 percent.
5.3.3 Distribution of the Fatal Cancer Risk from Hiqh-Level
Waste Disposal Facilities
The distribution of fatal cancer risks from all high-level
waste disposal facilities was obtained by adding the number of
people and the projected number of fatal cancers in each risk
interval at the two sites. This distribution is presented in
Table 5-6. There are no persons in the 0-80 km populations with
an estimated lifetime fatal cancer risk greater than 1E-6. The
projected deaths/year of operations in the regional populations
is 4E-6.
5.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS
The facilities that make up the High-Level Waste Disposal
Facility source category are designed with state-of-the-art
effluent control systems. The effectiveness of these systems is
enhanced by the performance requirements of the waste forms and
packages. Given these considerations, and the very small
projected risks to nearby individuals (all less than one in one
million lifetime) and populations (one fatal cancer per 10,000
years), this evaluation does not address supplementary control
options and costs.
5-9
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Table 5-6. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from high-level
waste disposal facilities.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0 0 0
1E-2 to 1E-1 0 0
1E-3 to 1E-2 0 0
1E-4 to 1E-3 0 0
1E-5 to 1E-4 0 0
1E-6 to 1E-5 0 0
< 1E-6 101,000 4E-6
Totals 101,000 4E-6
5-10
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5.5 REFERENCES
Ch88 Channell, J., Personal Communication with A. Goldin, SC&A,
Inc., June 1988.
DOE80 U.S. Department of Energy, "Final Environmental Impact
Statement, Waste Isolation Pilot Plant," Report DOE/EIS-
0026, October 1980.
DOE85 U.S. Department of Energy, "Mission Plan for the Civilian
Radioactive Waste Management Program," Report DOE/RW-0005,
1985.
DOE86a U.S. Department of Energy, "Preliminary Safety Analysis
Report, Waste Isolation Pilot Plant," Amendment 9, May
1986.
DOE86b U.S. Department of Energy, "Environmental Assessment,
Yucca Mountain Site, Nevada Research and Development Area,
Nevada," Report DOE/RW-0073, May 1986.
NRC72 U.S. Nuclear Regulatory Commission, "Assumptions Used for
Evaluating the Potential Radiological Consequences of a
Fuel Handling Accident in the Fuel Handling and Storage
Facility for Boiling and Pressurized Water Reactors," NRC
Regulatory Guide 1.25, 1972.
NRC78 U.S. Nuclear Regulatory Commission, "Generic Environmental
Impact Statement on Handling and Storage of Spent Light
Water Power Reactor Fuel," Report NUREG-0404, March 1978.
NWP83 "Nuclear Waste Policy Act of 1982," Public Law 97-425, 42
USC 10101-10226.
Sm82 Smith, C.B., D.J. Egan, Jr., W.A. Williams, J.M. Gruhlke,
C-Y Hung, and B.L. Serini, "Population Risks from Disposal
of High-Level Radioactive Wastes in Geologic Repositories"
(Draft Report), U.S. Environmental Protection Agency,
Report EPA 520/3-80-06, December 1982.
Wo83 Woodley, R.E., "The Characteristics of Spent LWR Relevant
to Its Storage in Geologic Repositories," Report HEDL-
TME 83-28, Hanford Engineering Development Laboratory,
Richland, WA, 1983.
5-11
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6. ELEMENTAL PHOSPHORUS PLANTS
The elemental phosphorus plant source category consists of
five operating and three standby facilities that produce
elemental phosphorus by the electric furnace method. These
plants have been evaluated in previous EPA assessments under
Section 112 of the Clean Air Act and are subject to the NESHAP
(40 CFR 61, Subpart K) promulgated on February 5, 1985. The
NESHAP established an emissions limit of 21 Ci/y for polonium-210
released from calciners and nodulizing kilns.
This chapter updates the assessment made during the 1983-
1984 NESHAPS rulemaking period for radionuclides (EPA84a).
Revisions have been made where necessary to reflect the changes
in emissions or control technology as reported to the EPA under
provisions of the NESHAP. It also incorporates the exposure and
risk assessments for two idle plants in Florida that were not
addressed in the risk assessment for the 1984 rulemaking.
6.1 DESCRIPTION OF THE SOURCE CATEGORY
6.1.1 Industry Profile
About eight 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. Production of elemental
phosphorus has declined from 330,000 metric tons (MT; one short
ton is equivalent to 0.9072 metric tons) reported in 1983 to
300,000 MT in 1985 and 240,000 MT in 1986 (BM88).
There are eight elemental phosphorus plants in the United
States, located in Florida, Idaho, Montana, and Tennessee.
Location, ownership, estimated capacity, and current status of
the plants are shown in Table 6-1. The three idle facilities,
the two located in Florida and the Monsanto Chemical Company
plant in Columbia, Tennessee, are not expected to reopen. The
decreasing demand for elemental phosphorus, 27 percent in three
years, and the high operating costs, particularly for electricity
in Florida, make these plants uneconomic.
Phosphate rock contains from 20 to 200 ppm uranium, 10 to
100 times higher than the 1 to 2 ppm found in typical rocks and
soil. Heating the phosphate rock to high temperatures in
calciners and electric furnaces, as is done in the production of
elemental phosphorus, volatilizes lead-210 and polonium-210 which
may result in the release of significant quantities of these
radionuclides to the atmosphere.
6.1.2 Process Description
The 1984 Background Information Document (BID) and the
supporting report on Airborne Emission Control Technology for the
6-1
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Elemental Phosphorus Industry (SAI84) provide detailed data on
each plant, including design, operation, source and radionuclide
content of phosphate rock processed, and analyses of particulate
and radionuclide emissions from various parts of the process.
Table 6-1. Elemental phosphorus plants,
Location Company
Capacitya
(MT/y of Phosphorus)
Florida
Pierce(b)
Tarpon Springs^ '
Idaho
Pocatello
Soda Springs
Montana
Silver Bow
Tennessee
Columbia
Columbia
Mt. Pleasant
Mobil Chemical Co.
Stauffer Chemical Co.
FMC Corporation
Monsanto Chemical Co.
Stauffer Chemical Co.^c^
Occidental Chemical Co.
Monsanto Chemical Co.
Stauffer Chemical Co.
18,000
21,000
122,000
95,000
36,000
45,000
121,000
41,000
(a) Estimated capacity in 1984 (SAI84, EPA84b).
(b) These facilities are currently idle (BM88).
(c) In September 1987, Rhone-Poulenc, a French company, acquired
the inorganic chemicals business that had belonged to the
Stauffer Chemical Company.
Crushed and screened phosphate rock is fed into calciners
and heated to the melting point, about 1,300| C. After calcining,
the hot nodules are passed through coolers and into storage bins
prior to being fed into electric furnaces. The furnace feed
consists of the nodules, silica, and coke. A simplified chemical
equation for the electric furnace reaction is:
6SiC-2 + IOC
10CO + 6CaSiC-3
Phosphorus and carbon monoxide (CO) are driven off as gases
and are vented near the top of the furnace. Furnace off -gases
pass through dust collectors and then through water spray
condensers where the phosphorus is cooled to the molten state.
6-2
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The mix of phosphorus and water (phossy water) and mud are then
processed to recover the phosphorus. Clean off-gases from the
condensers contain a high concentration of CO and are used as
fuel in the calciners.
6.1.3 Existing Effluent Controls
Emissions from the calciners are typically controlled by low
energy scrubbers. Since the 1984 assessment of this source
category, one plant has upgraded its calciner emission controls
by installing a high energy scrubber system. Emissions from
nodule coolers, transfer points, and furnace tap holes are
controlled by either fabric filters or wet scrubbers. Screening
plant emissions are usually controlled by fabric filters
6.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT
6.2.1 Emissions
6.2.1.1 Radionuclide Emission Measurements
6.2.1.1.1 Results of 1975-1980 Emission Testing
During 1975-1980, EPA measured the radionuclide emission
rates from three elemental phosphorus plants: FMC in Pocatello,
Idaho (EPA77); Stauffer in Silver Bow, Montana (AnSla); and
Monsanto in Columbia, Tennessee (AnSlb). Measurements were made
from release points representative of all of the major process
operations in the production of elemental phosphorus. The stack
emission rates measured during these studies are summarized in
Table 6-2.
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 more than 95 percent of the lead-210 and
polonium-210 emitted from these facilities are released from the
calciner stacks. The high calcining temperatures volatilize the
lead-210 and polonium-210 from the phosphate rock, resulting in
the release of much greater quantities of these radionuclides
than of the uranium, thorium, and radium radionuclides. Analyses
of doses and risks from these emissions show that the emissions
of polonium-210 and lead-210 are the major contributors to risk
from radionuclide emissions from elemental phosphorus plants (see
Section 6.3).
6.2.1.1.2 Results of the 1983-1984 Emission Testing
In 1983, EPA conducted extensive additional radionuclide
testing at the FMC plant in Pocatello, Idaho (EPA84c, Ra84a) and
at the Stauffer plant in Silver Bow, Montana (EPA84d, Ra84b). In
early 1984, limited emission testing was done at the Monsanto
plant in Soda Springs, Idaho (EPA84e, Ra84c). This testing was
limited to calciner off-gas streams and focused primarily on
lead-210 and polonium-210 emissions in order to obtain additional
6-3
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Table 6-2. Radionuclide stack emissions measured at elemental
phosphorus plants (1975-1980).(a)
FMC Stauffer Monsanto
Parameter Idaho Montana Tennessee
Rock processing rate (MT/y)(b) 1.6E+6 5.3E+5 1.7E+6
U-238 concentration
of rock (pCi/g)(c) 22.0 27.0 5.o(d)
Calciner stack emission rate (Ci/y):(e)
U-238 1.2E-3 2.4E-4 2.2E-3
U-234 1.3E-3 2.0E-4 3.2E-3
Th-230 2.2E-3 1.2E-4 1.4E-3
Ra-226 1.3E-3 3.5E-4 2.1E-3
Rn-222 ND(f) 8.0 9.6
Pb-210 3.0E-3 2.8E-1 4.8E-1
Po-210 6.9 2.0E-1 7.5E-1
Other stacks emission rates (Ci/y):
U-238 4.0E-2 6.2E-4 l.OE-2
U-234 4.6E-2 7.OE-4 l.OE-2
Th-230 5.3E-3 1.2E-3 1.2E-2
Ra-226 5.9E-3 1.1E-3 9.0E-3
Rn-222 ND ND ND
Pb-210 1.5E-2 2.5E-3 ND
Po-210 4.0E-1 5.9E-3 2.7E-3
Fraction of input radionuclides emitted:
U-238 1.2E-3 6.0E-5 1.4E-3
U-234 1.4E-3 6 2E-5 1.5E-3
Th-230 2.1E-4 9.0E-5 1.5E-3
Ra-226 2.OE-4 9.8E-5 1.7E-3
Rn-222 ND 5.7E-1 1.1
Pb-210 5.1E-4 2.0E-2 5.6E-2
Po-210 2.1E-1 1.4E-2 8.8E-2
(a) Emissions are in particulate form except for radon-222 which
is released in gaseous form.
(b) These processing rates were those estimated for these plants
at the time of emission testing.
(c) Uranium-238 and its decay products are assumed to be present
in equilibrium in the rock.
(d) Calciner feed material was a blend of Tennessee and Florida
phosphate rock.
(e) Based on 8,760 hours of plant operation.
(f) ND - Not determined.
6-4
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information on radionuclide concentrations, particle size
distribution, and the lung-clearance classification of these
radionuclides in the calciner off-gases. Sampling of the
calciner off-gases at the Monsanto plant in Soda Springs, Idaho,
was hampered by the unavailability of suitable sampling locations
(for details see Ra84c). The major results of the testing are
summarized below.
Process Samples
Table 6-3 presents the measured radionuclide concentrations
in the calciner feed material and product samples for the three
plants studied. At the Stauffer and Monsanto plants, the
concentrations of lead-210 and polonium-210 were significantly
lower in the calciner product samples than in the feed material,
indicating volatilization of these radionuclides during
calcining. At the FMC plant, only the polonium-210 concentration
was significantly lower in the product samples than in the feed
material, indicating lower volatilization of lead-210 during
calcining at this plant.
Radionuclide Emission Rates
Table 6-4 shows the measured radionuclide emission rates
(/iCi/h/calciner) and the estimated annual calciner emissions for
the three plants studied.
Particle Size Distribution
Table 6-5 presents the particle size distributions of
lead-210 and polonium-210 in the calciner off-gas streams at the
FMC and Stauffer plants (these data could not be obtained at the
Monsanto plant; see Ra84c). At both plants, most of the
polonium-210 (about 75 percent) was associated with particles
smaller than 1 urn.
Lung-Clearance Classification Studies
Table 6-6 summarizes the dissolution data for lead-210 and
polonium-210 in simulated lung fluid for particulate samples from
the FMC and Stauffer plants. The tests showed that both lead-210
and polonium-210 dissolved very slowly in the simulated lung
fluid; more than 99 percent of these radionuclides remained
undissolved after 60 days of testing. 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. A detailed
description of the tests and results is presented in PNL-5221
(Ka84) .
6.2.1.1.3 Results of 1988 Emission Testing
During 1988, EPA conducted additional radionuclide testing
at the FMC plant in Pocatello, Idaho (EPA88a) and at the Monsanto
plant in Soda Springs, Idaho (EPA88b). These measurements were a
6-5
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Table 6-3
Plant
Measured radionuclide concentrations in process
samples at elemental phosphorus plants - 1983-1984
results.
Radionuclide Concentrations (pCi/g)
Feedstock Calcined Product
U-238 Pb-210 Po-210
U-238 Pb-210 Po-210
FMC
Pocatello, ID
Stauffer
Silver Bow, MT
Monsanto(a)
Soda Springs, ID
21
42
32
26
46
150
21
40
91
22
42
37
27
8
(a) Blended feed material. This plant recycles both dropout
chamber dust and underflow solids from wet scrubber clarifier.
Table 6-4,
Plant and
Number of
Calciners
Radionuclide emissions from calciners at elemental
phosphorus plants - 1983-1984 results.
Average Measured
Radionuclide Emissions
(jiCi/h/calciner) (a)
U-238 Pb-210 Po-210
Estimated Total
Calciner Emissions
(Ci/y) (b) (c)
U-238 Pb-210 Po-210
FMC
Pocatello, ID 0.28
(2 calciners)
Stauffer
Silver Bow, MT 0.04
(2 calciners)
7.5 540
7.6
Monsanto
Soda Springs, ID
(1 calciner)
0.78
760
50
2,900
0.004 0.12
0.0006 0.11
8.6
0.74
0.006
5.6
21
(a) For 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 the total annual emissions, it is
assumed that both calciner units have the same emission rate,
(b) Based on 7,400 hours of calciner operation (i.e., 85 percent
operating factor).
(c) Conversion of measured emission rates to annual emission
estimates for the FMC plant includes an adjustment for
processing rate where applicable (see EPA84c).
6-6
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Table 6-5.
Measured distribution of lead-210 and polonium-210 by
particle size in calciner stack outlet streams at
elemental phosphorus plants - 1983 results.(a)
Plant
Particle Size
(Dp50) (urn) (*>)
Stauffer
Silver Bow, MT
0.5
0.9
1.5
3
10
Cumulative Activity
Percentages
Pb-210
Po-210
FMC
Pocatello, ID
0.5
0.9
1.5
3
10
44
58
68
77
90
73
78
84
88
93
54
76
90
95
99
50
74
90
96
98
(a) Particle size measurements using cascade impactors could not
be made at Monsanto, Soda Springs, ID, because suitable
sampling ports and locations were not available.
(b) Dp50 is defined in Ra84a and Ra84b.
Table 6-6.
Plant
Dissolution of lead-210 and polonium-210 from
particulate samples collected from off-gas streams at
FMC and Stauffer elemental phosphorus plants.(a)
Sample
Particle Size
(urn)
Dissolution
Time (days)
Fraction
of Pb-210
Remaining
Undissolved
Fraction
of Po-210
Remaining
Undissolved
FMC
Pocatello,
ID
Stauffer
Silver Bow,
MT
0-3
3-10
0-3
3-10
1.0
10
59
1.0
10
59
1.0
8.9
59
1.0
8.9
59
0.9984
0.9968
0.9950
0.9933
0.9682
0.9490
0.9999
0.9994
0.9978
1.0000
0.9990
0.9979
0.9997
0.9984
0.9978
0.9991
0.9979
0.9914
0.9997
0.9989
0.9980
0.9997
0.9992
0.9940
(a) Adapted from EPA84a.
6-7
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followup to those made earlier (1975-1980 and 1983-1984) , in
order to learn the effect of changes made to the emission systems
since the 1983-1984 study. The testing was limited to measuring
only lead-210 and polonium-210 in calciner off -gas streams and
particle size distributions of the activities emitted.
The emission rates in pCi/h from the calciners at these two
facilities for these radionuclides are listed in Table 6-7. At
the FMC plant, measurements were conducted only at calciner
number 1. Table 6-7 also lists the total curie amounts of
lead-210 and polonium-210 emitted annually from calciners tested.
Table 6-8 shows the particle size distributions of
lead-210 and polonium-210 in the calciner inlet and outlet
streams determined in 1988 at the FMC and Monsanto plants.
6-8
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Table 6-7.
Lead-210 and polonium-210 emissions measured in
calciner off-gas streams at two elemental phosphorus
plants - 1988.
Measured Emission
Rate per Calciner
Plant
Estimated Total
Calciner Emissions
(Ci/y)
Pb-210
Po-210
Pb-210
Po-210
FMC (EPA88a)
Pocatello, ID(a)
Monsanto (EPA88b)
(b)
41
1,208
172
(b) 7.l(°)
0.34(d) 1.4(d)
Soda Springs, ID
(a) Emission rates for Calciner 1 only.
(b) The large uncertainty in the lead-210 measurement data for
this plant made the lead-210 data invalid.
(c) Based on confidential data on plant production rates.
(d) Based on 8,300 hours of kiln operation.
Table 6-8. Measured distribution of lead-210 and polonium-210 by
particle size in calciner stack inlet and outlet
streams at elemental phosphorus plants - 1988
results.(a)
Cumulative Activity Percentages
Particle Size Po-210 Pb-210
Plant Dp50 (urn)
Inlet Outlet
Inlet Outlet
FMC (EPA88a)
Pocatello, ID
0.5
1
2.5
5
10
64
74
84
89
93
72
82
90
95
97
30
46
64
78
87
54
70
87
94
98
Monsanto
(EPA88b)
Soda Springs,
ID
0.5
1
2.5
5
10
60
90
96
98,
70
90
96
98,
99.4 99.4
60
90
97
99.3
99.9
60
90
98
99.3
99.7
(a) Measurements were made using cascade impactors,
6-9
-------
6.2.1.2 Source Terms Used in the Assessment
Table 6-9 shows the estimated annual calciner emission rates
for each of the eight elemental phosphorus plants.
Table 6-9. Estimated annual radionuclide emissions from elemental
phosphorus plants.
Annual Emissions (Ci/y)
Plant U-238laJPb-210Po-210
FMC Corporation(b)
Pocatello, ID
Monsanto Chemical Co.(c)
Soda Springs, ID
Stauffer Chemical Co.(d)
Silver Bow, MT
Stauffer Chemical Co.(e)
Mt. Pleasant, TN
Occidental Chemical Co.
Columbia, TN
Stauffer Chemical Co.(9)
Tarpon Springs, FL
Mobil Chemical Co.(9)
Pierce, FL
Monsanto Chemical Co.(n)
Columbia, TN
Operating Plants
3.2E-3
5.0E-4
6.0E-4
3.0E-4
l.OE-4
1.4E-1
3.5E-1
1.1E-1
5.8E-2
6.4E-2
Idle Plants
l.OE+1
1.4E+0
7.4E-1
2.8E-1
3.1E-1
3.5E-3
1.6E-3
2.0E-3
1.9E-1
1.2E-2
4.1E-1
1.5E-1
1.3E-2
6.4E-1
(a) Uranium-238 is assumed to be in radioactive equilibrium with
uranium-234, thorium-230, and radium-226 (see Table 6-2).
Uranium-238 emissions are estimated by multiplying the mass
emissions by the specific activity of uranium in the
feedstock.
(b) Based on EPA emission tests in 1983 (EPA84c) and 1988
(EPA88a).
(c) Based on Table 6-7.
(d) Based on Table 6-4.
(e) Assumed similar to emissions from the Occidental Chemical Co.
plant at Columbia, TN, and adjusted for production capacity
(41 MT/45 MT) (see Table 6-1).
(f) Based on reference Bu85 and an 85 percent operating factor.
(g) Based on reference SAI84.
(h) Based on Table 6-2 and an 85 percent operating factor.
6-10
-------
The emission rate estimates for the idle plants are those
that would occur if the plants were to resume operation. These
values were used to estimate the radiation dose equivalents and
fatal cancer risks from the plants.
The risk assessment is based upon the emissions from the
calciner stacks, since earlier studies have shown that over 95
percent of the lead-210 and polonium-210 are emitted in the
calciner off-gases (see Section 6.2.1.1.1). The sources of the
information used to estimate the annual emissions from each
facility are listed in the footnotes to Table 6-9. Where
available, actual measurements were used. The source terms for
uranium-234, thorium-230, and radium-226 were assumed to be equal
to uranium-238, since measurements at these facilities have shown
these radionuclides to approximate secular equilibrium in
calciner off-gases (see Table 6-2). Because it is unlikely that
the.idle facilities will ever operate again, they are listed
separately from the operating facilities.
Lung-clearance classifications and particle size
distributions (AMAD) used in this assessment (ICRP Task Group
Lung Model) are shown in Table 6-10. These values are the same
as those used in the previous assessment (EPA84a).
6.2.1.3 Other Parameters Used in the Assessment
The effluent from calciner stacks normally has a significant
heat content that can result in substantial buoyant plume rise.
Table 6-11 lists the stack parameters that were used for each of
the eight elemental phosphorus plants. However, because of the
low heat content of emissions at the Stauffer plant in Silver
Bow, Montana, plume rise is affected more by momentum than by
buoyancy.
Meteorological data used in the assessment come from nearby
weather stations. Population distributions used in the assess-
ment were generated by the computer code SECPOP using 1980 census
tract data. For FMC's Pocatello plant and Monsanto's Soda
Springs plant, these population data were augmented with actual
population distributions for the first 5 km. Table 6-12 shows
the number of people living within 80 km of these sites and the
source of the meteorological data used in the calculations.
The distance from each facility to the residence of the
maximum exposed individual is also listed in Table 6-12. The
locations of the individuals at the FMC and Monsanto facilities
in Idaho and the Stauffer facility in Montana were selected from
actual population distributions and confirmed by personal visits.
USGS topographic quadrangle maps were used to identify the nearby
residences at the Florida facilities, which were later verified
during a demographic survey. For the facilities located in
Tennessee, the individuals were placed at 1,500 m in the
predominant wind direction from the facilities.
6-11
-------
Appendix A provides details of the input parameters supplied
to the assessment codes.
Table 6-10. Lung clearance classification and particle sizes
used in the assessment.
Clearance Particle Size
Radionuclide Classification AMAD
Pb-210, Po-210 y(a) 0.3(a)
U-238, U-234, Th-230 y(b)
Ra-226 W(b)
(a) Based on experimental data obtained during emission testing.
(b) Based on values recommended by ICRP (ICRP66) when
experimental values are not available.
Table 6-11. Calciner stack emission characteristics.
Stack Height Heat Emission
Plant (meters) (calories/sec)
FMC, Pocatello, ID
Monsanto, Soda Springs, ID
Stauffer, Silver Bow, MT
Stauffer, Mt. Pleasant, TN
Occidental, Columbia, TN
Stauffer, Tarpon Springs, FL
Mobil, Pierce, FL
31
27
27
35
31
49
26
29
Operating Plants
9.5E+5
5.0E+5
3.0E+4(a)
6.0E+5
1.2E+6
Idle Plants
1.7E+5
2.3E+5
1.1E+5
Monsanto, Columbia, TN 35 l.OE+6
(a) Because of the low heat content, plume rise for the Stauffer,
Silver Bow, MT, plant was based on momentum rather than buoyancy
(see Appendix A).
6-12
-------
Table 6-12,
Plant
Populations within 80 km and distances to the
maximum exposed individuals of elemental phosphorus
plants with the source of meteorological data used
in dose equivalent and risk calculations.
Number of Distance to Source of
People Within Maximum Exposed Meteorological
80 km(a) Individual (m) Data(B)
Operatina Plants
FMC
Pocatello, ID
Monsanto
Soda Springs, ID
Stauffer
Silver Bow, MT
Stauffer
Mt. Pleasant, TN
Occidental
Columbia, TN
Stauffer
Tarpon Springs, FL
Mobil
Pierce, FL
Monsanto
170,
100,
71,
560,
920,
1,700,
1,800,
900,
000
000
000
000
000
000
000
000
1,
4,
2,
1,
1,
Idle
2,
1,
800
000
500
500
500
Plants
500
750
500
Pocatello
Pocatello
Butte, MT
Nashville
Nashville
, ID
, ID
, TN
, TN
Tampa , FL
Orlando,
Nashville
FL
, TN
Columbia, TN
(a) Based on 1980 Census.
(b) Data from National Climatic Center, Asheville, NC.
6.3 RESULTS OF THE EXPOSURE AND RISK ASSESSMENT
This section contains an assessment of the radiation
exposure and risk of cancer due to radionuclide emissions from
elemental phosphorus plants. The assessment addresses the
following specific topics:
1) dose equivalent rates to the maximum exposed individual
due to radioactive emissions from each facility;
6-13
-------
2) collective dose equivalent rates to the regional
population (the total number of people residing within
80 km) around each elemental phosphorus plant;
3) the lifetime fatal cancer risk to the maximum exposed
individual due to radioactive emissions from each plant;
and
4) the number of fatal cancers committed per year in the
regional population around each elemental phosphorus
plant.
The radiation dose equivalent rates and fatal cancer risks
due to radioactive emissions from elemental phosphorus plants
were estimated for the maximum exposed individual and the 80-km
regional population using AIRDOS-EPA (Mo79) and DARTAB (Be81)
codes. Input parameters to the codes are listed in Tables 6-9 to
6-12 and in Appendix A. The results for the idle and operating
facilities are listed separately, because it is doubtful that any
idle plant will ever reopen.
6.3.1 Radiation Dose Equivalent Rates
The dose equivalent rates to the maximum exposed individual
and the collective dose equivalent rates to the regional
population for each elemental phosphorus plant are listed in
Table 6-13 in order of decreasing rates. Only those organ dose
equivalents that contribute 10 percent or more to the risk are
listed. Except at the Mobil Chemical Company site near Pierce,
Florida, the lung is the only organ that met this criterion and
is possibly at significant risk. At the Pierce, Florida site,
about 12 percent of the risk results from exposure to endosteal
bone from inhaling larger amounts of uranium-238, uranium-234,
and thorium-230 relative to lead-210 and polonium-210.
The largest dose equivalent rate (180 mrem/y) is estimated
to occur to the lung of the maximum exposed individual near
FMC's Pocatello, Idaho plant; the lowest exposed nearby
individual resides near the plant at Pierce, Florida, and receives
a lung dose of about 7 mrem/y. The locations of these maximum
exposed individuals in relation to the elemental phosphorus
plants are given in Table 6-12. The largest exposure to people
within an 80-km region (1,200 person-rem/year) is also estimated
to occur around FMC's Pocatello, Idaho, facility, while the lowest
collective dose equivalent rate (72 person-rem/year) is estimated
to be to the regional population around the Stauffer Chemical
Company plant at Mt. Pleasant, Tennessee. The populations exposed
within the 80-km regions are given in Table 6-12. The exposures
estimated from the three idle plants in Table 6-13 will not occur
if predictions are correct and the facilities fail to reopen.
6-14
-------
Table 6-13. Estimated radiation dose equivalent rates to the
maximum exposed individual and to the 80-km regional
population from elemental phosphorus plants.
Maximum Exposed Regional
Individuals Population
Plant Organ (mrem/y) (person-rem/y)
Operating Plants •
FMC Corporation Lung 180 1,200
Pocatello, ID
Monsanto Chemical Lung 34 80
Soda Springs, ID
Stauffer Chemical Lung 23 120
Silver Bow, MT
Stauffer Chemical Lung 14 72
Mt. Pleasant, TN
Occidental Chemical Lung 13 150
Columbia, TN
Idle Plants
Monsanto Chemical Lung 45 460
Columbia, TN
Stauffer Chemical Lung 7.3 530
Tarpon Springs, FL
Mobil Chemical
Pierce, FL
Lung
Endosteum
7.3
4.1
240
190
6.3.2 Health Risks
Table 6-14 lists the highest individual risk for each of the
five operating and three idle plants considered in this
assessment. The locations of these individuals in relation to the
elemental phosphorus plants are shown in Table 6-12 and discussed
in Section 6.2.1.3.
Ninety-nine percent of the risk is the result of inhaling
effluents from the elemental phosphorus plants, and, with the
exception of the Mobil plant in Pierce, FLorida, 95 percent of
the risk is due to lead-210 and polonium-210 in those effluents.
The highest lifetime individual risks occur at the operating FMC
and Monsanto plants in Idaho and are estimated to be 6 and nearly
1 fatal lung cancers in 10,000, respectively. The locations of
6-15
-------
these individuals were selected from actual population distribu-
tions and verified by personal visits during the demographic
survey (see Section 6.2.1.3).
The collective risks to the 80-km regional population around
each operating elemental phosphorus plant due to airborne
effluents from the calciners at these plants are also listed in
Table 6-14. The populations within these 80-km regions are
listed in Table 6-12. The largest risk is estimated to be to the
population of 170,000 around FMC's Pocatello, Idaho, facility.
The risk to this population is estimated to be about one cancer
every 20 years. The smallest collective risk to an operating
plant's regional population occurs at Soda Springs, ID (100,000
persons) and Mt. Pleasant, Tennessee (560,000 persons), and is
estimated to be about three deaths in 1,000 years.
The collective risks to the regional populations surrounding
the three idle elemental phosphorus plants are also given in
Table 6-14. These collective risks are estimated to be about one
death in each regional population every 100 years. These risks,
however, are nonexistent until one of the plants resumes
operation, which is very unlikely due to the decreased demand for
phosphorus and high operating costs (see Section 6.1.1).
The DARTAB computer code provides the frequency distribution
of lifetime fatal cancer risks for each elemental phosphorus
plant. It gives the number of people in each of a series of
lifetime risk intervals and the number of cancer deaths that
occur annually within each risk interval. This information is
summarized in Tables 6-15 and 6-16 for all operating and idle
elemental phosphorus plants, respectively. Again, data on the
idle facilities are included in the unlikely case that a plant
recommences operations. These data reflect the number of deaths
expected to occur annually within the 0-80 km populations, which
are listed in the second column. For example, 1,800,000 people
are at risk in the five regional populations due to their expo-
sure to the radioactive effluents from calciners at all operating
elemental phosphorus plants. Within that population, about one
fatal lung cancer is expected to occur every 15 years.
6-16
-------
Table 6-14
Plant
Estimated fatal cancer risks to the maximum exposed
individual and to the 80-km regional population from
elemental phosphorus plants.(a/
Individual Lifetime
Fatal Cancer Risk
Regional (0-80 km)
Population (deaths/y)
FMC Corporation
Pocatello, ID
Monsanto Chemical
Soda Springs, ID
Stauffer Chemical
Silver Bow, MT
Stauffer Chemical
Mt. Pleasant, TN
Occidental Chemical
Columbia, TN
Monsanto Chemical
Columbia, TN
Stauffer Chemical
Tarpon Springs, FL
Mobil Chemical
Pierce, FL
Operating Plants
6E-4
8E-5
6E-5
3E-5
3E-5
9E-5
1E-5
1E-5
Idle Plants
6E-2
3E-3
5E-3
3E-3
6E-3
1E-2
2E-2
7E-3
(a) Radon-222 emissions are not included in these estimates.
Previous assessments (EPA83) show that radon-222 from
calciners of elemental phosphorus plants add little
additional risk of fatal cancer (about 1 percent or
less of the total risk).
6-17
-------
Table 6-15.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from operating
elemental phosphorus plants.
Risk Interval
Totals
Number of Persons
1,800,000
Deaths/y
1E-1 to
1E-2 to
1E-3 to
1E-4 to
1E-5 to
1E-6 to
< 1E-6
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
5,
110,
250,
1,500,
0
0
0
000
000
000
000
0
0
0
1E-2
4E-2
2E-2
6E-3
7E-2
Table 6-16.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from idle
elemental phosphorus plants.
Risk Interval
Totals
Number of Persons
4,400,000
Deaths/y
1E-1 to
1E-2 to
1E-3 to
1E-4 to
1E-5 to
1E-6 to
< 1E-6
1E+0
1E-1
1E-2
1E-3
1E-4
1E-5
6
490
3 , 900
0
0
0
0
,800
,000
,000
0
0
0
0
1E-3
1E-2
2E-2
4E-2
6.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS
The results of analyses to determine the efficiencies of
various alternatives for controlling the polonium-210 and lead-210
emissions from calciner off-gas systems at the five operating
elemental phosphorus plants are summarized in Tables 6-17 and
6-18, respectively. The control alternatives considered were the
installation of wet (venturi) scrubbers, electrostatic
precipitators, a spray dryer followed by a fabric filter, and
HEPA (high efficiency particulate air) filters. A detailed
description of the analyses of these control alternatives and
their efficiencies is presented in EPA88c.
The capital costs estimated to implement the control
alternatives and the annualized costs (in 1988 dollars) are
presented in Tables 6-19 and 6-20, respectively. Detailed
6-18
-------
analyses of the costs and risk reduction, as well as the economic
impact, of alternative polonium-210 and lead-210 emission rates
for the five operating facilities are presented in EPA88c.
Table 6-17. Estimated Po-210 emission levels achieved by control
alternatives.
Emission Levels (Ci/y)
Control Stauffer
Alternative FMC Monsanto Montana Tennessee Occidental
Baseline
emissions(a)
10
30
2.4
0.28
0.31
Wet scrubber
AP=2.5 kPa(b) 8.0 21 1.7 0.20 0.22
AP=6.2 kPa 4.0 14 1.1 0.13 0.14
AP=10 kPa 2.0 3.0 0.24 0.028 0.031
AP=20 kPa 1.0 1.5 0.12 0.014 0.016
ESP(C)
200 SCA(d) 2.9 7.4 0.59 0.07 0.08
400 SCA 1.0 2.7 0.19 0.02 0.02
600 SCA 0.38 0.84 0.07 0.01 0.01
800 SCA 0.14 0.29 0.02 <0.01 <0.01
Spray dryer/
fabric filter
HEPA filter
0.043
<0.001
0.15
<0.001
0.012
<0.001
0.001
<0.001
0.002
<0.001
(a) Emissions with only low energy or spray scrubber. Additional
systems are added to these wet scrubbers except for spray
dryer/fabric filter.
(b) kPa - kilopascal which equals 4 inches of water.
(c) ESP - electrostatic precipitator.
(d) SCA - specific collection area in ft2/1000 acfm.
6-19
-------
Table 6-18.
Estimated Pb-210 emission levels achieved by control
alternatives.
Control
Alternative
Emission Levels (mCi/y)
Stauffer
FMC Monsanto Montana Tennessee Occidental
Base Line
emissions(a) 140 9,500 320
Wet scrubber
AP=2.5 kPa(b) 70 6,600 220
AP=6.2 kPa 28 2,800 96
AP=10 kPa 9.8 950 32
AP=20 kPa 5.6 480 16
58
41
17
5.8
2.9
64
45
19
6.4
3.2
ESP(C)
200 SCA(d)
400 SCA
600 SCA
800 SCA
25
8.0
2.8
1.0
2,500
840
290
100
85
28
9.6
3.5
15
5.1
1.7
0.64
17
5.6
1.9
0.70
Spray dryer/
fabric filter
0.60
49
1.6
0.29
0.32
HEPA filter 0.003 0.19 <0.01 <0.01 <0.01
(a) Emissions with only low energy or spray scrubber. Additional
systems are added to these wet scrubbers except for spray
dryer/fabric filter.
(b) kPa - kilopascal which equals 4 inches of water.
(c) ESP - electrostatic precipitator.
(d) SCA - specific collection area in ft2/1000 acfm.
6-20
-------
Table 6-19. Capital cost of control alternatives (1,000 1988 $).
Plant
Control
Alternative
Stauffer
FMC
Monsanto Montana Tennessee Occidental
Wet scrubber
10
25
40
80
inch
inch
inch
inch
AP
AP
AP
5,
7,
8,
13,
940
810
500
280
2,
3,
4,
6,
530
200
460
590
I,
If
3,
690
690
890
870
If
If
2,
5,
460
870
460
230
2,020
2,510
3,230
6,120
Electrostatic precipitator
200 SCA(b)
400 SCA
600 SCA
800 SCA
Spray dryer/
fabric filter
HEPA filtration
10,640
15,500
20,280
24,790
17,330
4,200
6,630
9,860
12,890
15,720
10,380
2,870
2,350
3,310
4,080
4,750
7,540
620
3,140
4,390
5,950
7,390
6,580
1,020
(a) 1 inch of water = 0.25 kPa.
(b) SCA - specific collection area in ft2/iQOO acfm.
4,530
6,500
8,600
11,340
10,060
1,610
6-21
-------
Table 6-20.
Control
Alternative
Annualized cost of control alternatives
(1,000 1988 $).
Plant
FMC
Stauffer
Monsanto Montana Tennessee Occidental
Wet scrubber
10 inch Ap(a) 1,600
25 inch AP 2,110
40 inch AP 2,430
80 inch AP 3,750
Electrostatic precipitator
200 SCA(b)
400 SCA
600 SCA
800 SCA
Spray dryer/
fabric filter
HEPA filtration
2,010
2,840
3,650
4,430
9,700
10,140
970
1,200
1,530
2,220
1,260
1,820
2,330
2,820
5,430
15,700
660
680
740
1,110
790
830
870
910
3,070
2,960
590
750
930
1,610
640
850
1,120
1,370
3,120
7,450
740
920
1,150
1,910
970
1,320
1,670
2,030
4,630
10,070
(a) 1 inch of water =0.25 kPa.
(b) SCA - specific collection area in ft2/1000 acfm.
6-22
-------
6.5 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, NV, 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, NV, August 1981.
Be81 Begovich, C.L.; Eckerman, K.F.; Schlatter, E.G.; Ohr,
S.Y.; and Chester, R.O.; "DARTAB: A Program to Combine
Airborne Radionuclide Environmental Exposure Data with
Dosimetric and Health Effects Data to Generate
Tabulations of Predicted Health Impacts," ORNL-5692, Oak
Ridge National Laboratory, Oak Ridge, TN, August 1981.
BM88 U.S. Bureau of Mines, "Mineral Industry Surveys,
Phosphate Rock," Bureau of Mines, Washington, DC,
January 4, 1988.
Bu85 Buttrey, C.W., Occidental Chemical Co., Columbia, TN,
written communication to Winston Smith, EPA, Washington,
DC, March 29, 1985.
EPA77 U.S. Environmental Protection Agency, "Radiological
Surveys of Idaho Phosphate Ore Processing - The Thermal
Plant," ORP/LV-77-3, EPA, Office of Radiation Programs,
Las Vegas, NV, 1977.
EPA83 U.S. Environmental Protection Agency, "Draft Background
Information Document, Proposed Standards for
Radionuclides," EPA 520/1-83-001, EPA, Office of
Radiation Programs, Washington, DC, March 1983.
EPA84a U.S. Environmental Protection Agency, "Radionuclides:
Background Information Document for Final Rule," Volume
II, EPA 520/1-84-022-2, EPA, Office of Radiation
Programs, Washington, DC, October 1984.
EPA84b U.S. Environmental Protection Agency, "Regulatory Impact
Analysis of Emission Standards for Elemental Phosphorus
Plants," EPA 520/1-84-025, EPA, Office of Radiation
Programs, Washington, DC, October 1984.
EPA84c U.S. Environmental Protection Agency, "Emissions of
Lead-210 and Polonium-210 from Calciners at Elemental
Phosphorus Plants: FMC Plant, Pocatello, Idaho," EPA,
Office of Radiation Programs, Washington, DC, June 1984.
6-23
-------
EPA84d U.S. Environmental Protection Agency, "Emissions of
Lead-210 and Polonium-210 from Calciners at Elemental
Phosphorus Plants: Stauffer Plant, Silver Bow, Montana,"
EPA, Office of Radiation Programs, Washington, DC, August
1984.
EPA84e U.S. Environmental Protection Agency, "Emissions of
Lead-210 and Polonium-210 from Calciners at Elemental
Phosphorus Plants: Monsanto Plant, Soda Springs, Idaho,"
EPA, Office of Radiation Programs, Washington, DC,
August 1984.
EPA88a U.S. Environmental Protection Agency, "Elemental
Phosphorus Production - Calciner Off-gases: Final
Emission Test Report, FMC Elemental Phosphorus Plant,
Pocatello, Idaho," EMB Report No. 88-EPP-02, January
1989.
EPA88b U.S. Environmental Protection Agency, "Elemental
Phosphorus Production - Calciner Off-gases: Final
Emission Test Report, Monsanto Elemental Phosphorus
Plant, Soda Springs, Idaho," EMB Report No. 88-EPP-01,
January 1989.
EPA88c U.S. Environmental Protection Agency, "Characterization
and Control of Radionuclide Emissions from Elemental
Phosphorus Production," EPA 450/3-88-015, February 1989.
ICRP66 International Radiological Protection Commission Task
Group on Lung Dynamics, "Deposition and Retention Models
for Internal Dosimetry of Human Respiratory Tract,"
Health Physics 12:173-207, 1966.
Ka84 Kalkwarf, D.R., and Jackson, P.O., "Lung-Clearance
Classification of Radionuclides in Calcined Phosphate
Rock Dust," PNL-5221, Pacific Northwest Laboratories,
Richland, WA, August 1984.
Mo79 Moore, R.E.; Baes, C.F. Ill; McDowell-Boyer, L.M.;
Watson, A.P.; Hoffman, F.O.; Pleasant, J.C.; and Miller,
C.W.; "AIRDOS-EPA: A Computerized Methodology for
Estimating Environmental Concentrations and Dose to Man
from Airborne Releases of Radionuclides," EPA
520/1-79-009, Oak Ridge National Laboratory for U.S. EPA,
Office of Radiation Programs, Washington, DC, December
1979.
Ra84a Radian Corporation, "Emission Testing of Calciner Off-
gases at FMC Elemental Phosphorus Plant, Pocatello,
Idaho," Volumes I and II, prepared for the Environmental
Protection Agency under Contract No. 68-02-3174, Work
Assignment No. 131, Radian Corporation, P.O. Box 13000,
Research Triangle Park, NC, 1984.
6-24
-------
Ra84b Radian Corporation, "Emission Testing of Calciner Off-
gases at Stauffer Elemental Phosphorus Plant, Silver Bow,
Montana," Volumes I and II, prepared for the
Environmental Protection Agency under Contract No.
68-02-3174, Work Assignment No. 132, Radian Corporation,
P.O. Box 13000, Research Triangle Park, NC, 1984.
Ra84c Radian Corporation, "Emission Testing of Calciner Off-
gases at Monsanto Elemental Phosphorus Plant, Soda
Springs, Idaho," Volumes I and II, prepared for the
Environmental Protection Agency under Contract No.
68-02-3174, Work Assignment No. 133, Radian Corporation,
P.O. Box 13000, Research Triangle Park, NC, 1984.
SAI84 Science Applications, Inc., "Airborne Emission Control
Technology for the Elemental Phosphorus Industry," Final
Report to the Environmental Protection Agency, prepared
under Contract No. 88-01-6429, SAI, P.O. Box 2351,
La Jolla, CA, January 1984.
6-25
-------
-------
7. COAL-FIRED UTILITY AND INDUSTRIAL BOILERS
7.1 INTRODUCTION
The coal-fired boiler source category includes utility and
industrial boilers. Approximately 1,200 utility boilers burn
coal to generate electricity, while more than 50,000 industrial
boilers burn coal to provide electricity, process heat, and space
heat for in-house use. These two classes of facilities account
for approximately 90 percent of the coal burned in the United
States. The remaining 10 percent is consumed by residential and
commercial boilers used for space and hot water heating.
Coal contains trace quantities of natural uranium and
thorium. Isotopes of uranium and thorium and their decay
products are released to the air with the particulate matter in
fly ash. There are no Federal or state regulations that directly
limit emissions of radionuclides from coal-fired utility or
industrial boilers. However, since radionuclide emissions are
directly related to particulate emissions, regulations and
standards limiting particulate releases indirectly limit
radionuclide releases as well. The Federal Clean Air Act (the
Act) sets ambient air quality standards for several pollutants
emitted by coal-burning facilities. These ambient standards
limit emissions of sulfur dioxide, oxides of nitrogen, carbon
monoxide, lead, and particulate matter 10 microns or less in
diameter (40 CFR 50.6, 50.7, 50.8, 50.11, 50.12). In addition to
ambient air standards, the Act also establishes new source
performance and prevention of significant deterioration
standards. For particulate matter, the limits and standards
include:
The PM-10 Standard; Particulate matter 10 microns or
less in diameter emitted from a coal-burning facility
may not result in ambient levels of such particles in
excess of 150 ug/m3 in more than one 24-hour period per
year, or in excess of an annual average of 50 ug/m3.
Prevention of Significant Deterioration (PSD); PM-10
emissions from a coal-burning facility may not result
in an increase in ambient PM-10 levels of 10 ug/m3
24-hour maximum or 5 ug/m3 annual average in Class I
areas, and 37 ug/m3 24-hour maximum or 19 ug/m3 annual
average in Class II areas.
New Source Performance Standards; All new coal-fired
boilers with capacities greater than 73 MW thermal
input are subject to a particulate emission limit of
43.3 ng/J (0.10 Ib/million BTU) heat input, and new
utility coal-fired boilers of this size are limited to
13 ng/J (0.03 Ib/million BTU) heat input. New boilers
with capacities less than 73 MW are subject to limits
prescribed by State Air Quality Implementation Plans.
7-1
-------
The states (or local air quality control regions) set
emission standards for existing sources as part of the State Air
Quality Implementation Plans (SIPs). The SIPs are developed to
assure compliance with Federal ambient air quality and prevention
of significant deterioration standards.
7.1.1 Coal Use in the United States
In 1982, approximately 20 percent of U.S. energy
needs were met by burning coal: 74 percent to generate
electricity and about 24 percent for industrial use (DOE85). In
1982, combustion of coal at utility and industrial boilers
accounted for approximately 15,000 X 1012 BTU heat input. The
utility boilers consumed approximately 85 percent of the total
(12,500 X 1012 BTU), and the industrial sector consumed
approximately 15 percent (2,500 X 1012 BTU) (Me86). In utility
and .industrial applications, bituminous, sub-bituminous, and
lignite coals are much more widely used than anthracite.
Although natural gas, oil, and nuclear fission can be used
to generate electricity thermally, the cumulative use of these
energy sources has decreased in recent years. Indigenous natural
gas supplies have been tapped heavily, and most natural gas in
the United States is used for space heating, other residential
heating applications, and as a petrochemical and fertilizer
source. It is expected that coal will supply more than half of
the electricity generated in the United States in the foreseeable
future.
7.1.2 Radionuclides in Coal
The mineral matter contained in coal includes small
quantities of naturally-occurring uranium and thorium and their
decay products. Tables 7-1 and 7-2 present the half-lives and
principal radiations of the major decay products of uranium-238
and thorium-232, respectively. Data showing typical uranium and
thorium concentrations in coal are presented in Table 7-3 by
region and coal rank. The values presented for "All Coals" at
the end of the table represent more than 5,000 coal samples from
all major production areas in the United States. The
distribution of uranium concentrations in coal presented in
Table 7-4 indicates that 98 percent of all coals have uranium
concentrations of 10 ppm or less.
The release rates of uranium and thorium and their decay
products depend on their initial concentrations in the coal, the
ash content of the coal, and boiler-specific factors including
furnace design, heat rate, and effluent control system
efficiency. In this assessment, the values of 1.3 ppm
uranium and 3.2 ppm thorium (representing the geometric mean for
all coals) and an average value of 10 percent ash are used in
conjunction with boiler-specific emission factors.
7-2
-------
Table 7-1. Major decay products of uranium-238.
Radionuclide Half-life Principal radiation (Mev)
Uranium-238
Thorium-234
4.5 x 109 y
24 ,d
Protact inium-2 3 4m 1.2 m
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
y = years, d
Source: Le67
2.5 x 105 y
8.0 x 104 y
1.6 x 103 y
3.8 d
3.1 m
27 m
20 m
1.6 x 10"4 s
22 y
5.0 d
138 d
= days, h = hours,
Alpha Max. Beta
4.20
0.191
2.29
4.77
4.68
4.78
5.49
6.00
0.65
1.51
7.69
0.015
1.160
5.31
m - minutes, s - seconds
Gamma
0.093
1.001
0.186
0.352
0.609
0.047
Table 7-2. Major decay products of thorium-232.
Radionuclide Half-life Principal radiation (Mev)
Alpha
Thorium-232
Radium-228
Act inium-2 2 8
Thorium-228
Radium-224
Radon-220
Polonium-216
Lead-212
Bismuth-212
1
6
6
1
3
55
0
10
60
.4
.7
.1
.9
.6
s
.15
h
m
x 1010 y
y
h
y
d
s
4.
5.
5.
6.
6.
01
43
68
29
78
Max . Beta Gamma
0.055
1.11 0.
0.
0.
0.589 0.
2.25 0.
908
084
241
239
727
Polonium-212 3.1 x 10~7 s 8.78
Thallium-208 3.1m 1.80 2.614
y = years, d = days, h = hours, m = minutes, s = seconds
Source: Le67
7-3
-------
Table 7-3
Region/
Coal Rank
Typical uranium and thorium concentrations in coal.
Uranium Thorium
Range Geometric Range Geometric
mean mean Refer-
(ppm) (ppm) (ppm) (ppm) ence
Pennsylvania
Anthracite
Appalachian
Bituminous
NR
Bituminous
Bituminous
Illinois Basin
NR
Bituminous
Bituminous
0.3-25 1.2 2.8-1.4 4.7 Sw76
<0.2 - 11 1.0 2 - 48 2.8 Sw76
0.4-3 1.3 1.8-9 4.0 IGS77
NR 1.1 NR 2.0 SRI77
0.1-19 1.2 NR 3.1 ZU79
0.3- 5 1.3 0.7- 0.5 1.9 IGS77
0.2 - 43 1.4 <3 - 79 1.6 Sw76
0.2 - 59 1.7 0.1 - 79 3 Zu79
Northern Great Plains
Bituminous-
Sub-bituminous <0.2
Sub-bituminous <0.1
Lignite 0.2
3 0.7 <2 -8 2.4 SW76
16 1.0 0.1 - 42 3.2 ZU79
13 1.2 0.3 - 14 2.3 ZU79
Western
NR
Rocky Mountain
Bituminous-
Sub-bituminous
Sub-bituminous
Bituminous
All Coals
0.3-3 1.0 0.6-6 2.3 IGS77
0.2 - 24 0.8 <3 - 35 2.0 Sw76
0.1 - 76 1.9 0.1 - 54 4.4 Zu79
0.1 - 42 1.4 <0.2 - 18 3.0 Zu79
<0.1 - 76 1.3 <0. 1- 79 3.2
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.
NR Not reported
7-4
-------
Table 7-4. Uranium concentrations and distributions in coal.
Uranium Number Percent of Coals
Concentration of Coals Within Uranium
(ppm) Analyzed Concentration
Range
less than 2 2,669 71.5
2-4 666 17.9
4-6 207 5.5
6-8 67 1.8
8-10 39 1.0
10 - 12 26 0.7
12-14 17 0.5
14 - 16 12 0.3
16-18 7 0.2
18-20 5 0.1
20-30 9 0.2
30-60 5 0.1
60 -130 2 0.05
Source: Fa79
7.2 UTILITY BOILERS
7.2.1 General Description
7.2.1.1 Profile of Utility Boilers
In 1985, 2.47 trillion kilowatt-hours of electricity were
generated in the United States (WA87) of which 56.8 percent was
generated by burning coal. In 1986, there were approximately
1,200 coal-fired utility boilers in the United States, with a net
generating capacity of 305 GW (DOES6).
A few terms commonly used in discussions of electric
generation are:
"Capacity factor" (often referred to as "capacity") is
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.
"Availability" 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.
7-5
-------
"Capability" is the percentage of nameplate capacity
which is needed to meet an average seasonal demand;
this term is beginning to replace "capacity factor" as
a hallmark of plant operation.
Power plants are designed and operated to serve three load
classes:
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).
Intermediate-load (or cycling) plants, which operate at
varying levels of capacity each day (about 40 percent
utilization on an average annual basis).
Peaking plants, which operate only a few hours per day
(about 700-800 hours per year).
Fossil-fueled steam-electric generating plants now dominate
base-load and intermediate-load service. Coal is rarely the
primary fuel for a peaking plant. New units have historically
been used for base-load generation; cycling capacity has been
obtained by downgrading the older, less efficient base-load
equipment as more replacement capacity comes on line.
In 1979, the average capacity factor for coal-fired units
operating in the base-load mode was 65 percent; for units
operating in a cycling mode, 42 percent (TRI79). The
availability of a coal-fired unit generally declines with
increasing generating capacity. Generating units with capacities
of less than 400 MW have average availabilities of more than 85
percent; those with capacities of more than 500 MW, only 74 to 76
percent (An77). The operating mode affects the heat rate of the
plant; for example, changing the capacity factor from 42 to 70
percent changes the heat rate from 12.3 to 9.2 MJ/kWh.
7.2.1.2 Process Description
As coal is burned, the minerals in the coal melt and then
condense into a glass-like ash; the quantity of ash depends on
the mineral content of the coal. A portion of the ash settles to
the bottom of the boiler (bottom ash), and the remainder enters
the flue gas stream (fly ash). Partitioning between fly ash and
bottom ash for various types of coals and various boiler designs
is given in Table 7-5 (Me86).
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). Fuel-
firing equipment can be divided into three general categories:
stoker furnace (dry bottom), either spreader or non-spreader;
cyclone furnace (wet bottom); and pulverized-coal furnace (dry or
wet bottom).
7-6
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Table 7-5. Coal ash distribution by boiler type.
Percent Fly Ash/Percent Bottom Ash
Furnace Type Bituminous Lignite Anthracite
Pulverized dry bottom
Pulverized wet bottom
Cyclone
Stoker
80/20
65/35
13.5/86.5
60/40
35/65
-
30/70
35/65
85/15
-
-
5/95
Stoker furnaces are usually small, older boilers ranging in
thermal capacity from 7.3 to 73 MW. Of the coal-fired boilers
sold from 1965 to 1973, none exceeded 143 MW (thermal); 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.
Cyclones are high-temperature combustion chambers for
burning crushed coal. 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. As of 1974, only 9 percent of the coal-fired utility
boiler capacity was of the cyclone type, and no boilers of this
kind have been ordered by utilities in the past seven years
(Co75).
A pulverized-coal furnace burns coal which has been
pulverized to a fine powder (approximately 200 mesh) and which is
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).
The dry-bottom, pulverized-coal-fired boiler, in which the
furnace temperature is kept low enough to prevent the ash from
melting, 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 a slurry, which is transported to an ash-settling
pond.
7-7
-------
The distribution of utility coal-fired boiler types, by
percent, is:
Pulverized dry bottom: 76%
Pulverized wet bottom: 11%
Cyclone: 11%
Stoker: 2%
The use of fluidized bed combustors, which generally have
lower air emissions, continues to increase. In addition, the
Clean Coal Project of the U.S. Department of Energy is developing
technology for burning a mixture of coal and liquid fuel derived
from coal which should considerably reduce fly ash (Tr81).
Incorporation of clean coal technology into coal combustion uses
is expected to accelerate, but an accurate prediction of the rate
of acceleration is not now possible.
7.2.1.3 Current Status of Emission Control
As was noted in the introduction, the National Ambient Air
Quality Standards (NAAQS) require air emission controls for
virtually all coal-fired utility boilers in the United States.
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 a number of publications (for example, De77,
De79, Co77).
ESPs, wet scrubbers, and fabric filters are all theoretically
capable of better than 99.8 percent collection efficiencies for
ash as small as one micron in diameter. However, actual
collection efficiency for a specific unit can be considerably
less (as low as 50 percent) because of specific loading
parameters and ash characteristics. Operational collection
efficiencies of ESPs and fabric filters, in particular, have
improved during the last decade, so that, at present, almost all
collectors are at least 98 percent efficient during normal
operation. Hot-side precipitators have been developed to
overcome problems posed by resistive fly ash. The recent
development of high-temperature fabrics has resulted in an
increase in the use of fabric filters for controlling utility
boiler emissions.
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 SC>2 scrubber systems in a plant depends
on the type of scrubbers (wet or dry) installed; these devices
are located upstream of a wet scrubber system or downstream of a
spray dryer system.
7-8
-------
Table 7-6 gives the distribution of particulate control
equipment for utility boilers burning bituminous coal; this
distribution is representative of control equipment on boilers
using other types of coal.
Table 7-6. Distribution of particulate control equipment for
bituminous coal-fired utility boilers.
% Distribution of Particulate Control Equipment
Combustion ESP Centrifugal Other No
System Separator Control
Pulverized dry bottom
Number basis 60 17 15 8
Capacity basis 79 10 10 1.6
Fuel consumption basis 83 11 5 1.0
Pulverized wet bottom
Number basis 52 20 16 11
Capacity basis 66 11 9 14
Fuel consumption basis 77 9 7 7
Cyclone
Number basis 61 5 18 7
Capacity basis 83 8 5 4
Fuel consumption basis 89 5 33
Stoker
Number basis 8 36 25 32
Capacity basis 29 32 20 19
Fuel consumption basis 44 25 14 16
Source: Me86
7.2.2 Basis for the Risk Assessment of Utility Boilers
The risk assessment of utility boilers is based on
reference (actual) facilities selected to represent large and
typical utility boilers. The reference facilities were selected
from a data base of almost one thousand utility boilers
maintained by the EPA's Office of Air Quality Planning and
Standards (OAQPS). The boilers in the data base account for
virtually all of the coal used by utility boilers.
7-9
-------
In selecting the reference utility boilers, the boilers in
the data base were classified according to the number of persons
living within 50 km of the .plant. Urban plants were defined as
3,000,000 persons or more, suburban plants as 800,000 to
3,000,000 persons, rural plants as 100,000 to 800,000 persons,
and remote plants as less than 100,000 persons. This
classification shows 34 utility boilers located in urban areas,
234 located in suburban areas, 567 located in rural areas, and
150 located in remote areas.
For each location, the large reference plant and the typical
reference plant were chosen based on the estimate of total
particulate emissions. The large reference plants were used in
the evaluation of the risks to nearby individuals and the typical
reference plants were used to evaluate the magnitude and
distribution of the population risk.
7.2.2.1 Radionuclide Emissions
The trace amounts of uranium-238, thorium-232, and their
decay products present in coal are released to the atmosphere as
particulates in the fly ash. The quantities emitted depend on
the concentrations of the radionuclides in the coal burned, the
type of boiler and emissions controls operating, the capacity,
capacity factor, and heat rate for the boiler operation, and the
ash partitioning. The distribution of ash between the bottom and
fly ash depends on the firing method, the type of coal, and the
type of furnace (dry bottom or wet bottom). For pulverized-coal,
dry-bottom units, 80-85 percent of the ash is fly ash.
Measured emission factors for uranium-238 and thorium-232,
on both a weight and heat input basis, are given for various
types of boilers and control systems in Tables 7-7 and 7-8
(Me86). Although uranium and thorium are in secular equilibrium
with their progeny in coal, measurements have shown that certain
radionuclides are partitioned unequally between the bottom ash
and fly ash (Be78, Wa82). The concentration mechanism is not
fully understood; however, one explanation is that certain
elements are preferentially concentrated on the particle
surfaces, resulting in their depletion in the bottom ash and
their enrichment in the fly ash (Sm80).
The highest concentration of the trace elements in fly ash
is found in particulates in the 0.5 to 10 micron 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. Uranium is enriched in the fly
ash relative to the bottom ash, particularly in particles less
than 1 micron in diameter. The enrichment factor for uranium is
about 2. Thorium, on the other hand, shows virtually no small
particle enrichment and is only slightly enriched in the fly ash.
Enrichment factors based on measured values obtained at utility
boilers are shown in Table 7-9 for the radioisotopes in coal that
may present a health risk (EPA81).
7-10
-------
The total amount of uranium released from all utility
boilers can be estimated using the average uranium content of
coal (1.3 ppm), the average ash content of coal (10 percent), an
enrichment factor of 2, and the total quantity of particulate
matter released from utility boilers. The OAQPS estimates the
total quantity of suspendible particulate matter from all the
utility boilers in its data base to be 3 X 108 kg/y (EPA89).
Using this value, an estimated 3 Ci/y of uranium-238 is released
from all utility boilers.
Table 7-7. U-238 emission factors for coal-fired utility boilers.
Boiler Type/ Emission Factor (pCi/q) Emission Factor(pCi/MBTU)
Control Average Range Average Range
Pulverized Dry Bottom
ESP 6.55
ESP/Scrubber 7.1
Scrubber 5.6
Pulverized Slag Bottom
Mechanical/ESP
Cyclone
ESP
Scrubber
Stoker
Fabric Filter
ESP
Unspecified
ESP
0.004
1.5
13.9
0.003
0.5
16.1
3.3-9.2
295.3
22.5
73.7
0.005- 3.0
0.017-37.5
68.0
1757.8
6.3-675.9
301.2-3214.3
7-34.2
294
101.6-486.5
MBTU means million BTU.
Source: Table 3-173 of Me86.
7-11
-------
Table 7-8. Th-232 emission factors for coal-fired utility boilers.
Boiler Type/ Emission Factor (pCi/q) Emission Factor(pCi/MBTU)
Control
Pulverized Dry
ESP
ESP/Scrubber
Scrubber
Cyclone
ESP
Scrubber
Stoker
ESP
Average Range
Bottom
3.0 0.6-5.3
7.14
2.78
1.8
2.09 1.5-2.68
0.5
Average Range
170.0 50.3-180.7
22.7
36.5
40.8
170.0 110.2-229.7
13.8
MBTU means million BTU.
Source: Table
3-174 of Me86.
Table 7-9. Enrichment factors for radionuclides.
Nuclide
Enrichment Factor
Uranium series
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-222
Lead-210
Polonium-210
Thorium series
Thorium-232
Radium-228
Thorium-228
Radium-224
Radon-220
Lead-212
Bismuth-212
2
2
1
1.5
20
5
5
1
1.5
1
1.5
20
5
5
7-12
-------
7.2.2.2 Source Terms Used in the Risk Assessment
Source terms for the reference facilities were developed by
using the plant specific data in the OAQPS data base on boiler
types, heat input, and control systems. For each reference
plant, the average emission factors (in pCi/MBTU) from Tables 7-7
and 7-8 for the appropriate boiler type and control technique
were multiplied by the heat input (MBTU/y) to yield the
uranium-238 and thorium-232 source terms. Source terms for the
decay products were determined using the enrichment factors
presented in Table 7-9. The estimated uranium-238 and
thorium-232 source terms for the typical and large reference
boilers are presented in Tables 7-10 and 7-11, respectively.
7.2.2.3 Other Parameters Used in the Risk Assessment
The reference plants were assessed using site-specific data
for each plant. Releases were modeled using actual stack heights
and buoyant plume rise calculated on the basis of the units' heat
inputs and capacity factors. Meteorological data from nearby
airports were used, and the 0-80 km population distributions were
generated using the SECPOP computer code. Risks to nearby
individuals were assessed by assuming that individuals reside in
the predominant wind direction at a distance of 750 meters from
the plant.
Food fractions appropriate to the type of location were
assumed. Details of the parameters input to the assessment codes
are presented in Appendix A.
Table 7-10. Emissions for typical coal-fired utility boilers.
Facility
Remote
Rural
Suburban
Urban
U-238
(mCi/y)
5.6
5.6
9.4
5.1
Th-232
(mCi/y)
3.2
2.3
5.4
2.4
7-13
-------
Table 7-11. Emissions for large coal-fired utility boilers.
Facility
Remote
Rural
Suburban
Urban
U-238
(mCi/y)
32
42
40
39
Th-232
(mci/y)
19
25
24
22
7.2.3 Results of the Dose and Risk Assessment of Utility Boilers
7.2.3.1 Estimated Doses from Utility Boilers
The estimated dose rates for both the nearby individuals and
the regional population are presented in Table 7-12 for typical
utility boilers, and in Table 7-13 for large boilers. Organ dose
rates that represent 10 percent or more of the total risk are
reported.
7.2.3.2 Estimated Risks from Utility Boilers
The estimated lifetime fatal cancer risk to nearby
individuals and the estimated risk to the regional population are
given in Tables 7-14 and 7-15. The greatest lifetime fatal
cancer risk estimated is 3E-5. This estimate, obtained for the
large reference utility boiler in a rural location, reflects the
risk that could occur at the location of maximum offsite dose and
presumes that a large fraction of the foodstuffs consumed by the
individual are grown at that location.
7-14
-------
Table 7-12,
Facility
Estimated radiation dose rates from typical coal-
fired utility boilers.
Organ
Suburban
Nearby
Individuals
(mrem/y)
Gonads
Breast
Lung
Red Marrow
Remainder
Bone Surface
1.5E-1
1.4E-1
1.3E-1
1.1E-1
1.1E-1
Regional
Population
(person-rem/y)
Remote
Rural
Gonads
Breast
Remainder
Red Marrow
Lung
Bone Surface
Bone Surface
Remainder
Red Marrow
Gonads
Lung
1.6E-1
1.5E-1
1.3E-1
1.3E-1
1.2E-1
—
7.9E-1
1.2E-1
8.7E-2
4.7E-2
-
^m
-
1.5E+0
1.1E+0
1.8E+0
1.1E+1
1.2E+1
1.6E+0
1.2E+0
-
2.3E+0
6.1E+1
4.5E+0
5.9E+1
Urban
Lung
Gonads
Breast
Red Marrow
Remainder
Bone Surface
1.1E-1
8.7E-2
8.1E-2
7.0E-2
6.7E-2
1.6E+2
1.2E+2
7-15
-------
Table 7-13,
Facility
Estimated radiation dose rates from large coal-
fired utility boilers.
Organ
Nearby
Individuals
(mrem/y)
Regional
Population
(person-rem/y)
Remote
Bone Surface
Remainder
Gonads
Red Marrow
Lung
1.1E+0
3.1E-1
2.7E-1
2.7E-1
2.9E+1
4.4E+0
3.1E+0
1.6E+1
Rural
Bone Surface
Remainder
Red Marrow
Gonads
Lung
1.2E+1
2.1E+0
1.5E+0
l.OE+0
3.9E+1
5.6E+0
4.2E+0
2.0E+0
6.6E+0
Suburban
Urban
Gonads
Breast
Remainder
Red Marrow
Lung
Bone Surface
Gonads
Breast
Remainder
Red Marrow
Lung
Bone Surface
5.2E-1
4.9E-1
4.1E-1
4.0E-1
4.0E-1
—
3.5E-1
3.2E-1
2.7E-1
2.7E-1
2.6E-1
^m
5.3E+0
—
9.2E+0
7.9E+0
1.9E+1
5.9E+1
6.8E+0
—
9.6E+0
—
3.7E+1
6.5E+1
7-16
-------
Table 7-14. Estimated fatal cancer risk from typical coal-
fired utility boilers.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Remote 3E-6 2E-4
Rural 1E-6 2E-4
Suburban 3E-6 3E-3
Urban 2E-6 6E-3
Table 7-15. Estimated fatal cancer risk from large coal-
fired utility boilers.
Nearby Individuals Regional (0-80 km)
Facility Lifetime Fatal Population
Cancer Risk Deaths/y
Remote
Rural
Suburban
Urban
6E-6
3E-5
1E-5
7E-6
1E-3
9E-4
2E-3
3E-3
7.2.3.3 Projection of Fatal Cancer Risk to U.S. Population
The risks (deaths/year) and the distribution of the risks
estimated for the four typical reference utility boilers were
extrapolated to estimate the risk attributable to radionuclide
releases from all utility boilers. The extrapolation was made as
follows. First, the risk distribution for each of the four
typical reference facilities was multiplied by the number of
facilities in that population category (150 remote plants, 567
rural plants, 234 suburban plants, 34 urban plants). Next, the
distributions were summed for all four population categories.
The problem of overlap was addressed by limiting the population
at risk to the actual U.S. population. Finally, because the
emissions from the reference facilities are typical emissions and
not mathematical averages, a scaling factor had to be used so
that the risk being estimated for all plants corresponds to the
risk from the approximately 3 curies of uranium-238 that are
estimated to be emitted annually by all coal-fired utility
7-17
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boilers. The resulting distribution is presented in Table 7-16.
The total estimated number of deaths per year due to coal-fired
utility boiler radionuclide emissions is 0.4.
Table 7-16. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from all
coal-fired utility boilers.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
0
0
0
0
0
130,000
240,000,000
0
0
0
0
0
1E-3
4E-1
Totals 240,000,000 4E-1
The estimates of maximum individual risk and total deaths
per year obtained in this assessment agree closely with
the estimates made by OAQPS (EPA89). In making its estimates,
the OAQPS scales the risks estimated for a model plant, with
average stack characteristics, sited on typical urban, suburban,
and rural demographies to each of the plants in its data base.
The OAQPS uses two scaling factors. The first is the ratio of
the model plant's uranium-238 emissions to the estimated
uranium-238 emissions for each plant, calculated on the basis of
heat input and the appropriate boiler/control-type specific
emission factor. The second is the ratio of the population
within 50 km of the model plant to the actual population within
50 km of each plant.
7.2.4 Supplementary Control Options and Costs
Existing boilers can be retrofitted with additional electro-
static precipitators to reduce emissions to the level prescribed
for new sources (13 ng/J). With all coal-fired utility boilers
operating with particulate emissions of 13 ng/J (0.03 Ib/MBTU) of
heat input, the current 12,500 x 106 MBTU annual heat input would
result in about 1.7 x 108 kg of particulate releases. This is
roughly half of the current estimate of particulate releases.
The source term and potential health impact would therefore be
reduced by about a factor of 2. The estimate of the total
deaths per year would drop to 0.2.
7-18
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The EPA's Office of Air Quality Planning and Standards has
estimated the costs of retrofitting all existing utility coal-
fired boilers to meet the control level of 13 ng/J to be about
$13 billion in capital cost (1982 dollars) and about $3.4 billion
in annual costs (RC83).
7.3 INDUSTRIAL BOILERS
7.3.1 General Description
Coal-fired industrial boilers are used primarily to produce
process steam, generate electricity for the industrial producer's
own use onsite, and provide space and water heat. Boilers are
used in virtually every industry, from small manufacturing plants
to large concerns. Major users are smelters, steel, aluminum and
copper fabrication, pulp and paper manufacture, and the chemical
industry. In 1974, about 90 percent of the coal burned in
industrial boilers was consumed by the steel, aluminum, chemical,
and paper industries (EPA80). That fraction has not changed
materially.
7.3.1.1 Process Description
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 insides 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 more than 200 MW thermal input.
Fire tube boilers are designed to allow hot combustion gas
to flow through the tubes, while the water to be heated is
circulated outside the tubes. These boilers are usually smaller
than 9 MW thermal input.
Cast iron boilers are designed like fire tube boilers, with
heat transfer from hot gas inside the tubes to circulating water
outside the tubes, but cast iron is used rather than the steel
used in fire tube boilers. Cast iron has a lower heat capacity
and is a better conductor of heat than most steels. Cast iron
boilers generally have capacities of less than 3 MW.
Table 7-17 lists the approximate number of industrial
boilers in the United States, as of 1981, and their installed
capacities (EPA81). Water tube units represent 89 percent of the
total installed capacity in terms of heat input. Since the
amount of coal burned influences the level of emissions to the
environment, emissions from water tube boilers largely determine
the radiological impact of coal-fired industrial boilers.
7-19
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Table 7-17. Numbers and capacities of industrial boilers.
Unit Capacity (MW Thermal Input)
Boiler Type 0-3 3-15 15-30 30-75 >75
Water Tube Units
Total MW
Fire Tube Units
Total MW
Cast Iron Units
Total MW
683
835
8112
5650
35965
6330
2309
22225
1224
7780
1290 1181 423
27895 50825 59930
There are two main types of coal-fired industrial 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 feeds the coal into the furnace and provides a grate
upon which the coal is burned. Stokers are generally rated at
less than 120 MW heat input. The three main types of stoker
furnaces are spreader, overfeed (or chain grate), and underfeed.
Each of the boiler types is discussed below.
Pulverized Coal-Fired Boilers
Coal is pulverized to a light powder and pneumatically
injected through burners into the furnace. If the furnace is
designed to operate at a high temperature (typically 1,600" 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. "Dry bottom" units operate at lower
combustion temperatures (1,200 - 1,600° C) with the bottom ash
remaining in the solid state.
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 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.
7-20
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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.
7.3.1.2 Emissions and Emission Controls
7.3.1.2.1 Particulate Emissions by Boiler Type
The fractional distribution of ash between the bottom ash
and fly ash directly affects the particulate emission rate and is
a function of the following parameters:
Boiler firing method. The type of firing is the most
important factor in determining ash distribution.
Stoker-fired units emit less fly ash than pulverized
coal-fired boilers.
Wet or dry bottom furnaces. Dry bottom units produce
more fly ash.
Boiler load. Particulate emissions are directly
proportional to the amount (load) of coal burned.
7.3.1.2.2 Existing Control Technology
As in the case of utility boilers, radionuclides are emitted
with the particulates in the fly ash. The technologies commonly
used to remove particulates from effluent gas from coal-fired
industrial boilers are the same as those used on utility boilers
and have been discussed in a Section 7.2.1.3. However, unlike
the utility boilers, a large fraction of industrial boilers
operate without particulate emission controls or with low-
efficiency controls such as multiclones.
7.3.2 Basis for the Risk Assessment of Industrial Boilers
Characteristics of individual industrial boilers vary
considerably. The majority of these plants are very small, but
the larger plants have heat inputs comparable to those of utility
boilers. The risk assessment of industrial boilers is based on a
single reference plant. The reference plant has the largest
estimated release of total particulates of the industrial boilers
in OAQPS1 data base of about 500 industrial boilers (EPA89). The
boilers in the OAQPS data base represent a stratified random
sample of more than 2,000 industrial boilers located throughout
the United States.
The untypically large emissions from this plant, reflecting
its large heat input and relatively inefficient multiclone
7-21
-------
control system, provide a conservative estimate of the health
risks posed by radionuclide emissions from industrial boilers.
7.3.2.1 Radionuclide Emissions
Radionuclide release rates from industrial boilers have not
been measured. Therefore, the source term for the reference
facility is estimated using the actual heat input of the plant
and an emission factor derived for utility boilers. The annual
release of uranium-238 from this facility is estimated to be
8 mCi. The source term also includes 4 mCi/y of thorium-232.
Release rates of the uranium and thorium decay products are
estimated using the enrichment factors given in Table 7-9. This
is a conservative assumption, as these enrichment factors were
developed for utility boilers and probably overstate the amount
of polonium-210 and lead-210 actually released by industrial
boilers.
7.3.2.2 Other Parameters Used in the Risk Assessment
The reference plants were assessed using site-specific data
for each plant. Releases were modeled using actual stack height
and buoyant plume rise calculated on the basis of the unit's heat
input and capacity factor. Meteorological data from a nearby
airport were used, and the 0-80 km population distributions were
generated using the SECPOP computer code. Risks to nearby
individuals were assessed by assuming that individuals reside in
the predominant wind direction at a distance of 250 meters from
the plant.
As the reference facility is located in a rural area, food
fractions appropriate to a rural location were assumed. Details
of the parameters input to the assessment codes are presented in
Appendix A.
7.3.3 Results of Dose and Risk Assessment of Industrial Boilers
7.3.3.1 Estimated Doses and Risks from Industrial Boilers
The estimated dose rates from the large industrial reference
facility are presented in Table 7-18. Organ doses that represent
10 percent or more of the total risk are reported. The lifetime
fatal cancer risk for nearby individuals is estimated to be 7E-6.
This estimate reflects the risk that could occur at the location
of maximum offsite dose and presumes that a large fraction of the
foodstuffs consumed by the individual are grown at that location.
The radionuclide releases from the reference plant are estimated
to cause 1E-3 deaths/year in the regional population.
7-22
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Table 7-18. Estimated radiation dose rates from the reference
coal-fired industrial boiler.
Nearby Regional
Organ Individuals Population
(mrem/y) (person-rem/y)
Bone Surface
Remainder
Red Marrow
Lung
6.5E+0
9.0E-1
6.1E-1
-
5.6E+1
5.8E+0
—
2.1E+1
7.3.3.2 Distribution of the Fatal Cancer Risk
The magnitude and distribution of the fatal cancer risk
estimated for the reference facility were extrapolated to obtain
an estimate of the risk attributable to radionuclide releases
from all industrial boilers. It is estimated that the total
airborne release of uranium-238 from industrial coal-fired
boilers is about 3 Ci/y (EPA89). Using this estimate, the
results from the reference facility were scaled to obtain the
potential health impact of all industrial boilers. Table 7-19
presents the resulting distribution which indicates an estimated
0.4 deaths per year.
Table 7-19. Estimated distribution of the fatal cancer risk
to the regional (0-80 km) populations from all
coal-fired industrial boilers.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
<
to 1E+0
to 1E-1
to 1E-2
to 1E-3
to 1E-4
to 1E-5
1E-6
0
0
0
0
0
*
240,000,000
0
0
0
0
0
*
4E-1
Total 240,000,000 4E-1
The results of the risk assessment of the model facility
indicate that there may be individuals in this risk interval,
However, data are insufficient to provide quantitative
estimates.
7-23
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7.3.4 Supplementary Controls Options and Costs
A full evaluation of supplementary control options and costs
has not been performed for industrial boilers. Existing boilers
could be retrofitted with electrostatic precipitators (ESPs). It
is estimated that retrofitting ESPs at industrial boilers with
heat inputs >2 x 106 MBTU/hr would reduce particulate emissions
by a factor of approximately 2.
7-24
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7.4 REFERENCES
An77 Anson, D., Availability of Fossil-Fired Steam Plants.
EPRI-FP-422 SR, Electric Power Research Institute, Palo
Alto, CA, 1977.
Be78 Beck, H.L., Perturbation of the Natural Radiation
Environment Due to the Utilization of Coal as an Energy
Source. Proceedings, DOE/UT Symposium on the Natural
Radiation Environment, Houston, TX, 1978.
Co77 Considine, D.M. (ed.), Energy Technology Handbook. McGraw-
Hill, NY, 1977.
Co75 Cowherd, C. et al., Hazardous Emission Characteristics of
Utility Boilers. NTIS PB-245-915, 1975.
De77 Dennis, R., et al., Filtration Model for Coal Fly Ash
with Glass Fabrics. EPA 600/7-77-084, U.S. Environmental
Protection Agency, Research Triangle Park, NC, 1977.
De79 Dennis, R. and K.A. Klemm, Fabric Filter Model Format
Change. Vol.1, Detailed Technical Report, EPA
600/7-79-0432, U.S. Environmental Protection Agency,
Research Triangle Park, NC, 1979.
DOES5 U.S. Department of Energy, Annual Energy Outlook. Energy
Information Agency, Washington, DC, 1985.
DOES6 U.S. Department of Energy, Annual Energy Outlook. Energy
Information Agency, Washington, DC, 1986.
EPA81 U.S. Environmental Protection Agency, The Radiological
Impact of Coal-fired Industrial Boilers (Draft Report),
Office of Radiation Programs, Washington, DC, 1981.
EPA80 U.S. Environmental Protection Agency, Fossil Fuel-Fired
Industrial Boilers—Background Information for Proposed
Standards. Chapters 3-5, Research Triangle Park, NC, June
1980.
EPA89 U.S. Environmental Protection Agency, "Coal and Oil
Combustion Study: Summary and Results," draft report in
preparation, Office of Air Quality, Planning and
Standards, Research Triangle Park, NC, scheduled for
publication during 1989.
Fa79 Facer, J.F., Jr., Uranium in Coal. U.S. Department of
Energy Report, GJBX-56(79), Washington, DC, 1979.
IGS77 Illinois State Geological Survey, Trace Elements in Coal:
Occurrence and Distribution, NTIS Report No. PB-270-922,
June 1977.
7-25
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Le67 Lederer, C.M.; Hollander, J.M.; and I. Perlman, Table of
Isotopes. Sixth Edition, John Wiley and Sons, NY, 1967.
Me86 Mead, R.C., Post B.K.; and G.W. Brooks, Summary of Trace
Emissions from, and Recommendations of Risk Assessment
Methodologies for. Coal and Oil Combustion Sources. Radian
No. 203-024-41, Radian Corporation, Research Triangle
Park, NC, 1986.
RC83 Radian Corporation, Boiler Radionuclide Emissons Control;
The Feasibility and Costs of Controlling Coal-fired Boiler
Particulate Emissions, Prepared for the Environmental
Protection Agency, January 1983.
SRI77 Stanford Research Institute, "Potential Radioactive
Pollutants Resulting from Expanded Energy Programs," NTIS
Report No. PB-272-519, August 1987.
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.
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.
TRI79 Teknekron Research, Inc., "Utility Simulation Model
Documentation, Vol. 1, R-001-EPA-79, Prepared for the
Environmental Protection Agency, Washington, DC, July
1979.
Tr81 Trigillo, G., Volume Reduction Techniques in Low-Level
Radioactive Waste Management. NUREG-/CR 2206, U.S. Nuclear
Regulatory Commission, 1981.
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,
NM, 1982.
Zu79 Zubovic, P., et al., "Assessment of the Chemical Compo-
sition of Coal Resources," USGS Expert Paper Presented at
the United Nations Symposium on World Coal Prospects,
Katowice, Poland, April 15-23, 1979.
7-26
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8. INACTIVE URANIUM MILL TAILINGS
8.1 DESCRIPTION OF INACTIVE URANIUM MILL TAILINGS SITES
Twenty-four former uranium processing sites were designated
as Title I sites under the Uranium Mill Tailings Radiation Con-
trol Act (UMTRCA) of 1978. The Inactive Uranium Mill Tailings
• source category comprises 18 final disposal sites where the tail-
ings and other wastes from these site are being consolidated and
stabilized for long-term isolation. Radon-222, the decay product
of the residual radium-226 in the tailings, is emitted to the air
from the tailings. Radon emissions from licensed uranium mill
tailings sites are addressed in Chapter 9.
8.1.1 Rulemaking History and Current Regulations
In enacting the UMTRCA (Public Law 95-604, 42 USC 7901), the
Congress found that:
o "Uranium mill tailings located at active and in-
active mill operations may pose a potential and
significant radiation health hazard to the public,
and that..."
o "Every reasonable effort should be made to provide
for the stabilization, disposal, and control in a
safe and environmentally sound manner of such
tailings in order to prevent or minimize radon
diffusion into the environment and to prevent or
minimize other environmental hazards..."
To these ends, the Act required the EPA to set generally ap-
plicable standards to protect the public against both radiolog-
ical and nonradiological hazards posed by residual radioactive
materials at uranium mill tailings sites. Residual radioactive
material means (1) tailings waste resulting from the processing
of ores for the extraction of uranium and other valuable constit-
uents, and (2) other wastes, including unprocessed ores or low
grade materials at sites related to uranium ore processing. The
term tailings will be used to refer to all of these wastes.
The UMTRCA divided uranium mill tailings sites into two
groups: Title I covering inactive and abandoned sites and Title
II covering those sites for which licenses had been issued by the
Nuclear Regulatory Commission (NRC) or its predecessor or by an
Agreement State. Twenty-four sites have been designated Title I
sites under the UMTRCA. Under this Act, the EPA was required to
develop general standards to govern the remedial activities con-
ducted by the Secretary of Energy or his designee under section
275a. of the Atomic Energy Act of 1954, at the sites identified
under Title I. The Department of Energy (DOE) is responsible for
the cleanup and long-term stabilization of the tailings at these
sites, consistent with the generally applicable standards devel-
oped by the EPA.
8-1
-------
Under the UMTRCA, the EPA was required to promulgate stand-
ards before the DOE could begin cleanup of the Title I sites.
These standards were required, to the maximum extent practicable,
to be consistent with the requirements of the Solid Waste Dispos-
al Act (SWDA) as amended. The SWDA includes the provisions of
the Resource Conservation and Recovery Act (RCRA) .
Because some buildings had been found to be contaminated
with tailings resulting in high radiation levels, interim stand-
ards for cleanup of residual radioactivity that had contaminated
land and buildings were published in the Federal Register on
April 22, 1980. This allowed DOE to proceed with the cleanup of
offsite tailings contamination without waiting for the formal
promulgation of a regulation through the EPA rulemaking process.
At the same time, proposed standards for the cleanup of the in-
active mill tailings were published for comment.
The proposed cleanup standards were followed by proposed
disposal standards that were published in the Federal Register on
January 9, 1981. The disposal standards applied to the tailings
at the 24 designated sites and were designed to place them in a
condition which would be safe for a long time. Final standards
for the disposal and cleanup of inactive uranium mill tailings
were issued on January 5, 1983.
The American Mining Congress and others immediately peti-
tioned the Tenth Circuit Court of Appeals for a review of the
standards. On September 3, 1985, the Tenth Circuit Court upheld
the inactive mill tailings standards, with the exception of the
groundwater protection portions which were remanded to EPA for
revision. The EPA is currently developing new groundwater stand-
ards under this rule. The disposal standard that applies to the
24 Title I sites (40 CFR 192, Subpart A) requires long-term stab-
ilization of the tailings and establishes a design standard so
that post-stabilization radon-222 releases do not exceed an emis-
sion rate of 20
8.1.2 Identification and Status of Sites
The tailings contain residual radioactive materials, includ-
ing traces of unrecovered uranium and most of its decay products,
as well as various heavy metals and other elements, often at
levels exceeding established standards. Of the 24 processing
sites designated under Title I of the UMTRCA, 23 are situated in
the generally semi-arid to arid western United States. The site
locations vary from isolated, sparsely populated rural settings
to populated urban communities.
The DOE has developed and is implementing a program for re-
medial actions at these 24 sites. The DOE's Uranium Mill Tail-
ings Remedial Action Program (UMTRAP) calls for the removal of
tailings from sites in highly populated areas or where the long-
term stabilization is threatened by flooding or could result in
8-2
-------
the contamination of groundwater. Under Public Law 95-604, as
amended, the DOE is to complete disposal and stabilization by the
end of fiscal year (FY) 1994. To date, disposal at seven sites
has been completed, and tailings at all sites are scheduled to be
covered by February 1993 (DOE88). The quantity of tailings and
proposed remedial action are summarized for each site in Table 8-1.
The information is Table 8-1 shows that once the DOE has complet-
ed its program, there will be 19 disposal sites. However, since
the remedial action at the Converse County site calls for dispos-
al under 40 feet of cover, there will be 18 sites where there is
a potential for radon-222 emissions that could cause risks to
public health.
8.1.3 Existing Emission Controls
Previous analyses have shown that the only effective means
of controlling radon emissions from the tailings is to cover the
tailings with an earthen cover thick enough to attenuate the
radon fluxing from the tailings. As discussed in Appendix B,
earthen covers reduce the amount of radon released to the air by
retaining the radon in the cover long enough for it to decay.
The 40 CFR 192 standards require that the cover be designed so
that the average radon flux does not exceed 20 pCi/m2/s. Gen-
erally accepted models are available to demonstrate the adequacy
of the design (Ro84). The design flux from the covers that the
DOE has approved for these piles range from the UMTRCA limit of
20 pCi/m2/s to 0.5 pCi/m2/s (see Section 8.2, Table 8-2).
At the sites where remedial actions are pending, no controls
are currently in place to reduce radon emissions. Thin interim
earthen covers have been used at some sites and may reduce the
amount of radon released to the air, but these are intended pri-
marily to control wind erosion of the tailings. At sites where
long-term stabilization under,UMTRCA has been completed, thick
earthen covers have been placed on the tailings, and the radon
fluxes will likely be below the long-term design flux.
8.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT
Previous assessments have evaluated the risks from radon-222
releases from these sites under both the assumption that the
tailings remain unreclaimed and that the stabilization and dis-
posal of tailings under UMTRCA just meets the 20 pCi/m2/s cover
design. In this assessment, the risks that will be incurred once
disposal in accordance with UMTRCA is completed are evaluated,
along with alternatives of limiting post-disposal flux to 6 and 2
pCi/m2/s, respectively. The evaluation of the risks that would
be incurred if the tailings remain unreclaimed has been dropped.
This reflects the fact that the DOE is proceeding, as required by
Public Law 95-604, with the reclamation of these sites, and that
all sites are scheduled to be under cover by early 1993.
8-3
-------
Table 8-1. Quantity of tailings and planned remedial actions at inactive
uranium mill tailings sites.
Site
Quantity
of Tailings
(106 tons)
Proposed
Action
Schedule^)
Start Finish
Monument Valley, AZ
Tuba City, AZ
Durango, CO
Grand Junction, CO
Gunnison, CO
Maybell, CO
Naturita, CO
New Rifle, CO
Old Rifle, CO
Slick Rock (NC), CO
Slick Rock (UC), CO
Lowman, ID
Ambrosia Lake, NM
Shiprock, NM
Belfield, ND
Bowman, ND
Lakeview, OR
Canonsburg, FA
Falls City, TX
Green River, UT
Mexican Hat, UT
Salt Lake City, UT
Converse County, WY
Riverton, WY
1.2 Removal to Mexican Hat Site
0.8 Stabilization in place
1.6 Removal to Bodo Canyon Site
1.9 Removal to Cheney Site
0.5 Removal to Landfill Site
2.6 Stabilization in place
0.6 Removal to Dry Flats Site
2.7 Removal to Estes Gulch Site
0.4 Removal to Estes Gulch Site
0.04 Removal to Slick Rock (UC)
0.35 Stabilization in place
0.09 Stabilization in place
2.6 Stabilization in place
1.5 Stabilization in place
Removal to Bowman Site
Stabilization in place
0.13 Removal
0.4 Stabilization in place
2.5 Stabilization in place
0.12 Stabilization in place
2.2 Stabilization in place
1.7 Removal to S. Clive Site
0.19 Stabilization in place
0.9 Removal to UMETCO's Gas
Hills Licensed Site
FY90 FY91
UW FY90
UW
UW
FY90
FY91
FY91
UW
UW
FY92
UW
FY92
FY92
FY90
UW
UW
UW
UW
FY90
FY93
FY92
FY92
FY92
FY92
FY92
DONE
DONE
FY92
FY90
DONE
FY93
FY93
DONE
DONE
FY92
DONE
FY91
DONE
FY89
FY91
(a) DOE88
(b) The start and finish dates refer to construction activities to stabilize
and cover the tailings. The finish dates do not include development and
implementation of the Surveillance and Monitoring Program or Certification
that the remedial action is complete.
(c) UW - underway, i.e., remedial actions to stabilize the tailings have been
initiated.
(d) North Continent pile
(e) Union Carbide pile
8-4
-------
The radon releases from the tailings at the 18 disposal
sites that will remain once UMTRCA disposal is completed are as-
sessed on a site-by-site basis. The following sections detail
how the radon release rates were developed and the sources of the
meteorological and demographic data used in the assessment. De-
tails of the values that were provided to the AIRDOS-EPA/DARTAB/
RADRISK codes are presented in Appendix A.
8.2.1 Development of the Radon Source Terms
Radon source terms for the post-UMTRCA disposal of the tail-
ings at these sites are calculated on the basis of the DOE's est-
imated radon fluxes through the approved cover designs and the
areas of the disposal sites. The DOE's design fluxes and the
areas of the disposal sites are those reported in DOE88. For the
alternative fluxes of 6 and 2 pCi/m2/s, the source terms are
calculated using the lower of the value for the design flux or
the appropriate flux limit. The areas of the final disposal
sites, the cover design flux rate, and the radon source terms
calculated for each pile are shown for each alternative flux in
Table 8-2.
8.2.2 Demographic and Meteorological Data
In assessing the exposures and risks that result from the
release of radon, site-specific demographic data have been used.
Demographic data for the nearby individuals (0-5 km) were devel-
oped for each site by surveys conducted during site visits (PNL84).
For sites that were estimated to have the highest risks, these
data have been updated based on site visits made by SC&A during
1989 or on the basis of information provided by the DOE for new
disposal sites (see Appendix A for details). The results of
those surveys are shown in Table 8-3. The populations between 5-
80 km were generated using the computer code SECPOP. Meteorolog-
ical data were obtained from the nearest station with data in an
appropriate format for use in the assessment codes.
8.3 RESULTS OF THE RISK ASSESSMENT FOR INACTIVE MILLS
The AIRDOS-EPA/DARTAB/RADRISK codes were used to estimate
the lifetime fatal lung cancer risk for individuals living near
the tailings impoundments and the number of fatal cancers per
year in the regional (0-80 km) populations around these sites.
8.3.1 Exposures and Risks to Nearby Individuals
The estimates of the exposure and risk to nearby individuals
once UMTRCA disposal is completed are shown in Table 8-4. The
lifetime fatal cancer risks for individuals residing near these
disposal sites range from 4E-7 to 2E-4. The maximum lifetime fatal
cancer risk of about 0.02 percent (2 in 10,000) is estimated at
the Shiprock site in New Mexico at a distance of 750 meters from
the center of the impoundment.
8-5
-------
Table 8-2.
State/Site
Summary of radon-222 emissions from inactive uranium
mill tailings disposal sites.
Site Design UMTRCA
(acres) Flux Limit
(PCi/m2/s)
Arizona
Tuba City
Colorado
Durango -Bodo Canyon
Grand Junction - Cheney Site
Gunnison - Landfill Site
Maybell
Naturita - Mill Site
New/Old Rifle - Estes Gulch
Slick Rock - Combined
Idaho
Lowman
New Mexico
Ambrosia Lake
Shiprock
North Dakota
Bowman/Bel fie Id
Oregon
Lakeview
Pennsylvania
Canonsburg
Texas
Falls City
Utah
Green River
Mexican Hat
Salt Lake City - S. Clive
22
40
62
38
80
23
71
6
5
105
72
12
30
18
146
9
68
50
9.3
20
6.5
1.9
7.1
20
5.8
5.7
16.7
20
3.9
7.5
7
13.2
0.5
12
20
2 . 6E+1
1 . OE+2
5 . 1E+1
9 . 2E+0
7 . 3E+1
1 . 5E+1
1 . 8E+2
4 . 4E+0
3 . 6E+0
2 . 2E+2
1.8E+2
6 . OE+0
2 . 9E+1
1.6E+1
2 . 5E+2
5.7E-1
1 . OE+2
1 . 3E+2
6 pCi/mz/s
Limit
1 . 7E+1
3 . 1E+1
4.8E+1
9 . 2E+0
6 . 1E+1
1 . 5E+1
5 . 4E+1
4.4E+0
3 . 6E+0
8 . OE+1
5 . 5E+1
6 . OE+0
2 . 3E+1
1 . 4E+1
1 . 1E+2
5.7E-1
5 . 2E+1
3 . 9E+1
.. *,—/•« A
Limit
5 . 6E+0
l.OE+1
1 . 6E+1
9 . 2E+0
2 . OE+1
5 . 9E+0
1 . 8E+1
1 . 5E+0
1 . 3E+0
2 . 7E+1
1 . 8E+1
3 . 1E+0
7 . 7E+0
4 . 6E+0
3 . 7E+1
5.7E-1
1.7E+1
1 . 3E+1
Totals
857
1.3E+3
5.9E+2
2.2E+2
(a) For each case, emissions are calculated based on the area of the site and
the lower of the DOE-approved cover design flux or the appropriate 20, 6, or
2 pCi/m2/s limit.
(b) Final cover design not available, UMTRCA limit of 5 pCi/g radium assumed due
to the fact that only residual contamination exists at this site.
8-6
-------
Table 8-3. Estimated number of persons living within 5 km of the centroid of
tailings disposal sites for inactive mills(a).
State/Site
Distance (kilometers)
0.0-0.5 0.5-1.0 1.0-2.0 2.0-3.0 3.0-4.0 4.0-5.0 Total
Arizona
Tuba City
Colorado
Durango
Grand Junction
Gunnison
Maybe 11
Naturita
New/Old Rifle
Slick Rock
Idaho
Lowman
New Mexico
Ambrosia Lake
Shiprock
North Dakota
Bowman/Be If ield
Oregon
Lakeviev
Pennsylvania
Canonsburg
Texas
Falls City
Utah
Green River
Mexican Hat
Salt Lake
0
0
0
0
0
0
0
3
9
0
0
0
0
950
0
0
0
0
18
0
0
0
0
0
0
16
76
0
155
3
16
2,960
3
14
0
0
12
2
0
0
0
65
0
0
87
0
1,904
9
543
7,988
18
257
279
0
15
0
0
8
0
20
16
3
0
0
1,034
3
1,704
5,126
0
810
56
0
0
0
26
11
0
106
0
0
16
0
1,016
6
1,457
2,830
15
397
0
0
19
0
31
22
0
902
49
0
30
0
839
12
464
2,281
9
20
0
0
64
2
57
41
0
1,093
65
22
218
0
4,948
33
4,184
22,135
45
1,498
335
0
Total
962
3,261 11,164 8,795 5,880 4,678 34,740
(a) PNL84, updated per SC&A site visits and DOE data (see Appendix A)
8-7
-------
Table 8-5,
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around inactive tailings
disposal sites.
Mill
Fatal Cancers
per Year
Arizona
Colorado
Idaho
New Mexico
North Dakota
Oregon
Pennsylvania
Texas
Utah
Tuba City
Durango
Grand Junction
Gunnison
Maybell
Naturita
New/Old Rifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/ Bel field
Lakeview
Canonsburg
Falls City
Green River
Mexican Hat
Salt Lake City
1.3E-4
6.7E-4
9.9E-4
7.5E-5
l.OE-4
3.5E-5
5.3E-4
6.4E-6
9.7E-6
5.3E-4
3.0E-3
4.0E-6
1.3E-4
4.7E-3
7.1E-3
3.3E-6
3.4E-4
4.9E-5
Total
1.8E-2
8-10
-------
Table 8-6. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from inactive
uranium mill tailings disposal sites.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
4,
89,
4,900,
0
0
0
130(a)
500
000
000
0
0
0
4E-4
2E-3
2E-3
1E-2
Totals 5,000,000 2E-2*
(a) All of the individuals in this risk interval reside near the
Shiprock disposal site in New Mexico.
* Totals may not add due to independent rounding.
8.3.4 Exposures and Risks Under Alternative Standards
Once all the tailings piles are stabilized and disposed of
in accordance with the UMTRCA disposal standard, the radon-222
emission rates will all be at or below 20 pCi/mvs. Alterna-
tive flux limits of 6 and 2 pCi/m2/s are also evaluated. Esti-
mates of what the risks would be for these alternative levels are
shown in Tables 8-7 through 8-9 for the 6 pCi/m2/s alternative
and in Tables 8-10 through 8-12 for the 2 pCi/m2/s alternative.
The estimates are obtained using the methodology described in
Section 8.2, but assuming all piles will achieve the lower of the
cover design flux or the radon flux rate assumed for the alterna-
tive.
These estimates show that for nearby individuals the maximum
lifetime fatal cancer risk could be reduced from 2E-4 at the ex-
isting UMTRCA standard to 7E-5 at a limit of 6 pCi/m2/s or 2E-5
at a limit of 2 pCi/m2/s. The number of deaths/year that will
occur in the regional populations would be reduced by about one-
half (from 2E-2 to 1E-2) at a limit of 6 pCi/m2/s. At a limit
of 2 pCi/m2/s, the deaths/year would be reduced by almost nine-
tenths (from 2E-2 to 3E-3).
8-11
-------
Table 8-7. Estimated exposures and risks to Individuals living near inactive
tailings sites assuming a 6 pCi/m^/s radon flux limit.(*)
State/Site
Arizona
Tuba City
Maximum
Radon
Concentration
(pCi/1)
1.3E-3
Maximum
Exposure
(WL)
4.4E-6
Maximum Lifetime
Fatal Cancer Risk
to Individual
6E-6
Distance^)
(meters)
1.500
Colorado
Durango
Grand Junction
Gunnison
Maybe11
Naturita
New/Old Rifle
Slick Rock
Idaho
Lowman
New Mexico
Ambrosia Lake
Shiprock
North Dakota
Bowman/Belfield
Oregon
Lakeview
Pennsylvania
Canonsburg
Texas
Falls City
Utah
Green River
Mexican Hat
Salt Lake City
3.
1.
1.
7.
1.
,3E-3
,3E-3
.6E-4
.4E-4
,3E-2
8.0E-4
3.6E-3
4.4E-3
1.4E-4
1.6E-2
7.5E-4
1.5E-3
1.7E-2
6.0E-3
2.1E-4
5.6E-3
1.3E-5
1.1E-5
5.4E-6
7.0E-7
4.8E-6
3.5E-5
2.9E-6
l.OE-5
1.2E-5
6.9E-7
4.8E-5
2.2E-6
5.4E-6
4.7E-5
2.0E-5
6.2E-7
1.9E-5
8.2E-8
2E-5
7E-6
IE-6
7E-6
5E-5
4E-6
1E-5
2E-5
9E-7
7E-5
3E-6
7E-6
7E-5
3E-5
9E-7
3E-5
1E-7
1,500
4,500
4,500
15,000
250
2,500
250
250
7,500
750
750
2,500
250
1,500
750
750
15,000
(a) The exposures and risks reflect the emissions calculated from the area of
the site and the lower of the DDE-approved cover design flux (see Table 8-2)
or the alternative 6 pCi/m^/s limit.
(b) Distance from center of a homogenous circular equivalent impound-
ment to the point where the exposures and risks were estimated.
8-12
-------
Table 8-8,
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around inactive tailings
disposal sites assuming a 6 pCi/m2/s radon flux limit.
Mill
Fatal Cancers
per Year
Arizona
Colorado
Idaho
New Mexico
North Dakota
Oregon
Pennsylvania
Texas
Utah
Tuba City
Durango
Grand Junction
Gunnison
Maybe 11
Naturita
New/Old Rifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/Bel field
Lakeview
Canonsburg
Falls City
Green River
Mexican Hat
Salt Lake City
8.8E-5
2.1E-4
9.3E-4
7.5E-5
8.5E-5
3.5E-5
1.6E-4
6.4E-6
9.7E-6
1.9E-4
9.2E-4
4.0E-6
1.1E-4
4.1E-3
3.1E-3
3.3E-6
1.7E-4
1.5E-5
Total
l.OE-2
8-13
-------
Table 8-9. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from inactive
uranium mill tailings disposal sites assuming a
6 pCi/m2/s radon flux limit.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0 0 0
1E-2 to 1E-1 0 0
1E-3 to 1E-2 0 0
1E-4 to 1E-3 0 0
1E-5 to 1E-4 2,500 1E-3
1E-6 to 1E-5 28,000 1E-3
< 1E-6 5,000,000 8E-3
Totals 5,000,000 1E-2*
Totals may not add due to independent rounding.
8-14
-------
Table 8-10. Estimated exposures and risks to individuals living near inactive
tailings sites assuming a 2 pCi/m2/8 radon flux limit.(a)
State/Site
Arizona
Tuba City
Maximum
Radon
Concentration
(pCi/1)
4.4E-4
Maximum
Exposure
(WL)
1.4E-6
Maximum Lifetime
Fatal Cancer Risk
to Individual
2E-6
Distance^)
(meters)
1,500
Colorado
Durango
Grand Junction
Gunnison
Maybell
Naturita
New/Old Rifle
Slick Rock
Idaho
Lowman
New Mexico
Ambrosia Lake
Shiprock
North Dakota
Bowman/Belfield
Oregon
Lakeview
Pennsylvania
Canonsburg
Texas
Falls City
Utah
Green River
Mexican Hat
Salt Lake City
1.1E-3
4.2E-2
1.6E-4
2.4E-4
5.0E-3
2.7E-4
1.2E-3
1.9E-3
4.6E-5
5.2E-3
3.6E-4
4.9E-4
5.6E-3
2.0E-3
2.1E-4
1.8E-3
4.2E-6
3.7E-6
1.8E-6
7.0E-7
1.6E-6
1.4E-5
9.8E-7
3.4E-6
5.4E-6
2.3E-7
1.6E-5
1.2E-6
1.8E-6
1.6E-5
6.6E-6
6.2E-7
6.1E-6
2.7E-8
5E-6
2E-6
1E-6
2E-6
2E-5
1E-6
5E-6
6E-6
3E-7
2E-5
2E-6
2E-6
2E-5
9E-6
9E-7
8E-6
4E-8
1,500
4,500
4,500
15,000
250
2,500
250
250
7,500
750
750
2,500
250
1,500
750
750
15,000
(a) The exposures and risks reflect the emissions calculated from the area of
the site and the lower of the DOE-approved cover design flux (see Table 8-2)
or the alternative 2 pCi/m2/s limit.
(b) Distance from center of a homogenous circular equivalent impound-
ment to the point where the exposures and risks were estimated.
8-15
-------
Table 8-11.
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around inactive tailings
disposal sites assuming a 2 pCi/m2/s radon flux limit.
Mill
Fatal Cancers
per Year
Arizona
Colorado
Idaho
New Mexico
North Dakota
Oregon
Pennsylvania
Texas
Utah
Tuba City
Durango
Grand Junction
Gunnison
Maybe 11
Naturita
New/Old Rifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/Bel field
Lakeview
Canonsburg
Falls City
Green River
Mexican Hat
Salt Lake City
2.9E-5
6.7E-5
3 . 1E-4
7.5E-5
2.8E-5
1.4E-5
5.3E-5
2.2E-6
3.6E-6
6.5E-5
3.0E-4
2.1E-6
3.6E-5
1.4E-3
1.1E-3
3.3E-6
5.7E-5
4.9E-6
Total
3.5E-3
8-16
-------
Table 8-12. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from inactive
uranium mill tailings disposal sites assuming a
2 pCi/m2/s radon flux limit.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
1,
7,
5,000,
0
0
0
0
100
500
000
0
0
0
0
2E-4
3E-4
3E-3
Totals 5,000,000 3E-3*
* Totals may not add due to independent rounding.
8.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS
Previous studies have examined the feasibility, effective-
ness, and cost associated with various options for controlling
releases of radioactive materials from uranium mill tailings
(NRC80, EPA82, EPA83, EPA86b). These studies have concluded that
long-term stabilization and control will be required to protect
the public from the hazards associated with these tailings. The
standards for long-term disposal established for these Title I
sites under the UMTRCA provide for controls to prevent misuse of
the tailings, protect water resources, and limit releases of
radon-222 to the air. The UMTRCA standard established a design
standard to limit long-term radon releases to an average flux not
to exceed 20 pCi/m2/s. As shown in Table 8-2, the DOE has ap-
proved cover designs ranging from 0.5 to 20 pCi/m2/s.
Both active and passive controls to reduce radon-222 emis-
sions from tailing are available. Active controls require that
some institution, usually a government agency, take the responsi-
bility for continuing oversight of the piles and for repairing to
the control system when needed. Fencing, warning signs, periodic
inspections and repairs, and restrictions on land use are active
control measures that may be used by the oversight agency. Pas-
sive controls, on the other hand, are measures of sufficient
permanence to require little or no active intervention. Passive
controls include thick earth or rock covers, barriers (dikes) to
protect against floods, burial below grade, and moving piles out
of flood prone areas, or away from population centers. Of the
two methods, active or institutional controls are not preferred
for long-term control of radon-222 emissions, since institutional
performance over a long period of time is not reliable.
8-17
-------
8.4.1 Long-Term Control Options
Previous studies (see above) have identified a number of
options to provide long-term control of radon-222 emissions from
the tailings. These include earthen or synthetic covers, extrac-
tion of radium from the tailings, chemical fixation, and sinter-
ing. The following paragraphs give a brief summary of these
options and provide the rationale for limiting the discussion of
costs and effectiveness to earthen covers.
8.4.1.1 Earth Cover
Covering the dried tailings with dirt is an effective method
for reducing radon-222 emissions (Ro84) and is already in use at
inactive tailings impoundments. The depth of soil required for a
given amount of control varies with the type of earth and radon-
222 exhalation rate.
Earth covers decrease radon-222 emissions by retaining the
radon-222 released from the tailings long enough so that a sig-
nificant portion will decay in the cover. A rapid decrease in
radon-222 emissions is initially achieved by applying almost any
type of earth. The high-moisture content earths provide greater
radon-222 emission reduction because of their smaller diffusion
coefficient.
In practice, earthen cover designs must take into account
uncertainties in the measured values of the specific cover mater-
ials used, the tailings to be covered, and predicted long-term
values of equilibrium moisture content for the specific location.
The uncertainty in predicting reductions in radon-222 flux in-
creases rapidly as the radon-222 emission limit is reduced.
The cost of adding earth covers varies widely with the loca-
tion of the tailings impoundment, its layout, availability of
earth, the topography of the disposal site, its surroundings, and
hauling distance. Another factor affecting costs of cover mater-
ial is its ease of excavation. In general, the more difficult
the excavation, the more elaborate and expensive the equipment
and the higher the cost. The availability of materials such as
gravel, dirt, and clay will also affect costs. If the necessary
materials are not available locally, they must be purchased and/
or hauled and costs could increase significantly.
8.4.1.2 Water Covers
Maintaining a water cover over the tailings reduces radon-
222 emissions (EPA86b). The degree of radon-222 control increas-
es with the depth of the water and decreases with the radium-226
content of the water. The diffusion coefficient of water is very
low (about one thousandth that of a 9 percent moisture content
soil) and water is thus an effective barrier for radon-222. In
shallow areas, however, radon-222 release is increased by thermal
gradients and wave motion, and emissions approach those of satur-
8-18
-------
ated tailings. Increased radium-226 content in the water reduces
its effectiveness in controlling radon-222 since it releases
radon-222. For a water depth less than 1 meter, the radon flux
is similar to saturated bare tailings.
Additional factors affecting the feasibility and/or effec-
tiveness of water covers include the evaporation and precipita-
tion rates at the site, pile construction and slope, the poten-
tial for groundwater contamination, and dike or dam stability.
Since the inactive tailings piles are currently dry and are
located in arid and semi-arid parts of the country, water covers
would require recontouring of the piles to contain the water and
active controls to monitor and maintain the water levels. Ac-
tive surveillance would also be needed to determine if there is
any seepage through the dam or sides, and groundwater samples
might be required periodically as a check for groundwater contam-
ination from seepage. For these reasons, water covers are not
suitable to provide long-term passive stabilization.
8.4.1.3 Synthetic Covers and Chemical Sprays
Synthetic material such as a polyethylene sheet can also re-
duce radon-222 emissions if carefully placed and sealed on dry
tailings. The overall effectiveness of synthetic covers is not
known since leaks occur around the edges and at seams and breaks.
Synthetic covers also have a limited life, especially in dry,
sunny, windy areas, and will not provide a long-term barrier to
radon-222. Chemical stabilization sprays that form coatings on
the dry tailings are effective for controlling dust, but are not
effective in controlling radon-222 since an impermeable cover is
not obtained. The lack of long-term stability of synthetic cov-
ers and the ineffectiveness of chemical sprays make these options
unsuitable for long-term passive control.
8.4.1.4 Thermal Stabilization
Thermal stabilization is a process in which tailings are
sintered at high temperatures. The Los Alamos National Labora-
tory has conducted a series of tests on tailings from four dif-
ferent inactive mill sites (Dr81). The results show that thermal
stabilization is effective in preventing the release (emanation)
of radon from tailings. However, before thermal stabilization
can be considered as a practical disposal method, information is
needed on the following: (1) the long-term stability of the sin-
tered material; (2) the interactions of the tailings and the re-
fractory, materials lining the kiln; (3) the gaseous and particu-
late emissions produced during sintering of tailings; and (4) re-
vised engineering and economic analysis as more information is
developed. Since gamma radiation is still present, protection
against the misuse of sintered tailings is required. While the
potential health risk from external gamma radiation is not as
great as that from the radon decay products, it can produce unac-
ceptably high exposure levels in and around occupied buildings.
8-19
-------
Also, the potential for groundwater contamination may require the
use of liners in a disposal area.
Given the experimental nature of this option and the uncer-
tainties involving the risk from external gamma radiation, ther-
mal stabilization will not be considered further in this analysis.
8.4.1.5 Chemical Processing
The Los Alamos National Laboratory has also studied various
chemical processes such as nitric acid leaching to extract
thorium-230 and radium-226 from the tailings, along with other
materials (Wm81). After removal from the tailings, the thorium
and radium can be concentrated and fixed in a matrix such as
asphalt or concrete. This greatly reduces the volume of these
hazardous materials and allows disposal with a higher degree of
isolation than economically achievable with unextracted tailings.
The major question regarding chemical extraction is whether
it reduces the thorium and radium values in the stripped tailings
to safe levels. If processing efficiencies of 80 to 90 percent
were attained, radium concentrations in tailings would still be
in the 30 to 60 pCi/g range. Thus, careful disposal of the
stripped tailings would still be required to prevent misuse.
Another disadvantage of chemical processing is the cost, although
some of the costs might be recovered from the sale of other min-
erals recovered in the processing (Th81).
8.4.1.6 Soil Cement Covers
A mixture of soil and Portland cement, called soil cement,
is widely used for stabilizing and conditioning soils (PC79).
The aggregate sizes of tailings appear suitable for soil cement,
which is relatively tough, withstands freeze/thaw cycles, and has
a compressive strength of 300 to 800 psi. When combined in a
disposal system with a 1-meter earth cover, soil (tailings)
cement would likely provide reasonable resistance to erosion and
intrusion, substantially reduce radon releases, and shield
against penetrating radiation. A previous study (EPA82) has est-
imated, based upon design specifications, that soil cement cover
will control emission to approximately the same level as a 2-
meter earth cover. Costs are expected to be comparable to those
of thick earth covers. The long-term performance of soil cement
is unknown, especially as tailings piles shift or subside with
age. Soil cement cracks at intervals when placed over large sur-
face areas. The importance of this cracking on the effectiveness
of soil cement has not been evaluated but is expected to be small.
8.4.1.7 Deep-Mine Disposal
Disposal of tailings in worked-out deep mines offers several
advantages and disadvantages compared to surface disposal options.
The probability of intrusion into and misuse of tailings in a
deep mine is much less than in the case of surface disposal.
8-20
-------
Radon releases to the atmosphere would be eliminated, for practi-
cal purposes, as would erosion and external radiation. The major
disadvantage of deep mine disposal is the potential contamination
of groundwater resulting from leaching of radionuclides and other
toxic chemicals from the tailings. Overall, while this method
can provide a relatively high level of protection against expo-
sure to radon and misuse of tailings, it has a high potential for
causing serious groundwater contamination and is very costly.
8.4.1.8 Caliche Cover
Caliche (calcium deposits that form within or on top of soil
in arid or semi-arid regions) cover material for mill tailings
pile has been suggested (Br81) as a control method. This mater-
ial may be effective in precluding excessive mobilization of cer-
tain radionuclides and toxic elements. However, the effectiveness
and long-term performance of such covers are not as yet known.
8.4.2 Comparison of Earth Covers to Other Control Techniques
In comparison to other control technologies, earth covers
have been shown to be cost effective (NRC80). Apart from cost
considerations, other benefits accrue by using earth covers as a
method to control radon-222 emissions. For example, synthetic
covers, such as plastic sheets, do not reduce gamma radiations.
However, earth covers that are thick enough to reduce radon-222
emissions will reduce gamma radiation to insignificant levels.
Further, chemical and physical stresses over a substantial period
of time destabilize synthetic covers, while earthen covers are
stable over the long run, provided the erosion caused by rain and
wind is contained with vegetation or rock covers, and appropriate
precautions are taken against natural catastrophes, e.g., floods
and earthquakes.
Earthen covers also reduce the likelihood of groundwater
contamination resulting from either storing radioactive materials
in underground mines (typically located under the water table) or
from using the leaching process to extract radioactive and non-
radioactive contaminants from mill tailings. Moreover, although
underground mine disposal is an effective method to protect
against degradation and intrusion by man (this maintains the long-
term stability of the cover), it nevertheless incurs a social cost.
For example, storing tailings in underground mines eliminates the
future development of the mines' residual resources. Again,
earthen covers with proper vegetation and rock covers can protect
against human intrusion, without incurring such social costs.
Finally, earth covers provide more effective long-term stab-
ilization than either water or soil cement covers. Albeit, soil
cement covers are comparable to earthen covers in terms of cost
effectiveness, their long-term performance is as yet unknown.
Water covers, on the other hand, do not provide the long-term
stability required for the needed time periods, which are at
8-21
-------
least 1,000 years. Moreover, earth covers are more practical
than water covers in arid regions.
8.4.3 Cost Estimates for Inactive Tailings Impoundments
For the reasons described above, the supplemental control
selected for long-term radon-222 control at inactive tailings im-
poundments is the earth cover control option. The cost estimates
developed below are for covers designed to meet the lower of the
DOE-approved cover design flux or the three alternative radon
emission levels: 20 pCi/m2/s (the level established by the
UMTRCA standard), 6 pCi/m2/s, and 2 pCi/m2/s. The basis for
the effectiveness of various depths of cover and the unit costs
used in this analysis are documented in Appendix B, Generic Unit
Costs for Earth Cover Based Radon-222 Control Techniques.
The thicknesses of the covers required to achieve a given
radon flux are a function of the soil type and the initial radon
flux from the pile. In this assessment, soil type B (see
Appendix B) is assumed. Table 8-13 presents the estimated radium
content and base area for each pile and the estimated thickness
of cover needed to achieve the lower of the DOE-approved design
flux or the flux limit for each of the three cases.
Five basic steps or operations are required to place earthen
covers on inactive tailings piles: regrading the slopes of the
pile to achieve long-term stability; procurement and placing of
the dirt cover; placing gravel on the pile tops; placing of rip-
rap on the pile sides; and reclamation of the borrow pits.
The first step is to regrade the inactive tailings piles, as
necessary, to prepare for the placement of the dirt cover. It is
assumed that existing piles have a slope of 2:1, and that the
placement of a dirt cover requires a slope no greater than 5:1
(EPA86b). The total cost for this operation is the product of
the volume regraded and the unit cost of grading. The volumes to
be regraded are based on the set of equations presented in
Appendix B, and two additional assumptions about the geometric
configuration of the piles. First, it is assumed that the length
of each base side of the pile is the square root of the area of
the pile. Second, it is assumed that the ratio between the
height and base side lengths of the piles is equal to 40 feet of
height per 2,100 feet in base side length. The unit cost of re-
grading is $1.36 per cubic yard.
The second step is the procurement and placement of the
earthen cover. In the case of inactive tailings piles, it is as-
sumed that dirt is available onsite at an average distance of one
mile from the pile (two miles round trip). The cost of the dirt
cover is a product of the volume required and unit costs for
excavating (on trucks), hauling, spreading, and compacting. The
volume is estimated by multiplying the surface area of the pile
(including the sides) by the depth of cover required to meet each
8-22
-------
Table 8-13. Estimated depths of earth cover needed to achieve given radon flux
rates.
State/Mill
Base Area
of Pile
(acres)
Estimated
Radium
Content
(pCi/g)
DOE Cover
Design Flux
(PCi/m2/s)
Cover Depth
Design
Flux
6 pCi/m2/
Flux
(meters)
's 2 pCi/m2/s
Flux
Arizona
Tuba City 22
Colorado
Durango 40
Grand Junction 62
Gunnison 38
Maybell 80
Naturita 23
Rifle 71
Slick Rock 6
550
670
665
315
200
45
745
115
9.3
20.0
6.5
1.9
7.1
5.0
20.0
5.8
4.4
3.7
4.9
5.5
3.6
2.3
3.9
3.2
4.8
5.0
5.0
5.5
3.7
2.3
5.1
3.2
6.0
6.2
6.2
5.5
4.9
3.3
6.3
4.3
Idaho
Lowman
160
5.7
3.6
3.6
4.7
New Mexico
Ambrosia Lake 105
Shiprock 72
North Dakota
Bowman/Belfield 12
Oregon
Lakeview
30
570
420
50
110
16.7
20.0
3.9
7.5
3.8
3.2
2.7
2.9
4.9
4.5
2.7
3.1
6.0
5.7
3.4
4.3
Pennsylvania
Canonsburg
Texas
.Falls City
Utah
Green River
Mexican Hat
Salt Lake City
18
146
9
68
50
2,315
190
75
670
480
7.0
13.2
0.5
12.0
20.0
6.2
2.8
5.3
4.3
3.4
6.4
3.7
5.3
5.0
4.7
7.5
4.9
5.3
6.2
5.8
(a) Estimated cover depths based on the radium content of the pile and the lover
of the DOE-approved cover design flux or the stated flux limit.
8-23
-------
of the three alternative radon flux rates. The equations used to
estimate surface areas, cover depths, and the total unit cost of
$6.01 per cubic yard for excavation, hauling, spreading, and com-
pacting are documented in Appendix B.
The third and fourth steps are erosion controls required to
provide long-term stabilization, after the final earth cover has
been put in place. The erosion control system is an essentially
maintenance-free gravel and rock system designed for arid condi-
tions. In this system, gravel is placed on the top of the pile,
and riprap (random broken stone) is placed on the sides of the
pile. The cost of each is a product of surface area, depth, and
unit costs. The depth required for adequate erosion protection
is assumed to be one-half yard (EPA86b). The equations used to
calculate the relevant surface areas and the unit costs of $7.55
per cubic yard for gravel and $23.00 per cubic yard for rip-rap
are documented in Appendix B.
The final operation is the reclamation of the borrow pits,
from which the earthen cover is extracted. The costs of borrow
pit reclamation is assumed to include regrading the sides of the
pits from 2:1 to 8:1. Regrading of the pit is calculated using
the same methodology as is used for estimating pile regrading.
The volume of the pit is based on the volume of dirt required for
cover. The ratio of height to base side length is the same as
given above, as is the unit cost for grading.
Table 8-14 presents the calculated volumes and surface areas
that were used in the development of the cost estimates. Tables
8-15, 8-16, and 8-17 summarize the costs of achieving the altern-
ative levels of control. The total cost of achieving the DOE-
approved cover fluxes under the UMTRCA limit of 20 pCi/m2/s at
all sites is approximately $127 million. The estimated total
costs at all sites for the 6 and 2 pCi/m2/s alternatives are
approximately $147 and $176 million, respectively.
Three overhead cost factors are used in conjunction with the
cost of earth cover described above. The first cost factor is
1.07, used to reflect overhead costs based on general industry
experience. The second factor of 3.3 represents the DOE's proj-
ect costs based its experience with the UMTRAP to date. The
project cost factor of 3.3 includes the additional costs to the
government of community participation, technology development and
evaluation, site acquisition, costs for a planning contractor,
management suppport, and design construction management and as-
sociated services. Since many of these project costs are sunk
costs, a third cost factor of 2.4, .is also provide. This altern-
ative project cost factor is based only on future costs.
In numerous cases (see Table 8-1) piles have already been
covered or are being covered under the UMTRCA design standard to
the DOE-approved cover flux of 20 pCi/m2/s or less. The cost
methodology, described above, assumes that no cover operations
had been done previously on the individual piles. Thus, the
8-24
-------
costs shown for achieving the UMTRCA limit includes the estimated
costs for piles where the work has already been accomplished.
Furthermore, in estimating the incremental costs of achieving the
alternative limits of 6 and 2 pCi/m^/s, no attempt has been
made to include the costs of redesigning covers and/or removing
and replacing existing erosion controls.
Table 8-14. Major volumes and surface areas used to calculate the costs to
achieve given radon-222 flux rates.(a)
Mill
Volume of
Tailings
Regraded
(m3)
Total Area
of Regraded
Tailings
(m2)
Volume of
Dirt Cover
20 pCi/m2/s
(m3)
Volume of
Dirt Cover
6 pCi/m2/s
(m3)
Volume of
Dirt Cover
2 pCi/m2/s
(m3)
Tuba City
Durango
Grand Junction
Gunnison
Maybell
Naturita
New/Old Rifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/Belfield
Lakeview
Canonsburg
Falls City
Green River
Mexican Hat
Salt Lake
52,661
129,323
249,410
119,863
365,781
58,167
304,949
7,513
5,694
549,525
311,346
21,250
83,480
38,958
901,125
13,774
9,045
178,730
89,696
163,266
252,962
155,204
326,532
95,843
289,243
24,490
20,358
428,322
293,274
48,980
121,946
73,369
595,619
36,684
27,715
202,571
390,554
611,864
1,249,416
846,533
1,163,327
224,748
1,116,735
78,072
72,452
1,613,738
952,914
133,351
349,514
454,250
1,695,198
196,171
118,975
687,066
432,507
821,648
1,271,025
846,533
1,221,989
224,748
1,488,390
78,072
72,452
2,081,669
1,329,748
133,351
378,555
466,320
2,196,394
196,171
139,477
947,354
537,673
1,013,074
1,567,617
846,533
1,604,841
318,473
1,827,521
105,900
95,207
2,583,866
1,673,606
168,260
521,533
552,344
2,894,744
196,171
171,972
1,184,864
(a) Volumes calculated to achieve the lower of the stated flux or the DOE-
approved design flux (see Table 8-2).
8-25
-------
Table 8-15. Estimated costs of achieving the UMTRCA limit of 20 pCi/m2/s.(a)
oo
I
to
CTi
Mill
Tuba City
Durango
Grand Junction
Gunnison
Maybe 11
Naturita
New/OldRifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/Be If ield
Lakeview
Canonsburg
Falls City
Green River
Mexican Hat
Salt Lake City
Regrade
Slopes
0.09
0.23
0.44
0.21
0.65
0.10
0.54
0.01
0.01
0.98
0.55
0.04
0.15
0.07
1.60
0.02
0.02
0.32
Apply
Dirt
Cover
3.07
4.81
9.82
6.65
9.14
1.77
8.77
0.61
0.57
12.68
7.49
1.05
2.75
3.57
13.32
1.54
0.93
5.40
Apply
Riprap
0.41
0.75
1.16
0.71
1.50
0.44
1.33
0.11
0.09
1.97
1.35
0.22
0.56
0.34
2.74
0.17
0.13
0.93
Apply
Gravel
0.20
0.37
0.57
0.35
0.74
0.22
0.66
0.06
0.05
0.97
0.67
0.11
0.28
0.17
1.35
0.08
0.06
0.46
(Millions of 1988
Reclaim
Borrow Total
Fits Cost
0.15
0.23
0.48
0.32
0.45
0.09
0.43
0.03
0.03
0.62
0.37
0.05
0.13
0.17
0.65
0.08
0.05
0.26
3.93
6.39
12.47
8.25
12.48
2.61
11.73
0.82
0.75
17.21
10.42
1.47
3.86
4.32
19.66
1.89
1.19
7.37
Dollars)
Total Cost Total Cost Total Cost
Including Including Including
1.07 DOE Cost 2.4 DOE Cost 3.3 DOE Cost
Factor Factor Factor
4
6
13
8
13
2
12
0
0
18
11
1
4
4
21
2
1
7
.20
.84
.35
.83
.35
.80
.55
.88
.80
.42
.15
.58
.14
.62
.03
.02
.27
.88
9
15
29
19
29
6
28
1
1
41
25
3
9
10
47
4
2
17
.42
.34
.94
.81
.94
.27
.15
.98
.79
.31
.00
.53
.28
.36
.17
.54
.85
.68
12.96
21.09
41.16
27.23
41.17
8.62
38.70
2.72
2.46
56.80
34.38
4.86
12.75
14.24
64.86
6.25
3.92
24.32
Totals
6.05 93.92 14.91
7.36
4.58 126.81
135.69
304.35
418.49
(a) Based on costs of achieving the lower of the DOE-approved cover design flux or the UMTRCA limit of
20 pCi/m2/s.
-------
Table 8-16. Estimated costs of achieving an average limit of 6
03
I
to
Mill Regrade
Slopes
Tuba City
Durango
Grand Junction
Gunnison
Maybell
Naturita
New/Old Rifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/Belfield
Lakeview
Canons burg
Falls City
Green River
Mexican Hat
Salt Lake City
0.09
0.23
0.44
0.21
0.65
0.10
0.54
0.01
0.01
0.98
0.55
0.04
0.15
0.07
1.60
0.02
0.02
0.32
Apply
Dirt
Cover
3.40
6.46
9.99
6.65
9.60
1.77
11.69
0.61
0.57
16.35
10.45
1.05
2.97
3.66
17.26
1.54
1.10
7.44
Apply
Riprap
0.41
0.75
1.16
0.71
1.50
0.44
1.33
0.11
0^09
1.97
1.35
0.22
0.56
0.34
2.74
0.17
0.13
0.93
Apply
Gravel
0.20
0.37
0.57
0.35
0.74
0.22
0.66
0.06
0.05
0.97
0.67
0.11
0.28
0.17
1.35
0.08
0.06
0.46
(Millions of 1988 Dollars)
Total Cost Total Cost Total Cost
Reclaim Including Including Including
Borrow Total 1.07 DOE Cost 2.4 DOE Cost 3.3 DOE Cost
Fits Cost Factor Factor Factor
0.17
0.31
0.49
0.32
0.47
0.09
0.57
0.03
0.03
0.80
0.51
0.05
0.15
0.18
0.84
0.08
0.05
0.36
4.27
8.12
12.65
8.25
12.96
2.61
14.79
0.82
0.75
21.07
13.52
1.47
4.10
4.41
23.78
1.89
1.36
9.51
4.57
8.69
13.54
8.83
13.87
2.80
15.83
0.88
0.80
22.54
14.47
1.58
4.39
4.72
25.45
2.02
1.45
10.18
10
19
30
19
31
6
35
1
1
50
32
3
9
10
57
4
3
22
.25
.49
.36
.81
.10
.27
.50
.98
.79
.56
.45
.53
.85
.60
.08
.54
.25
.83
14
26
41
27
42
8
48
2
2
69
44
4
13
14
78
6
4
31
.10
.80
.75
.23
.76
.62
.81
.72
.46
.52
.62
.86
.54
.57
.49
.25
.47
.39
Totals
6.05 112.55 14.91
7.36
5.49 146.35
156.60
351.25
482.97
(a) Based on costs of achieving the lower of the DOE-approved cover design flux or the UMTRCA limit of
6 pCi/m2/s.
-------
Table 8-17. Estimated costs of achieving an average limit of 2
oo
I
to
00
Mill
Tuba City
Durango
Grand Junction
Gunnison
Maybe 11
Naturita
New/Old Rifle
Slick Rock
Lowman
Ambrosia Lake
Shiprock
Bowman/Belfield
Lakeview
Canonsburg
Falls City
Green River
Mexican Hat
Salt Lake City
Regrade
Slopes
0.09
0.23
0.44
0.21
0.65
0.10
0.54
0.01
0.01
0.98
0.55
0.04
0.15
0.07
1.60
0.02
0.02
0.32
Apply
Dirt
Cover
4.22
7.96
12.32
6.65
12.61
2.50
14.36
0.83
0.75
20.30
13.15
1.32
4.10
4.34
22.74
1.54
1.35
9.31
Apply
Riprap
0.41
0.75
1.16
0.71
1.50
0.44
1.33
0.11
0.09
1.97
1.35
0.22
0.56
0.34
2.74
0.17
0.13
0.93
Apply
Gravel
0.20
0.37
0.57
0.35
0.74
0.22
0.66
0.06
0.05
0.97
0.67
0.11
0.28
0.17
1.35
0.08
0.06
0.46
(Millions of 1988
Reclaim
Borrow Total
Pits Cost
0.21
0.39
0.60
0.32
0.61
0.12
0.70
0.04
0.04
0.99
0.64
0.06
0.20
0.21
1.11
0.08
0.07
0.45
5.14
9.70
15.09
8.25
16.11
3.38
17.58
1.05
0.93
25.20
16.35
1.76
5.28
5.12
29.54
1.89
1.62
11.47
Dollars)
Total Cost Total Cost Total Cost
Including Including Including
1.07 DOE Cost 2.4 DOE Cost 3.3 DOE Cost
Factor Factor Factor
5.50
10.38
16.15
8.83
17.24
3.62
18.81
1.13
1.00
26.97
17.50
1.88
5.65
5.48
31.61
2.02
1.74
12.27
12
23
36
19
38
8
42
2
2
60
39
4
12
12
70
4
3
27
.33
.27
.23
.81
.67
.12
.20
.53
.24
.49
.25
.22
.68
.30
.89
.54
.90
.53
16
32
49
27
53
11
58
3
3
83
53
5
17
16
97
6
5
37
.96
.00
.81
.23
.17
.17
.03
.48
.08
.18
.97
.81
.43
.91
.48
.25
.36
.85
Totals
6.05 140.34 14.91
7.36
6.85 175.50
187.79
421.21
579.16
(a) Based on costs of achieving the lower of the DOE-approved cover design flux or the UMTRCA limit of
2 pCi/m2/s.
-------
8.4.4 Effectiveness of the Control Options
The effectiveness of the various cover options can be evalu-
ated by comparing the current average flux rate with the flux
rates achieved by each of the options. The emission of radon-222
from the inactive tailings sites once UMTRCA disposal is achieved
is estimated to be about 1,300 curies per year. Given the total
areas of the disposal sites, approximately 857 acres, this is
equivalent to an average post-UMTRCA flux of 12 pCi/m2/s. The
post-UMTRCA emissions are estimated to result in 2E-2 deaths per
year in the regional populations; reducing the emission limit to
6 pCi/m2/s would lower the deaths per year in the regional pop-
ulation to 1E-2 (see Table 8-8). Similarly, reducing the average
radon flux to 2 pCi/m2/s would reduce the deaths per year in
the regional populations to 3E-3.
8-29
-------
DOES 8
Dr81
8.5 REFERENCES
Br81 Brookins, P.G., "Caliche-Cover for Stabilization of
Abandoned Mill Tailings," in Proceedings of the Fourth
Symposium on Uranium Mill Tailings Management, Fort
Collins, Colorado, October 26-27, 1981, Geotechnical
Engineering Program, Civil Engineering Department,
Colorado State University, 1981.
U.S. Department of Energy, "Annual Status Report on the
Uranium Mill Tailings Remedial Action Program,"
Washington, D.C., December 1988.
Dreesen, D.R., Williams, J.M., and Cokal, E.J., "Thermal
Stabilization of Uranium Mill Tailings," in Proceedings of
the Fourth Symposium on Uranium Mill Tailings Management,
Fort Collins, CO, October 1981.
EPA82 U.S. Environmental Protection Agency, "Final Environmental
Impact Statement for Remedial Action Standards for
Inactive Uranium Processing Sites (40 CFR 192)," Vol.1,
EPA 520/4-82-013-1, Office of Radiation Programs,
Washington, D.C., October 1982.
EPA83 U.S. Environmental Protection Agency, "Final Environmental
Impact Statement for Standards for the Control of By-
product Materials from Uranium Ore Processing (40 CFR
192)," Vol.1, EPA 520/1-83-008-1, Office of Radiation
Programs, Washington, D.C. 1983
EPA86a U.S. Environmental Protection Agency, "Radon Flux
Measurements on Gardinier and Royster Phosphogypsum Piles
Near Tampa and Mulberry, Florida," EPA 520/5-85-029,
Office of Radiation Programs, Washington, DC, January
1986.
EPA86b U.S. Environmental Protection Agency, "Final Rule for
Radon-222 Emissions from Licensed Uranium Mill Tailings,"
EPA 520/1-86-009, Office of Radiation Programs,
Washington, D.C., August 1986.
NRC80 U.S. Nuclear Regulatory Commission, "Final Generic
Environmental Impact Statement on Uranium Milling," NUREG-
0706, Washington D.C., September 1980.
PC79 Portland Cement Association, "Soil-Cement Construction
Handbook," EB003.095, Skokie, II, 1979.
PNL84 Pacific Northwest Laboratory. "Estimated Population Near
Uranium Tailings," PNL-4959, WC-70, Richland, WA, January
1984.
8-30
-------
Ro84 Rogers, V.C., Neilson, K.K., and Kalkwarf, D.R., "Radon
Attenuation Handbook for Uranium Mill Tailings Cover
Design," NUREG/CR-3533, prepared for the U.S. Nuclear
Regulatory Commission, Washington, D.C., April 1984.
Sh85 Shiager, K.J., "Disposal of Uranium Mill Tailings,"
presented at the NCRP annual meeting, April 1985.
Th81 Thode, E.F., and Dreesen, D.R., "Technico-Economic
Analysis of Uranium Mill Tailings Conditioning
Alternatives," in Proceedings of the Fourth Symposium on
Uranium Mill Tailings Management, Fort Collins, CO,
October 1981.
Wm81 Williams, J.M., Cokal, E.J., and Dreesen, D.R., "Removal
of Radioactivity and Mineral Values from Uranium Mill
Tailings," in Proceedings of the Fourth Symposium on
Uranium Mill Tailings Management, Fort Collins, CO,
October 1981.
8-31
-------
-------
9. LICENSED URANIUM MILL TAILINGS FACILITIES
9.1 DESCRIPTION OF LICENSED URANIUM MILL TAILINGS
The licensed uranium mill tailings source category comprises
the tailings impoundments and evaporation ponds created by con-
ventional acid or alkaline leach processes at uranium mills li-
censed by the Nuclear Regulatory Commission (NRC) or the Agree-
ment States. Recovery of uranium by conventional milling results
in the release of uranium and its decay products to the air. The
risks associated with the release of uranium and other radionu-
clides in the form of particulates are addressed in the Uranium
Fuel Cycle source category (see Chapter 4). This assessment ad-
dresses only radon-222 released from the tailings impoundments
and their associated evaporation ponds. Previous evaluations
have shown that radon releases from other milling operations are
insignificant (NRC80, EPA83, EPA86).
9.1.1 Rulemakinq History and Applicable Standards
On January 13, 1977, the EPA issued Environmental Protection
Standards for Nuclear Power Operations (40 CFR 190). These stand-
ards limit the total individual radiation dose during normal oper-
ations from uranium fuel cycle facilities, including licensed
uranium mills. However, when 40 CFR 190 was promulgated, consid-
erable uncertainty existed regarding the public health risk from
radon-222 and the best method for managing new man-made sources of
this radionuclide. Therefore, the doses caused by emission of
radon-222 were excluded from the limits established in 40 CFR 190.
On April 6, 1983, the Agency proposed National Emission
Standards for Hazardous Air Pollutants (NESHAPS) for radionu-
clides under Section 112 of the Clean Air Act (CAA). At that
time, it determined that uranium fuel cycle facilities should be
exempted from the NESHAP for NRC-Licensed Facilities since they
were already subject to the dose limits of 40 CFR 190. During
the comment period, it was noted that radon-222 emissions from
operating uranium mills could pose significant public health
risks, and that such emissions were not subject to any current or
proposed EPA standards.
On September 30, 1983, under the authority of the Uranium
Mill Tailings Radiation Control Act (UMTRCA), the Agency issued
final standards (40 CFR 192) for the management of mill tailings
at licensed facilities. Although the UMTRCA standard requires
procedures to maintain radon-222 emission as low as reasonably
achievable (ALARA) during operations, it does not impose a numer-
ical limit on radon-222 emissions until after closure of a facil-
ity. Current NRC regulation imposes a concentration limit at the
boundary. After closure, the tailings must be disposed of in ac-
cordance with the standard and the post-disposal radon-222 emis-
sion rate cannot exceed an average of 20 pCi/m2/s. At the time
the UMTRCA standard was promulgated, taking into account the com-
9-1
-------
ments received during the radionuclide NESHAPS rulemaking, the
Agency stated that it would issue a Notice of Proposed Rulemaking
(under Section 112 of the CAA) with respect to control of radon-
222 emissions from uranium tailings piles during the operational
period of a uranium mill. This notice was published on October
31, 1984.
On September 24, 1986, the Agency promulgated a NESHAP (40
CFR 61, Subpart W) for radon-222 emissions from licensed uranium
mills during operations. The NESHAP imposes a work practice
standard of either phased or continuous disposal on all new tail-
ings impoundments and prohibits the use of existing tailings
piles after December 31, 1992.
9.1.2 Industry Profile
In December of 1988, the conventional uranium milling indus-
try in the United States consisted of 26 licensed facilities.
Three other mills have been licensed, but two never were con-
structed and one was built but never operated. The licensed con-
ventional uranium mills that have operated are in Colorado, New
Mexico, South Dakota, Texas, Utah, Washington, and Wyoming. Cur-
rently, 4 of the 26 licensed facilities are operating; 8 are on
standby status; and 14 are being or have been decommissioned.
The mills on standby status are being maintained, but they are
not processing uranium ore. When the demand for uranium in-
creases, these standby mills could resume milling. At the 14
facilities where decommissioning is in progress or completed, the
mills have been or are being dismantled; therefore these facili-
ties will never resume operations. The tailings at these 14 fa-
cilities have either been stabilized and reclaimed in conformance
with the UMTRCA requirements or reclamation activities are under-
way. The operational status of each conventional licensed mill
and the current extent of tailings reclamation are shown in
Table 9-1.
9.1.3 Process Description
Recovery of uranium by conventional milling methods is de-
scribed in Chapter 4, Section 4.2.2. Since the uranium ores typ-
ically contain only 0.05 to 0.5 percent uranium, virtually all of
the ore input to the mill remains as waste which is disposed of
in the tailings impoundment. The tailings wastes from the mill
are discharged into an impoundment. Impoundment technology has
changed with time. At older facilities, the pond areas were gen-
erally formed from dikes built with tailings sands or from soil
and rock from the pond area. As the pond is filled, the dikes
are raised with mill tailings sands. This practice is discourag-
ed but continues at some of the sites. At newer facilities, the
impoundment dikes were engineered and constructed with either
natural clay and/or man-made synthetic liners. The tailings dis-
charged to these impoundments are almost entirely covered by the
tailings pond.
9-2
-------
Table 9-1. Operating status of licensed conventional uranium mills as of
June 1989.
State/Mill
Owner
Operating
Status
Colorado
Canon City
Uravan
New Mexico
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Homestake
Cotter Corp.
Umetco Minerals
BP American
United Nuclear
Anaconda
Kerr-McGee
Homestake
Standby
Standby
Decommission
Decommission
Decommission
Standby
Active
Future
In Progress^)
Cover in Place
In Progress
In Progress
In Progress(e)
Future*^)
South Dakota
Edgemont
Texas
Panna Maria
Conquista
Ray Point
Utah
White Mesa
Rio Algom
Moab
Shootaring
Washington
Dawn
Sherwood
Wyoming
TVA
Chevron
Conoco/Pioneer
Exxon
Umetco Minerals
Rio Algom
Atlas
Plateau Resources
Dawn Mining
Western Nuclear
Decommission Completed
Active Future
Decommission In Progress
Decommission Completed
Active
Standby
Decommission
Standby
Future
In Progress
In Progress
Future
Decommission In Progress
Standby Future
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Pathfinder
Western Nuclear
Umetco Minerals
Rocky Mt. Energy
Pathfinder
Minerals Expl.
Exxon
American Nuclear
Corporation
Petrotomics
Standby
Decommission
Decommission
Decommission
Active
Standby
Decommission
Decommission
Future
In Progress
In Progress
In Progress
Future
Future
Cover in Place
Unknown
Decommission Design Approval Pending
9-3
-------
Table 9-1. Operating status of licensed conventional uranium mills as of
June 1989.(a)(continued)
(a) Data obtained from conversations with cognizant personnel in Agreement
States and the NRG, comments submitted by individual companies and the
American Mining Congress during the public comment period, and site
visits. Does not include mills licensed but not constructed.
(b) Active mills are currently processing ore and producing yellowcake.
Standby mills are not currently processing ore but are capable of
restarting. At mills designated by "Decommission", the mill structure has
been or is being dismantled and no future milling will occur at the site.
(c) Reclamation to the UMTRCA requirements is in various stages of completion,
creating a dynamic situation. The terms used to describe the reclamation
status are as follows: "Future" mean that the impoundment is being
maintained to accept additional tailings and that reclamation activities
have not been started; "Design Approval Pending" means that the final
disposal design has been submitted for regulatory approval and that
preliminary reclamation activities are underway; "In Progress" means that
active reclamation has begun, but the final cover is not completed; "Cover
in Place" designates that the final earthen cover has been completed, but
final stabilization has not been completed; and "Completed" means that
disposal and stabilization have been accomplished in accordance with the
UMTRCA requirements.
(d) According to UMETCO, the mill is being held on standby but the entire
impoundment area is being reclaimed. Thus, if future milling is done at
this facility a new impoundment will have to be constructed. For the
purposes of this analysis, the facility is grouped with other
decommissioning mills.
(e) The main impoundment, which is filled, arid the unlined evaporation ponds
are being reclaimed. The secondary impoundment and lined evaporation
ponds are being maintained to accept future tailings.
(f) The inactive impoundment containing tailings generated for the AEC is
covered with several feet of soil.
(g) The upper impoundment, which is filled, is being reclaimed. The lower
impoundment is being maintained to accept future tailings.
9-4
-------
9.1.4 Existing Emission Controls
During the operating period of the mill, radon releases from
the tailings are required to be maintained ALARA. The addition
of wet tailings provides a water cover which reduces the radon
emissions. The beaches are sprayed to prevent wind erosion and
control the radon. During operations and standby periods an in-
terim cover can be placed on portions or all of the tailings pile
to reduce radon and wind erosion before final reclamation. At
the end of the operating period, the tailings pond is dewatered
and the spraying of water on the beaches is discontinued. This
is done so that the tailings can dry sufficiently to provide a
stable base for the heavy equipment needed to regrade the impound-
ment and place the earthen covers required to meet the long-term
disposal criteria of the UMTRCA standard.
9.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT
The evaluation of the exposures and risks caused by emis-
sions of radon from licensed conventional uranium mills involves
three distinct assessments: the risks that result from the con-
tinued use of existing impoundments at the 11 facilities that are
operating or on standby; the risks that will occur once all ex-
isting piles are disposed of; and the risks that will result from
future tailings impoundments. As was done in the 1986 NESHAPS
rulemaking for this source category, the exposures and risks for
existing impoundments are assessed on a site-by-site basis, while
risks from future impoundments are assessed using model impound-
ments to represent the alternative technologies. The following
sections explain the basis for the assessments of existing sites,
while the emissions and risks estimated for future impoundments
are discussed in Section 9.4.2.
9.2.1 Assessment of Risks from Operating and Standby Mills
The overall risk from operating and standby mills includes
the risks that result from emissions during the operating or
standby phase, the drying out and disposal phase, and the post-
disposal phase. The following sub-sections detail how the radon
release rates were developed for each of these three phases to
obtain the source terms for the 11 operating and standby mills.
The sources of the meteorological and demographic data used in
the assessment are also discussed. Detailed information on the
inputs to the assessment codes is presented in Appendix A.
9.2.1.1 Development of the Radon Source Terms
Measured radon-222 release rates are not available for all
of the licensed tailings piles. Therefore, the radon source
terms are estimated for each phase based on the radon flux rate
per unit area and the area of the tailings. This assessment uses
the same basic methodology for estimating the radon releases and
the radon source terms that was used in the 1986 NESHAPS rule-
makings (EPA86). For each phase, the methodology involves three
9-5
-------
estimates: the radon flux per unit area, the fluxing area of the
tailings pile, and the duration, in years, of the phase.
For both the operating or standby phase and the drying and
disposal phase, the radon flux per unit area is calculated on the
assumption that 1 pCi/m2/s radon-222 is emitted per pCi/g radium-
226 in the tailings. While the EPA recognizes that this number
could be lower because of moisture and other factors, the conser-
vative value was used due to the lack of site-specific measured
values. In the calculations of the specific flux rates, the ra-
dium concentrations of the tailings used are those reported in
previous studies by the EPA and the NRC (EPA83, NRC80) or updated
values provided by the industry during the public comment period
(see Appendix A). For the post-disposal phase, the assumed radon
flux per unit area is the design flux of the approved cover, if
known, or the 20 pCi/m2/s (2 pCi/m2/s for facilities in Colorado)
limit established by the regulatory authorities responsible for
the implementation of the UMTRCA disposal standard.
Since water and dirt covers effectively attenuate radon,
during the operating or standby phase the calculated radon flux
rates are applied only to the dry areas of the operable pile and
any associated evaporation ponds. The areas of the piles that
are ponded, wet, covered with dirt, and dry have been updated
from information obtained during the public comment period.
Where no new information was provided, the areas were estimated
from aerial photographs taken of each pile in 1986.
During the drying and disposal phase the calculated radon
flux rates are applied to the total areas of the impoundment and
any associated evaporation ponds. This could lead to an over-
estimation of the radon releases during this period since cover
operations can proceed while the the piles are drying. For the
post-disposal phase, the radon flux is applied only to the area
of the impoundment. The areas of any associated evaporation
ponds are not included since the radium contamination in these
ponds is removed and transferred to the main impoundment prior to
stabilization.
The total areas of the piles, along with the areas that are
estimated to be non-fluxing (ponded, wet, or covered) and fluxing
(dry) and the radium concentrations in the tailings are shown in
Table 9-2.
To obtain the radon source term for each facility, it was
necessary to define the duration of each of the three phases.
The operating or standby phase is defined to be fifteen years.
While it is recognized that some of the impoundments do not have
15 years of capacity remaining at full production, the limited
processing that is now occurring makes it possible that these im-
poundments could remain operational for that length of time. The
drying out disposal period is defined to require five years,
based on industry and DOE experience to date. Finally, the post-
disposal period is defined as fifty years. Total emissions were
9-6
-------
Table 9-2. Summary of operable tailings impoundment areas and radium-226
content at operating and standby mills.
Surface Area (acres)
State/Impoundment
Colorado
Canon City - Primary
Canon City - Secondary
Canon City - Total
New Mexico
Ambrosia Lake - Secondary
Ambrosia Lake - Evap. Ponds
Ambrosia Lake - Total
Homestake - Primary
Homestake - Secondary
Homestake - Total
Texas
Panna Maria
Utah
White Mesa
Rio Algom - Lower
Shootaring
Washington
Sherwood
Wyoming
Lucky Me - Pile 1-3
Lucky Me - Evap. Ponds
Lucky Me - Total
Shirley Basin
Sweetwater
Totals
Total
90
40
130
121
280
401
170
40
210
160
130
47
7
80
203
104
307
275
37
1,784
Covered
0
0
0
13
0
13
0
40
40
80
0
0
0
0
108
0
108
0
0
241
Ponded
88
40
128
0
162
162
100
0
100
40
55
18
2
0
35
104
139
179
30
853
Wet
2
0
2
0
0
0
0
0
0
40
70
29
1
40
0
0
0
36
0
218
Dry
0
0
0
108
118
226
70
0
70
0
5
0
4
40
60
0
60
60
7
472
Ra-226
(pCi/g)
400
400
400
237
22
87
300
300
300*
198
981
420
280
200
220
22
153
208
280
--
* The sand and slime fractions of the tailings are separated by a mobile
cyclone, and the exposed sands average 65 pCi/g Ra-226.
9-7
-------
estimated by simply summing the estimated emissions for each
period. The total was then divided by 70 to obtain the average
release per year for input to the assessment codes. The radon
source terms calculated for each pile are given in Table 9-3.
9.2.1.2 Sources of Demographic and Meteorological Data
Site-specific demographic data were used in assessing the
exposures and risks that result from the release of radon from
operable mills. Demographic data for the nearby individuals (0-5
km) were developed for each site by site visits made during late
1983 (PNL84). These data were verified and or updated for the
mills that were estimated to have the highest post-disposal risks
in the draft assessment (see Appendix A). The results of these
surveys for all 26 licensed facilities are shown in Table 9-5.
The population data between 5-80 km were generated using the com-
puter code SECPOP. Meteorological data were obtained from on-
site meteorological towers where available or from the nearest
meteorological station with suitable joint-frequency data.
9.2.2 Assessment of the Post-Disposal Risks
The UMTRCA rule-making (40 CFR 192) established requirements
for the long-term stabilization and disposal of uranium mill
tailings. In addition to protection of groundwater and long-term
isolation to prevent misuse of tailings, the UMTRCA standards re-
quire that the tailings cover be designed to limit the radon flux
through the cover to 20 pCi/m2/s or less. The NRC and the Agree-
ment States, which are responsible for implementing the UMTRCA
requirements at licensed facilities, require licensees to demon-
strate that the cover designs will achieve the 20 pCi/m2/s at the
end of 1,000 years.
9.2.2.1 Development of the Radon Source Terms
As was done for the assessment of Inactive Tailings (see
Chapter 8), the post-disposal source terms for each of the sites
was estimated on the basis of the area of the tailings impound-
ment (s) and the design flux or measured performance of the cover.
Where information on the design flux or performance of the cover
was unavailable, the UMTRCA limit of 20 pCi/m2/s (2 pCi/m2/s for
facilities in Colorado) was used. Table 9-4 summarizes the
areas, radon flux rates through the covers, and estimated annual
emissions for each of the 26 licensed facilities once disposal is
complete.
9.2.2.2 Sources of Demographic and Meteorological Data
The demographic and meteorological data used to assess the
post-UMTRCA disposal risks were obtained in the same manner as
those used in the assessment risks from operable and standby
impoundments. Table 9-5 summarizes the 0-5 kilometer populations
around each of the sites.
9-8
-------
Table 9-3. Summary of radon source terms calculated for operable mill
tailings impoundments.
Radon Emissions
State/Impoundment
Operating/
Standby
Phase
(Ci/y)
Drying/
Disposal
Phase
(Ci/y)
Post-
Disposal
Phase
(Ci/y)
Total
Over All
Phases
(Ci)
Average
Over All
Phases
(Ci/y)
Colorado
Canon City
New Mexico
Ambrosia Lake
Homestake
0.OE+0
6.6E+3
2.5E+3 4.4E+3
5.8E+2* 8.OE+3
3.3E+1 3.5E+4 5.0E+2
9.4E+2 1.1E+5 1.5E+3
5.4E+2 7.6E+4 1.1E+3
0.OE+0
4.OE+3
4.1E+2
4.1E+4
5.8E+2
6 . 3E+2
0 . OE+0
1 . 4E+2
1 . 6E+4
5 . OE+3
2 . 5E+2
1 . 2E+2
2 . 4E+2
1 . 8E+1
9 . 7E+4
3 . 7E+4
4 . 3E+3
1 . 4E+3
5 . 3E+2
6 . 1E+1
l.OE+3
2.OE+3
2.OE+2
3.6E+4
5.1E+2
Panna Maria
Utah
White Mesa
Rio Algom
Shootaring
Washington
Sherwood
Wyoming
Lucky Me
Shirley Basin
Sweetwater
* The source term for the operating/standby phase is based on the reported 65
pCi/g Ra-226 in the exposed sand fraction of the tailings. The average Ra-
226 content of 300 pCi/g is used to calculate the source term for the
drying/disposal phase, since once the water from the pond is decanted both
the sands and slimes will be exposed and drying.
1 . 2E+3
1 . 6E+3
2 . 5E+2
6 . OE+3
7 . 3E+3
1.3E+3
5 . 2E+2
7 . OE+2
9 . 5E+1
7 . 3E+4
9 . 6E+4
1.5E+4
1 . OE+3
1 . 4E+3
2 . 2E+2
9-9
-------
Table 9-4.
Summary of uranium mill tailings impoundment areas,
flux rates, and post-UMTRCA radon-222 release rates.
Owner/Impoundment
Surface
Area
(acres)
Radon Flux
Rate
(PCi/m2/s)
Radon-222
Release Rate
(Ci/y)
Colorado
Canon City 130
Uravan 7 0
New Mexico
L-Bar 128
Churchrock 100
Bluewater 305
Ambrosia Lake 368
Homestake 210
South Dakota
Edgemont 123
Texas
2
2
20
20
20
20
20
20
3.3E+1
1.8E+1
3.3E+2
2.6E+2
7.8E+2
9.4E+2
5.4E+2
3.1E+2
Panna Maria
Conguista
Ray Point
Utah
White Mesa
Rio Algom
Moab
Shootaring
Washington
Dawn
Sherwood
Wyoming
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
160
240
47
130
93
147
7
128
80
220
156
218
90
275
37
200
117
140
20
20
20
7
20
20
20
10
20
20
20
20
20
20
20
20
20
20
4 . 1E+2
6.1E+2
1.2E+2
1.2E+2
2.4E+2
3.8E+2
1.8E+1
1.6E+2
2 . OE+2
5.2E+2
4. OE+2
5.6E+2
2.3E+2
7. OE+2
9.5E+1
5.1E+2
3 . OE+2
3.6E+2
9-10
-------
Table 9-5. Estimated number of persons living within 5 km of the centroid of
tailings impoundments of licensed mills.(a)
Distance (kilometers)
State/Impoundment
Colorado
Canon City*
Uravan*
New Mexico
L-Bar
Churchrock*
Bluewater*
Ambrosia Lake*
Homes take*
South Dakota
Edgemont
Texas
Panna Maria
Conquista
Ray Point
Utah
White Mesa
Rio Algom*
Moab
Shootaring
Washington
Dawn*
Sherwood*
Wyoming
Lucky Me
Split Rock*
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
0.0-0.5
0
0
0
0
0
0
0
.0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0.5-1.0
0
0
0
0
0
0
0
0
12
0
0
0
0
0
0
3
0
0
0
0
0
0
0
0
0
0
1.0-2.0
0
0
0
18
0
0
187
0
42
3
21
0
0
9
0
93
0
0
0
0
0
0
0
0
0
0
2.0-3.0
184
0
0
52
25
0
104
0
33
12
21
0
0
33
0
157
0
0
30
0
0
0
0
0
0
0
3.0-4.0
2,767
0
42
51
220
0
42
286
81
9
30
0
0
1,094
0
96
32
0
75
0
0
0
0
6
0
96
4.0-5.0
2,982
0
124
150
294
0
57
1,182
285
18
58
8
40
1,225
171
62
17
0
40
0
0
0
0
0
0
0
Total
5,933
0
166
271
539
0
390
1,468
453
42
130
8
40
2,361
171
411
49
0
145
0
0
0
0
6
0
96
0
15
373
651 4,927
6,713 12,679
(a) Based on information developed by Pacific Northwest Laboratory during 1983
(PNL84). At facilities marked with an asterisk the data were verified and
updated as necessary during site visits made in 1989.
9-11
-------
9.3 RESULTS OF THE RISK ASSESSMENTS FOR LICENSED MILLS
9.3.1 Exposures and Risks from Operating and Standby Mills
The estimates of the risks to nearby individuals and the
deaths/year caused by operable and standby mills are substantially
lower than previous estimates. The differences are due to several
factors including:
o the elimination from the assessment of the licens-
ed mills that are decommissioning, reflecting the
fact that disposal of tailings is progressing un-
der the UMTRCA standards and that the regulatory
authorities responsible for implementing those
standards are requiring closure activities once
the impoundments are filled and/or the mill itself
is dismantled;
o the updated demographic data which show signific-
antly fewer people in the immediate vicinity of
these mills; and
o updated information on mill characteristics in-
cluding average radium content, partial reclama-
tion activities, and additional information on the
interim covers that have been placed at some mills
which allows radon reduction credit to be given
due to their thickness and/or moisture content.
These changes, along with changes in the meteorological data
(including correction of day/nite data sets inadvertently used in
the draft assessment) are detailed in Appendix A.
9.3.1.1 Exposures and Risks to Nearby Individuals
The AIRDOS-EPA and DARTAB,model codes were used to estimate
the increased chance of lung cancer for individuals living near
an operable or standby tailings impoundment and receiving the
maximum exposure. The results for exposure to the average emis-
sions from all phases, in terms of radon concentration (pCi/1),
exposure (WL), and lifetime fatal cancer risk are shown in Table
9-6. Table 9-6 also presents the lifetime fatal cancer risks
that are attributable to the 15 year operating or standby period.
The lifetime fatal cancer risks from all phases for individuals
residing near these mill sites range from 4E-4 to 5E-6. The max-
imum risk of about 4E-4 (4 in 10,000) is estimated at the Panna
Maria mill in Texas. The lifetime fatal cancer risks to nearby
individuals from the operating or standby periods range from 3E-5
to nil, with the highest risk estimated at the Homestake mill in
New Mexico. The negligible risks during the operating or standby
phase estimated for the Panna Maria, Canon City, and La Sal mills
results from the fact that the design of these impoundments al-
lows them to be kept totally wet.
9-12
-------
Table 9-6. Estimated exposures and risks to individuals living near operable
tailings impoundments.
State/Mill
Maximum
Radon
Concentration
(pCi/1)
Maximum
Exposure
(WL)
Lifetime
Fatal Cancer
Risk to
Individuals
(All Phases)
Lifetime
Fatal Cancer
Risk to
Individuals
(Operations)
Distance^3)
(meters)
Colorado
Canon City 4.2E-3
New Mexico
Ambrosia Lake 2.7E-3
Homestake 5.8E-2
Texas
Panna Maria l.OE-1
Utah
White Mesa 2.2E-3
Rio Algom 1.5E-3
Shootaring 8.8E-4
Washington
Sherwood 4.8E-3
Wyoming
Lucky Me 1.2E-3
Shirley Basin 2.2E-3
Sweetwater 6.1E-4
1.7E-5
1.4E-5
1.9E-4
3.0E-4
1.5E-5
6.4E-6
3.8E-6
1.9E-5
8.4E-6
1.6E-5
4.2E-6
2E-5
2E-5
3E-4
4E-4
2E-5
9E-6
5E-6
3E-5
1E-5
2E-5
6E-6
OE+0
9E-6
3E-5
OE+0
2E-6
OE+0
3E-6
1E-5
3E-6
5E-6
1E-6
3,500
7,500
1,500
750
25,000
4,500
4,500
3,500
25,000
25,000
25,000
(a) Distance from center of a homogenous circular equivalent impoundment
to the point where the exposures and risks were estimated.
9-13
-------
9.3.1.2 Exposures and Risks to the Regional Population
Collective population risks for the region around each mill
site were calculated from the annual exposure in person-WLM for
the population in the assessment area. Collective exposure cal-
culations expressed in person-WLM were performed for each mill by
multiplying the estimated concentration in each annular sector by
the population in that sector. Table 9-7 presents the estimated
regional fatal cancers from operable tailings impoundments for
all phases of operations and for the operating or standby phase
only.
The estimates indicate that these operable impoundments
cause 4E-2 deaths/year (4 deaths in 100 years) in the regional
(0-80 km) populations. The emissions from the operating or
standby period are estimated to cause 4E-3 deaths/year in the
regional population; approximately 10 percent of the risk from
all phases of operations.
9.3.1.3 Distribution of the Fatal Cancer Risk
The frequency distribution of the estimated lifetime fatal
cancer risk for all operable uranium mill tailings is presented
in Table 9-8. This distribution was developed by simply summing
the distributions projected for each of the 11 facilities. The
distribution does not account for overlap in the populations ex-
posed to radionuclides released from more than a single mill.
Given the remote locations of these facilities and the relatively
large distances between mills, this simplification does not sig-
nificantly understate the lifetime fatal cancer risk to any indi-
vidual.
9-14
-------
Table 9-7,
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around operable tailings
impoundments-.
Mill
All Phases
Fatal Cancers per Year
Operating Phase
Colorado
New Mexico
Texas
Utah
Washington
Wyoming
Total
Canon City
Ambrosia Lake
Homestake
Panna Maria
White Mesa
Rio Algom
Shootaring
Sherwood
Lucky Me
Shirley Basin
Sweetwater
6.6E-3
3.1E-3
7.7E-3
1.4E-2
1.1E-3
2.8E-4
2.2E-5
2.9E-3
6.0E-4
1.8E-3
1.2E-4
3.9E-2
O.OE+0
1.5E-3
8.3E-4
O.OE+0
1.1E-4
O.OE+0
1.1E-5
1.2E-3
1.6E-4
4.5E-4
3.0E-5
4.3E-3
Table 9-8.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from operable
uranium mill tailings piles.
Risk Interval
Number of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
Totals
0
0
0
230
31,000
1,000,000
850,000
1,900,000
0
0
0
6E-4
9E-3
2E-2
5E-3
4E-2
9-15
-------
9.3.2 Post-Disposal Exposures and Risks
The exposures and risks that will remain once the impound-
ments at these 26 licensed sites are disposed of are estimated
for the existing UMTRCA disposal design standard of 20 pCi/m2/s
and alternative fluxes of 6 and 2 pCi/m2/s. As was done in the
case of inactive tailings (see Chapter 8), the source terms for
each site were calculated based on the lower of the design (or
measured flux rate) or the applicable flux standard and the areas
of the impoundments. The estimates for all three alternatives
reflect the current demography around these sites.
9.3.2.1 Exposures and Risks Under the UMTRCA Standard
Once all the tailings piles are stabilized and disposed of
in accordance with the UMTRCA disposal standard, the radon-222
emission rates will all be at or below 20 pCi/m2/s. Estimates
of what the post-UMTRCA disposal risks will be are shown in
Tables 9-9 through 9-11.
The estimates show that for nearby individuals the maximum
lifetime fatal cancer risk will range from 3E-4 to 9E-7 once
disposal activities are completed. The number of deaths/year
that will occur in the regional populations around these 26 sites
is estimated to be 5E-2. The individuals at the highest risks
(>lE-4) reside near the Homestake and Panna Maria piles.
9.3.2.2 Exposures and Risks Under Alternative Disposal Standards
Risks to nearby individuals and the regional populations are
shown in Tables 9-12 through 9-14 for the alternative of 6
pCi/m2/s, and Tables 9-15 through 9-17 for the alternative of 2
pCi/m2/s.
At 6 pCi/m2/s, the maximum individual lifetime fatal
cancer risk is 9E-5 at the Pahna Maria site, a factor of approxi-
mately three lower than the risks under the UMTRCA disposal
standard. The estimated deaths per year are reduced from 5E-2 to
2E-2. Similarly, at the alternative of 2 pCi/m2/s, the maximum
individual risk is reduced by another factor of three to 3E-5,
and the deaths/year from all 26 sites is reduced to 6E-3.
9-16
-------
Table 9-9. Estimated exposures and risks to individuals living near licensed
tailings impoundments post-UMTRCA disposal.
Maximum
State/Mill
Colorado
Canon City
Uravan
New Mexico
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Homes take
South Dakota
Edgemont
Texas
Fanna Maria
Conquista
Ray Point
Utah
White Mesa
Rio Algom
Moab
Shoo tar ing
Washington
Dawn
Sherwood
Wyoming
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Radon Maximum Maximum Lifetime
Concentration Exposure Fatal Cancer Risk Distance^)
(pCi/1) (WL) to Individual (meters)
2.8E-4
1.3E-4
6.1E-3
1.2E-2
1.1E-2
2.3E-3
2.9E-2
2.6E-3
7.1E-2
1.2E-2
3.1E-3
1.9E-4
1.3E-3
1.6E-2
2.6E-4
1.2E-2
1.9E-3
6.3E-4
8.4E-3
6.9E-4
2.8E-4
1.1E-3
2.6E-4
7.9E-4
4.1E-4
3.9E-3
1.1E-6
6.4E-7
2.4E-5
4.1E-5
4.4E-5
1.2E-5
9.5E-5
l.OE-5
2.1E-4
3.9E-5
1.1E-5
1.3E-6
5.7E-6
5.9E-5
1.1E-6
3.7E-5
7.4E-6
4.4E-6
3.1E-5
4.7E-6
1.8E-6
7.8E-6
1.8E-6
5.1E-6
2.7E-6
1.6E-5
2E-6
9E-7
3E-5
6E-5
6E-5
2E-5
1E-4
1E-5
3E-4
5E-5
2E-5
2E-6
8E-6
8E-5
2E-6
5E-5
1E-5
6E-6
4E-5
6E-6
2E-6
1E-5
2E-6
7E-6
4E-6
2E-5
3,500
7,500
3,500
1,500
3,500
7,500
1,500
3,500
750
1,500
2,500
25,000
4,500
2,500
4,500
750
3,500
25,000
2,500
25,000
15,000
25,000
25,000
15,000
15,000
3,500
(a) Distance from center of a homogenous circular equivalent impoundment
to the point where the exposures and risks were estimated.
9-17
-------
Table 9-10.
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around licensed tailings
impoundments post-UMTRCA disposal.
Mill
Fatal Cancers per Year
Colorado
New Mexico
South Dakota
Texas
Utah
Washington
Wyoming
Canon City
Uravan
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Homestake
Edgemont
Panna Maria
Conquista
Ray Point
White Mesa
Rio Algom
Moab
Shootaring
Dawn
Sherwood
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Total
4.3E-4
4.2E-5
4.2E-3
1.5E-3
4.3E-3
2.7E-3
3.8E-3
3.7E-4
l.OE-2
1.7E-2
5.2E-4
9.1E-5
2.5E-4
1.3E-3
6.5E-6
1.3E-3
1.1E-3
3.1E-4
3.2E-4
3.3E-4
2.8E-4
9.2E-4
5.3E-5
6.8E-4
1.9E-4
4.5E-4
5.2E-2
9-18
-------
Table 9-11. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from licensed
uranium mill tailings piles post-UMTRCA disposal.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0 0 0
1E-2 to 1E-1 0 0
1E-3 to 1E-2 0 0
1E-4 to 1E-3 75 1E-4
1E-5 to 1E-4 28,000 6E-3
1E-6 to 1E-5 1,200,000 3E-2
< 1E-6 3,200,000 2E-2
Totals* 4,500,000 5E-2
* Totals may not add due to independent rounding.
9-19
-------
Table 9-12. Estimated exposures and risks to individuals living near licensed
tailings impoundments post-disposal to 6
Maximum
State/Mill
Colorado
Canon City
Uravan
New Mexico
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Home stake
South Dakota
Edgemont
Texas
Panna Maria
Conquista
Ray Point
Utah
White Mesa
Rio Algom
Moab
Shootaring
Washington
Dawn
Sherwood
Wyoming
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Radon Maximum Maximum Lifetime
Concentration Exposure Fatal Cancer Risk Distance (a)
(pCi/1) (WL) to Individual (meters)
2.8E-4
1.3E-4
1.8E-3
3.6E-3
3.3E-3
6.9E-4
8.5E-3
7.9E-4
2.1E-2
3.5E-3
9.2E-4
1.6E-4
3.9E-4
4.7E-3
7.8E-5
7.6E-3
5.7E-4
1.9E-4
2.5E-3
2.1E-4
8.4E-5
3.3E-4
7.7E-5
2.3E-4
1.2E-4
1.2E-3
(a) Distance from center of a homogenous
to the point where
1.1E-6
6.4E-7
7.2E-6
1.2E-5
1.3E-5
3.5E-6
2.8E-5
3.2E-6
6.3E-5
1.1E-5
3.4E-6
1.1E-6
1.7E-6
1.7E-5
3.3E-7
2.3E-5
2.3E-6
1.3E-6
9.3E-6
1.4E-6
5.5E-7
2.3E-6
5.4E-7
1.5E-6
8.1E-7
4.9E-6
circular
2E-6
9E-7
1E-5
2E-5
2E-5
5E-6
4E-5
4E-6
9E-5
2E-5
5E-6
1E-6
2E-6
2E-5
5E-7
3E-5
3E-6
2E-6
1E-5
2E-6
7E-7
3E-6
7E-7
2E-6
1E-6
7E-6
equivalent
3,500
7,500
3,500
1,500
3,500
7,500
1,500
3,500
750
1,500
2,500
25,000
4,500
2,500
4,500
750
3,500
25,000
2,500
25,000
15,000
25,000
25,000
15,000
15,000
3,500
impoundment
the exposures and risks were estimated.
9-20
-------
Table 9-13,
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around licensed tailings
impoundments post-disposal to 6 pCi/m2/s.
Texas
Utah
Washington
Wyoming
Mill
Fatal Cancers per Year
Colorado
New Mexico
South Dakota
Canon City
Uravan
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Homestake
Edgemont
4.3E-4
4.2E-5
1.2E-3
4.4E-4
1.3E-3
8.0E-4
1.1E-3
1.1E-4
Panna Maria
Conquista
Ray Point
White Mesa
Rio Algom
Moab
Shootaring
Dawn
Sherwood
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Total
3.0E-3
4.9E-3
1.7E-4
7.6E-5
7.6E-5
3.8E-4
2.0E-6
8.1E-4
3.5E-4
l.OE-4
9.7E-5
l.OE-4
8.4E-5
2.8E-4
1.6E-5
2.0E-4
5.8E-5
1.4E-4
1.6E-2
9-21
-------
Table 9-14. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from licensed
uranium mill tailings piles post-disposal to
6 pCi/m2/s.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0 0 0
1E-2 to 1E-1 0 0
1E-3 to 1E-2 0 0
1E-4 to 1E-3 0 0
1E-5 to 1E-4 520 2E-4
1E-6 to 1E-5 110,000 4E-3
< 1E-6 4,400,000 1E-2
Totals* 4,500,000 2E-2
* Totals may not add due to independent rounding.
9-22
-------
Table 9-15. Estimated exposures and risks to individuals living near licensed
tailings impoundments post-disposal to 2
Maximum
State/Mill
Colorado
Canon City
Uravan
New Mexico
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Homes take
South Dakota
Edgemont
Texas
Panna Maria
Conquista
Ray Point
Utah
White Mesa
Rio Algom
Moab
Shoo tar ing
Washington
Dawn
Sherwood
Wyoming
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Radon Maximum Maximum Lifetime
Concentration Exposure Fatal Cancer Risk Distance^)
(pCi/1) (WL) to Individual (meters)
2.8E-4
1.3E-4
6.1E-4
1.2E-3
1.1E-3
2.3E-4
2.7E-3
2.6E-4
7.1E-3
1.2E-3
3.1E-4
5.1E-5
1.3E-4
1.6E-3
2.6E-5
2.6E-3
1.9E-4
6.3E-5
8.4E-4
6.8E-5
2.8E-5
1.1E-4
2.6E-5
7.9E-5
4.1E-5
3.9E-4
(a) Distance from center of a homogenous
to the point where
the exposures and
1.1E-6
6.4E-7
2.4E-6
4.1E-6
4.4E-6
1.2E-6
9.5E-6
l.OE-6
2.1E-5
3.9E-6
1.1E-6
3.6E-7
5.7E-7
5.9E-6
1.1E-7
7.6E-6
7.4E-7
4.4E-7
3.1E-6
4.7E-7
1.8E-7
7.8E-7
1.8E-7
5.1E-7
2.7E-7
1.6E-6
circular
2E-6
9E-7
3E-6
6E-6
6E-6
2E-6
1E-5
1E-6
3E-5
5E-6
2E-6
5E-7
8E-7
8E-6
2E-7
1E-5
1E-6
6E-7
4E-6
6E-7
2E-7
1E-6
2E-7
7E-7
4E-7
2E-6
equivalent
3,500
7,500
3,500
1,500
3,500
7,500
1,500
3,500
750
1,500
2,500
25,000
4,500
2,500
4,500
750
3,500
25,000
2,500
25,000
15,000
25,000
25,000
15,000
15,000
3,500
impoundment
risks were estimated.
9-23
-------
Table 9-16,
State
Estimated fatal cancers per year in the regional
(0-80 km) populations around licensed tailings
impoundments post-disposal to 2 pCi/m2/s.
Mill
Fatal Cancers per Year
Colorado
New Mexico
South Dakota
Texas
Utah
Washington
Wyoming
Canon City
Uravan
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Homestake
Edgemont
Panna Maria
Conquista
Ray Point
White Mesa
Rio Algom
Moab
Shootaring
Dawn
Sherwood
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Total
4.3E-4
4.2E-5
4.2E-4
1.5E-4
4.3E-4
2.7E-4
3.8E-4
3.7E-5
l.OE-3
1.7E-3
5.2E-5
2.5E-5
2.5E-5
1.3E-4
6.5E-7
2.7E-4
1.1E-4
3.1E-5
3.2E-5
3.3E-5
2.8E-5
9.2E-5
5.3E-6
6.8E-5
1.9E-5
4.5E-5
5.8E-3
9-24
-------
Table 9-17. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) populations from licensed
uranium mill tailings piles post-disposal to
2 pCi/m2/s.
Risk Interval Number of Persons Deaths/y
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
to
to
to
to
to
to
1E+0
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1E-2
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1E-4
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0
0
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0
0
0
0
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* Totals may not add due to independent rounding.
9-25
-------
9.4. SUPPLEMENTARY CONTROL OPTIONS AND COSTS
Previous studies have examined the feasibility, effective-
ness, and cost associated with various options for controlling
releases of radioactive materials from uranium mill tailings
(NRC80, EPA82, EPA83, EPA86). These studies have concluded that
long-term stabilization and control will be required to protect
the public from the hazards associated with these tailings. The
standards for long-term disposal established for these licensed
sites under the UMTRCA provide for controls to prevent misuse of
the tailings, protect water resources, and limit releases of
radon-222 to the air. The UMTRCA standard established a design
standard to limit long-term radon releases to an average flux not
to exceed 20 pCi/m2/s. In addition, the NESHAP promulgated
under Section 112 of the Clean Air Act provides for the phasing
out of existing tailings impoundments by 1992 and for all future
tailings to be disposed of either continuously or in a phased
disposal impoundment.
In this section, the costs of long-term isolation of both
existing and future tailings impoundments are evaluated.
9.4.1 Control Options for Existing Licensed Tailings
Impoundments
For the reasons described in Chapter 8, the control selected
for long-term radon-222 control at existing licensed tailings im-
poundments is the earth cover option.
9.4.1.1 Cost Estimates for Earthen Covers
As in the case of inactive tailings, the cost estimates de-
veloped below consider covers designed to meet three radon emis-
sion levels: 20 pCi/m2/s (the level established by the UMTRCA
standard), 6 pCi/m2/s, and 2 pCi/m2/s. The basis for the ef-
fectiveness of various depths of cover and the unit costs used in
this analysis are documented in the "Radon Attenuation Handbook
for Uranium Mill Tailings Cover Design" (Ro84) and Appendix B,
"Generic Unit Costs for Earth Cover Based Radon-222 Control Tech-
niques. "
Even though existing impoundments may still be in use or on
standby with additional available capacity, the control options
evaluated in this analysis are based on the simplifying assump-
tion that operations have ceased, that the tailings are dry
enough to allow the use of heavy equipment, and that the piles
have their current dimensions.
The thickness of cover required to achieve a given radon
flux is a function of the soil type and the initial radon flux
from the pile. In this assessment, soil type B (see Appendix B)
is assumed. Table 9-18 presents the current radon flux rate at
each pile and the estimated thickness of cover needed to achieve
each of the three levels.
9-26
-------
Table 9-18. Estimated depths of earth cover needed to achieve given radon
flux rates.(a'
_ f
Colorado
Canon City
Uravan
New Mexico
L-Bar
Churchrock
Bluewater
Ambrosia Lake
Home stake
South Dakota
Edgemont
Texas
Panna Maria
Conquista
Ray Point
Utah
White Mesa
Rio Algom
Moab
Shootaring
Washington
Dawn
Sherwood
Wyoming
Lucky Me
Split Rock
Umetco
Bear Creek
Shirley Basin
Sweetwater
Highland
FAP
Petrotomics
Radon Flux
(pCi/m2/s)
400
480
500
290
305
416
300
.560
196
224
520
981
420
540
280
240
200
220
100
364
85
275
280
450
420
570
of Pile
-------
Five basic steps or operations are required to place earthen
covers on uranium tailings piles. These are: regrading the
slopes of the pile to achieve long-term stability, procuring and
placing the dirt cover, placing gravel on the pile tops, placing
riprap on the pile sides, and reclaiming the borrow pits. A pre-
liminary step, reclaiming radium-bearing materials from evapora-
tion ponds and regrading the ponds, is required at sites where
the tailings water was decanted to evaporation ponds.
The total cost of excavating evaporation ponds is calculated
by multiplying the volume of waste material by the unit cost of
$6.01 per cubic yard for excavation, hauling, spreading, and com-
pacting. The derivation of this unit cost is given in Appendix B.
Once all of the contaminated materials are placed on the
pile, the pile is regraded, as necessary, to prepare for the
placement of the dirt cover. It is assumed that existing piles
have a slope of 2:1 and that the placement of a dirt cover re-
quires a slope no greater than 5:1 (EPA86). The total cost for
this operation is the product of the volume regraded and the unit
cost of grading. The volumes to be regraded are based on the set
of equations presented in Appendix B and two additional assump-
tions about the geometric configuration of the piles. First, it
is assumed that the length of each base side of the pile is the
square root of the area of the pile. Second, it is assumed that
the ratio between the height and base side lengths of the piles
is equal to 40 feet of height per 2,100 feet in base side length.
The unit cost of regrading is $1.36 per cubic yard.
The third step is the procurement and placement of the
earthen cover. As in the case of inactive tailings piles (see
Chapter 8), it is assumed that dirt is available onsite at an
average distance of one mile from the pile (two miles round
trip). The cost of the dirt cover is the product of the volume
required and unit costs for excavating (on trucks), hauling,
spreading, and compacting. The volume is estimated by multiply-
ing the surface area of the pile (including the sides) by the
depth of cover required to meet each of the three alternative
radon flux rates. The equations used to estimate surface areas,
cover depths, and the total unit cost of $6.01 per cubic yard for
excavation, hauling, spreading, and compacting are documented in
Appendix B.
The fourth and fifth steps are erosion controls required to
provide long-term stabilization, after the final earthen cover
has been put in place. The erosion control system is an essen-
tially maintenance-free gravel and rock system designed for arid
conditions. In this system, gravel is placed on the top of the
pile, and riprap (random broken stone) is placed on the sides of
the pile. The cost of each is a product of surface area, depth,
and unit costs. The depth required for adequate erosion protec-
tion is assumed to be one-half yard (EPA86). The equations used
to calculate the relevant surface areas, and the unit costs of
$7.55 per cubic yard for gravel and $23.00 per cubic yard for
riprap are documented in Appendix B.
9-28
-------
The final operation is the reclamation of the borrow pits,
where the earthen cover is extracted. The cost of borrow pit re-
clamation is assumed to include regrading the sides of the pits
from 2:1 to 8:1. Regrading of the pit is calculated using the
same methodology used for estimating pile regrading. The volume
of the pit is based on the volume of dirt required for cover.
The ratio of height to base side length is the same as given
above, as is the unit cost for grading.
Tables 9-19 through 9-21 summarize the costs of achieving
the alternative levels of control. The total cost of achieving
the 20 pCi/m2/s opti.on at all sites is approximately $599
million. The estimated total costs at all sites for the 6 and 2
pCi/m2/s options are approximately $779 million and $944
million, respectively. These costs, as discussed in Appendix B,
include an overhead and profit factor of 7 percent.
The cost methodology, described above, assumes no previous
cover operations have been initiated on the individual piles.
However, as shown in Table 9-1, cover operations are preceding
and/or have been completed at a number of these sites. In
estimating the costs of achieving the alternative fluxes, no
attempt has been made to include the costs of possible redesign
and re-work that would be required if a lower flux limit has to
be achieved at these piles.
9-29
-------
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9.4.1.2 Effectiveness of the Earth Cover Control Options
Once all piles have been disposed of in accordance with the
current designs under the UMTRCA standard, it is estimated that
that maximum individual's lifetime fatal cancer risk will be 3E-4
(three chances in 10,000) and that the emissions from all piles
will cause approximately one death every 20 years (5E-2 deaths
per year) in the population of 4.5 million persons living within
80 kilometers of these sites.
At the alternative 6 pCi/m2/s flux limit, it is estimated
that the maximum individual's lifetime fatal cancer risk would be
reduced by a factor of approximately three, to 9E-5 (9 chances in
100,000). Similarly, the deaths per year in the regional popula-
tion would be reduced to approximately 2E-2 (one death every 50
years). Adopting the alternative 2 pCi/m2/s flux limit would
achieve another factor of three reduction in risks. The maximum
individual risk at 2 pCi/m2/s is estimated to be reduced to 3E-5,
and the deaths per year are estimated to be reduced to 6E-3.
9.4.2 Work Practices for New Tailings Impoundments
Tailings impoundments constructed in the future must, at
minimum, meet current Federal standards for prevention of ground-
water contamination and airborne particulate emissions. The
baseline tailings impoundment will have a synthetic liner, be
built partially below grade, and have earthen dams or embankments
to facilitate decommissioning.* A means for dewatering the tail-
ings after the area is filled should also be incorporated. This
conventional design allows the maintenance of a water cover over
the tailings during the milling and standby periods, thus main-
taining a very low level of radon-222 emissions. Dewatering of
the tailings can be accelerated using built-in drains. A syn-
thetic liner is placed along the sides and bottom. Cover mater-
ial may be added after the impoundment has reached capacity or is
not going to be used further and the tailings have dried. Two
alternatives to the work practices assumed in this baseline model
new tailings impoundment are evaluated in the following sections.
9.4.2.1 Phased Disposal
The first alternative work practice being evaluated for mod-
el new tailings impoundments is phased disposal. In phased or
multiple cell disposal, the tailings impoundment area is parti-
tioned into cells which are used independently of other cells.
After a cell has been filled, it can be dewatered and covered,
and another cell used. Tailings are pumped to one initial cell
* It may in some cases be feasible to replace synthetic with clay
liners. This option, however, is not evaluated here. In addi-
tion, it is possible but not cost-effective to construct below-
grade tailings impoundments. Section 9.4.3 provides a comparison
of the cost-effectiveness of below-grade versus partially below-
grade impoundments.
9-33
-------
until it is full. Tailings are then pumped to a newly construct-
ed second cell, and the first cell is dewatered and then left to
dry. After the first cell dries, it is covered with earth ob-
tained from the construction of a third cell. This process is
continued sequentially. This system minimizes emissions at a
given time since a cell can be covered after use without inter-
fering with operation as opposed to the case of a single cell.
Less total surface area is thus exposed at any one time.
Phased disposal is effective in reducing radon-222 emissions
since tailings are initially covered with water and finally with
earth. Only during a drying-out period of about 5 years for each
cell are there any radon-222 emissions from a relatively small
area. During mill standby periods, a water cover could be main-
tained on the operational cell. For extended standby periods,
the cell could be dewatered and a dirt cover applied.
Radon emissions from a model six-cell phased disposal im-
poundment are estimated to be 13.5 kCi during the 20-year operat-
ing life of the impoundment (EPA86). The 13.5 kCi of radon re-
leased during the operating period is about 55 percent of the
24.5 kCi estimated to be released from the baseline single-cell
impoundment (EPA86). Once the phased disposal impoundment is
filled and covered with three meters of soil, annual radon-222
releases are estimated to be 0.33 kci/y, comparable to the esti-
mated releases (0.3 kCi/y) for a single-cell impoundment covered
with the same depth of soil.
9.4.2.2 Continuous Disposal
The second alternative work practice, continuous disposal,
is based on removal of water from the tailings slurry prior to
disposal. The relatively dry dewatered (25 to 30 percent moist-
ure) tailings can then be dumped and covered with soil almost im-
mediately. No extended drying phase is required, and therefore
very little additional work would be required during final clo-
sure. Additionally, groundwater problems are minimized. To
implement a dewatering system requires added planning, design,
and modification of current designs. Additional holding ponds
with ancillary piping and pumping systems would be required to
handle the liquid removed from the tailings. Using trucks or
conveyor systems to transport the tailings to disposal areas
might also be more costly than slurry pumping. Thus, although
tailings are more easily managed after dewatering, this practice
would have to be carefully considered on a site-specific basis.
Various filtering systems such as rotary vacuum and belt
filters are available and could be adapted to a tailings dewater-
ing system. Experimental studies would probably be required for
a specific ore to determine the filter media and dewatering prop-
erties of the sand and slime fractions. Modifications to the
typical mill ore grinding circuit may be required to allow effi-
cient dewatering and to prevent filter plugging or blinding.
Corrosion-resistant materials would be required in any tailings
9-34
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dewatering system due to the highly corrosive solutions that must
be handled. Continuous tailings dewatering is not practiced at
any uranium mills in the United States, but it has been proposed
at several sites in the southwestern and eastern part of the
country (MA83). Tailings dewatering systems have been used suc-
cessfully at nonferrous ore beneficiation mills in the United
States and Canada (Ro78).
Radon emissions from a model continuous disposal (single-
cell) impoundment are estimated to be 7.5 kCi during the 15-year
operating life of the impoundment (EPA86). The 7.5 kCi of radon
released during the operating period is about 30 percent of the
24.5 kCi estimated to be released from the baseline single-cell
impoundment (EPA86). Much of this reduction is attributable to
the fact that the 5-year drying out period (when much of the
radon-222 is released) is avoided with a continuous disposal
system. Once the continuous disposal impoundment is filled and
covered with three meters of soil, annual radon-222 releases are
estimated to be 0.3 kCi/y, the same as the releases estimated
(0.3 kCi/y) for single-cell impoundment covered with the same
depth of soil.
9.4.3 Comparison of Control Options for New Tailings
Impoundments
To meet current Federal radon-222 emission standards, new
tailings areas will have synthetic liners with either earthen
dams or embankments, and also incorporate a means of dewatering
the tailings at final closure. These new tailings can either be
stored below or partially above grade. Although below-grade
storage provides the maximum protection from windblown emissions
and water erosion and eliminates the potential for dam failure,
it is not cost-effective compared to partially above-grade dis-
posal technology and has a greater potential for contaminating
groundwater.
Previous analysis of work practices for new model tailings
impoundments has estimated costs and radon releases for a number
of alternative control technologies (EPA86). These estimated
costs are listed in Table 9-22. The estimated radon releases are
summarized in Table 9-23. These estimates suggest that storage
of tailings piles partially above grade is cost-effective, when
compared to fully below-grade designs. Completely below-grade
designs are estimated, on average, to increase costs by 20
percent.
Partially below-grade piles have been shown to be cost-
effective compared to above-grade impoundments. Excavation costs
for the final dirt cover are incurred in both cases. Using the
excavated pit from which the earth cover is taken to store
tailings provides no-cost benefits in terms of windblown
emissions, water erosion, and dam failure. In addition, dam
construction cost is minimized because the sides of the excavated
pit replace part of the dam.
9-35
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Table 9-22. Estimated total costs for new tailings control
technologies.(a)
(in Millions of 1985 Dollars)
Technology Below Grade Partially Below Grade
Single Cell 41.33 29.71
Phased Disposal
Six Cells 47.78 41.54
Continuous Disposal
Trench Design 54.16 47.75
Single Cell Design NA 37.44
(a) Based on comparable dimensions for cells,
Source: EPA86
Table 9-23. Summary of estimated radon-222 emissions for new tailings control
technologies.(a)
Operational Emissions Post-Operational Emissions Cumulative
(kCi/y) (kCi/y) Total Emissions
Technology
Single Cell NA 0.30 9 15 21
NA - Not Applicable.
(a) Emissions estimates based on 280 pCi/g Ra-226 and a specific flux of
1 pCi/m2/s per pCi/g Ra-226.
(b) Final cover to meet 20 pCi/m2/s UMTRCA standard.
(c) Assumes 20 percent of impoundment area is dry beach during active phase.
(d) Assumes 15-year active life.
Adapted from EPA86.
9-36
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The 20 percent increase in costs for fully below-grade dis-
posal does not appear to be justified by additional benefits.
The increased costs are incurred for additional excavation. The
additional material is not needed for dirt cover, and the bulk of
the benefits to be derived from reducing windblown emissions,
water erosion, and dam failure have already been captured by the
partially below-grade design. Therefore, only designs that are
partially above-grade are considered.
Also dropped from consideration is the continuous trench
pile design. This technology has little operational advantage
over the continuous single cell design and is not cost-effective.
9.4.4 Engineering Design for New Model Tailings Impoundments
New tailings disposal impoundments at uranium mills can be
designed to incorporate current Federal regulations on radon-222
emissions. Three types of new model impoundments are considered:
Single-Cell, Phased Disposal, and Continuous Disposal. Engineer-
ing designs for each type of impoundment are discussed in the
following sections. These models will later be used to generate
cost estimates.
9.4.4.1 Model Single Cell Impoundment
The single cell impoundment can be constructed partially be-
low grade. The basic design and layout (consistent with earlier
uranium mill tailings studies) of a single cell impoundment, as-
suming a capacity of 1,800 tons per day, 310 working days, and a
15-year active life of the mill, are a square sloping pit (an in-
verted truncated pyramid) with a tailings depth of 12-meters, ex-
cluding a 3-meter final cover. Further, the final surface area
of this impoundment is 47 ha (116 acres), with a tailings capaci-
ty of 8.4 million tons and a tailings volume of 5.25 million
cubic meters.
The final surface area is obtained by taking the square of
the length at final cover (685 meters) and converting this value
into hectares, using appropriate rates of conversion. Tailings
capacity (in millions of tons) is the product of 1,800 tons per
day, 310 working days, and a 15-year active life of the mill.
Tailings volume is tailings capacity converted into meters, using
a conversion rate of 1.6 (EPA86).
The size, shape, and layout for a model single cell impound-
ment partially below grade are shown in Figures 9-1 and 9-2. The
model has a base with a width and length of 637 meters and a
slope of 2:1. The height to final cover is 12 meters, with a
length, at final cover, of 685 meters. Synthetic liners are
placed along the sides and bottom; tailings are stored 6 meters
each above and below grade; and earthen dams are constructed with
a berm 6 meters wide with a height of 9 meters, an outside slope
of 5:1, and an inside slope of 2:1.
9-37
-------
3m final
cover
tailings
Shape and dimensions of the single-cell impoundment
697m
685m
6
24m
Layout of model single-cell Impoundment
» Diagrams are not drawn to scale
Figure 9-1.
Shape and layout of the model single-cell
impoundment.
9-38
-------
erm
697m
6mf
dam
6m below
grade
synthetic liner
Diagram not drawn to scale.
Figure 9-2.
Size of partially above-grade model single cell
impoundment.
9-39
-------
305.7m
305.
3m final
couer ~
tailings
\ \
245.7m
Dimension of each cell of phased disposal impoundment
Layout of phased dlsposol impoundment
Diagrams are not drawn to scale.
Figure 9-4.
Shape and layout of model phased disposal
impoundment.
9-42
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Table 9-24. Unit cost categories for partially below-grade impoundments.
Cost Component Single Cell Phased Disposal Continuous Disposal
(all Cells) (Single Cell)
Excavation
Synthetic Liner
Grading
Drainage System
Dam Construction
Cover (3 meters)
Gravel Cap
Riprap
Evaporation Pond
Vacuum Filter
Indirect Cost
Adapted from EPA86.
Required
Required
Required
Required
Required
Required
Required
Required
Not Required
Not Required
Required
Required
Required
Required
Required
Required
Required
Required
Required
Required
Not Required
Required
Required
Required
Required
Not Required
Required
Required
Required
Required
Required
Required
Required
9-43
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Total costs for each design, shown in Tables 9-25 through
9-27, indicate that the phased partially above grade disposal im-
poundment is the most expensive design (about $54 million), while
the single cell partially above-grade impoundment (about $37 mil-
lion) is the least expensive. Costs for the continuous single
cell design (about $41 million) are marginally different from
those of the single cell impoundment, although the uncertainties
surrounding the technology used in this design are the largest.
The volumes or surface areas and the unit costs that were used in
calculating the cost figures are also provided in Tables 9-25 to
9-27. The equations used to calculate volumes and surface areas
are discussed in detail in Appendix B, as are the sources and
methodologies used to calculate unit costs. The assumptions and
rationales used in developing estimates for each cost category
are discussed in the following paragraphs.
For each design, costs for excavation are calculated by
multiplying the volume of the tailings cells that are below grade
by unit cost of excavation by 21 cubic yard scrappers for a 5,000
foot haul. It is assumed that the dirt is not hauled by truck,
but rather pushed aside for later use in dam construction and for
dirt cover.
Dam construction is required for each design, and the dams
are assembled during the excavation stage. Unit costs for dam
construction are a sum of costs for grading and compacting.
While unit costs for compacting are on a square unit rather than
a cubic unit basis, both are multiplied by the volume of the dam
because the dam materials must be compacted as each meter of ma-
terial is graded into place. This procedure insures stability of
the dam. The volumes of the dams are derived by calculating the
entire aboveground volume of the pile and dams and then subtract-
ing the aboveground volumes of the piles and their covers.
Synthetic liners are placed on the bottom and the sides of
the tailings impoundment. Cost for synthetic liners are derived
from the product of the unit post ($13.35 per square meter) and
surface areas of the interior of the cells, excluding the final
three meters where the dirt cover is placed. Design specific
volumes and surface areas are calculated using dimensions given
in Figures 9-1 through 9-4.
Evaporation ponds are required for both the phased disposal
and continuous single cell impoundments. Evaporation ponds are
used to regulate or control the water level in the waste impound-
ment. The surface area required for evaporation is assumed to be
equal to approximately one-third of the surface area of the
single impoundment or two of the phased disposal impoundments.
This assumption is based on the ratio of the surface areas of
evaporation ponds to the surface areas of tailings impoundments
at existing mills. Since phased piles will have only one cell in
operation at a time, this design requires an evaporation pond
with a surface area equal to the surface area of one cell. As
the continuous pile is assumed to store only dried tailings, it
9-44
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Table 9-25. Costs for a single cell partially below-grade new
model tailings impoundment.
Cost Component
(1988 Dollars)
Volume or Area
(m3 or m2)
Unit Cost
($/m3 or $/m2)
Total Cost
($ X 106)
Excavation 2,527,494
Grading 469,225
Cover
Grade
Compact
Total 1,432,479
Gravel Cap 251,341
Riprap 138,408
Dam Construction
Grade
Compact
Total 1,010,232
Synthetic Liner 442,405
Drainage System 641,089
Subtotal: Direct Cost
Indirect Cost @ 7%
Total Cost
4.92
1.78
1.78
1.49
3.27
9.87
30.07
1.78
1.49
3.27
13.35
0.60
12.42
0.83
4.68
2.48
4.16
3.30
5.91
0.38
34.17
2.39
36.56
9-45
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Table 9-26. Costs for a phased design, partially below-grade,
new model tailings impoundment.
Cost Component
(1988 Dollars)
Volume or Area
(m3 or m2)
Unit Cost
($/m3 or $/m2)
Total Cost
($ X 106)
Excavation 2,392,462
Grading 517,558
Cover
Grade
Compact
Total 1,616,978
Gravel Cap 442,835
Riprap 181,013
Dam Construction
Grade
Compact
Total 4,382,475
Synthetic Liner 451,901
Drainage System 1,066,682
Evaporation Pond
Excavate
Synthetic Liner
Total 88,387
Subtotal: Direct Cost
Indirect Cost @ 7%
Total Cost
4.92
1.78
1.78
1.49
3.27
9.87
30.07
1.78
1.49
3.27
13.35
0.60
4.91
14.59
19.50
11.76
0.92
5.28
4.37
5.44
14.32
6.03
0.64
1.72
50.49
3.53
54.02
9-46
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Table 9-27.
Costs for a continuous design, partially below-
grade, new model tailings impoundment.
Cost Component
(1988 Dollars)
Volume or Area Unit Cost Total Cost
(m3 or m2) ($/m3 or $/m2) ($ X 106)
Excavation 2,527,494
Grading 469,225
Cover
Grade
Compact
Total 1,432,479
Gravel Cap 251,341
Riprap 138,408
Dam Construction
Grade
Compact
Total 1,010,232
Synthetic Liner 442,405
Evaporation Pond
Excavate
Synthetic Liner
Total 176,775
Vacuum Filter NA
Subtotal: Direct Cost
Indirect Cost @ 7%
Total Cost
4.92
1.78
1.78
1.49
3.27
9.87
30.07
1.78
1.49
3.27
13.35
4.91
14.59
19.50
NA
12.42
0.83
4.68
2.48
4.16
3.30
5.91
3.45
0.92
38.15
2.26
40.83
9-47
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will require an evaporation pond twice the size of that required
by the phased disposal design. Evaporation ponds are assumed to
be excavated to a 1-meter depth and to employ a synthetic liner
to protect groundwater.
Once each cell is filled it is assumed that the tailings are
graded prior to cover. Grading volume is assumed to be a product
of the surface area of the top portion of the pile and a depth of
1 meter.
Costs for earthen cover are based on a depth of 3 meters and
unit costs for grading and compacting. It is assumed, as was the
case for dam construction, that compacting is done after each
meter of dirt is put in place.
Riprap and gravel caps are needed for erosion control and
are required to maintain long-term stability of the tailings im-
poundment. Typically, gravel is placed on the top of the pile
and rock (riprap) is placed on the sides of the pile. The cost
of each is the product of surface area, depth, and unit costs.
The depth required for adequate erosion protection is assumed to
be one-half meter (EPA86). Equations for calculating the rele-
vant surface areas, and the unit costs for gravel cap ($9.87 per
cubic meter) and riprap ($30.07 per cubic meter) are given in
Appendix B.
Except for the continuous single cell impoundment for which
the tailings are dried prior to disposal, all other designs re-
quire a drainage system. Costs for drainage systems are $0.60
per square meter, for both the single cell and phased disposal
impoundments. The surface area is assumed to be the entire
above-ground surface area of the pile.
Vacuum filters are required to dewater tailings in the con-
tinuous single cell impoundment. Dewatering and continuously
covering tailings is an attractive but untried method for tail-
ings disposal. Tailings dewatering systems have been used suc-
cessfully at nonferrous ore beneficiation mills in tjie United
States and Canada (EPA86). Several uranium mills have proposed
the use of continuous disposal systems. For example, Pioneer
Uravan, Inc., submitted plans to build the San Miguel Mill using
continuous tailings disposal at Slick Rock, Colorado (NRC80).
The planned tailings disposal operation consisted of below-grade
burial of horizontal belt filtered tailings in a series of ten
trenches. The mill, however, has not been constructed. An ad-
vantage of dewatering the tailings slurry prior to disposal is
that the tailings can be placed and covered with soil immediate-
ly. Thus, no extended dry phase is necessary, and groundwater
problems are reduced.
To implement a dewatering system, factors such as added
placing, design, and modification of current designs should be
evaluated. Further, adaptation of horizontal belt vacuum fil-
ters, to enhance the capability of the dewatering system, should
9-48
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also be considered. A horizontal belt vacuum filter basically
filters sand and slimes fractions from the tailings slurry.
Previous studies provide costs for such systems, but these
reported costs have not been consistent with each other. For ex-
ample, the cost estimate ranges from $1.46 million (in 85 dollars
(EPA86), to $465,000 (NRC80). Given this discrepancy, manufac-
turers sales representatives were contacted to provide current
cost estimates (EC88). Their estimate, based on a 316 frame with
a carbon frame for the wetted part and including auxiliary parts,
is $845,000. Costs for transportation and installation are ex- ,
eluded. Freight costs depend on location of site and are assess-
ed at $15,000 for sites in Arizona (EC88). Installation costs,
based on installation costs for similar equipment, are assumed to
be 7.5 percent of the cost of the horizontal belt vacuum filter.
Therefore, the total cost for this equipment, installed, is cal-
culated at $923,375.
Added to the cost of operations, as described above, is and
overhead and profit factor estimated at 7 percent. The calcula-
tion of this factor is described in Appendix B.
9.4.6 Work Practices at Existing Operable Impoundments
Radon releases during the operating and standby periods at
existing operable impoundments can be reduced by active controls
that minimize the area of the tailings that are dry and exposed.
Unlike the case of long-term isolation, where active institution-
al controls are not deemed to be reliable, active controls during
the operable phase of a mill can be assured simply by making them
a condition of the facility's license. Two active techniques
have been identified to minimize the area of dry tailings at ex-
isting impoundments: water and earthen covers.
As noted in Section 9.2 (see also Chapter 8), both water and
earthen covers can efficiently attenuate the radon generated in
the tailings. Thus, maximizing the extent of the tailings pond,
maintaining the moisture content in the exposed tailing at or
near the saturation point, and/or placing earth covers on por-
tions of the impoundment that are filled and/or inactive can re-
sult in a significant reduction in radon releases. Table 9-2
shows the extent to which these managment practices are currently
used at the 11 operable impoundments. Portions of the tailings
are either ponded or wet at all of the mills, and earthen covers
have been placed on portions of the operable impoundments at the
Panna Maria, Ambrosia Lake, and Lucky Me mills. While the extent
of control varies from mill to mill, the combined ponded, wet,
and covered acreages at all 11 mills represents almost 75 percent
of the the total impoundment areas.
To evaluate the potential effectiveness of these management
options, an estimate was made for each mill of the extent of
cover necessary to achieve a flux averaged over all areas of the
impoundment equal to the UMTRCA disposal limit of 20 pCi/m2/s.
9-49
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For both water and earth covers, the estimate assumes complete
attenuation of the the radon from the covered areas. Given ac-
tive control, virtually complete attenuation is achievable if the
wet tailings and interim earth covers are maintained at or near
the saturation point.
Site-specific design and other factors will determine the
work practice or combination of practices selected at a given
mill. However, to evaluate the costs associated with these work
practices a single control, either wetting or earth cover, was
assigned to each mill. Wetting was assigned as the work practice
at mills where the impoundments are lined with clay or a synthet-
ic liner. At these mills, the addition of water to the tailings
should not result in the degradation of groundwater. At mills
that lack such a liner, earthen covers were selected.
Table 9-28 shows the extent of coverage that would be re-
quired to achieve an average flux of 20 pCi/m2/s at each of the
operable mills. At three sites, Chevron's Panna Maria mill,
Cotter's Canon City mill, and Rio Algom's La Sal mill, no change
from existing practice would be required to achieve an average
flux of 20 pCi/m2/s. At the Shootaring mill, with only seven
acres of tailings, achieving an average flux of 20 pCi/m2/s
may not leave sufficient beach area to allow future disposal
operations. Thus, unless the work practice applies to the
licensed impoundment area rather than the current tailings area,
the Shootaring mill would have to close.
Costs of the alternative work practices of additional
wetting and partial cover with earthen covers have been estimated
based on the additional areas to be controlled shown in Table 9-
28. For sites where the water option was selected, the costs are
based on the net evaporation rate for the site and maintaining
the moisture of the controlled areas at 20 percent water. Since
sprinkling systems and/or water trucks are already in place, no
capital costs for this equipment are assessed. At the sites
where earthen covers are needed, the costs include both the costs
of placing the earthen covers and the cost of additional water to
maintain the covers near the saturation point. The total
annualized costs, assuming a 5 percent real interest rate, for
these work practices are estimated to be $1.25 million/year.
The risks that will remain when these work practices are
implemented will be roughly comparable to the risks that are
estimated for the piles post-UMTRCA disposal. As an example, for
the Sherwood mill, the lifetime fatal cancer risk from all phases
of operations (see Table 9-6) is 3E-5. This would be reduced to
approximately 1E-5 when operating controls that meet the long-term
disposal emissions limits are implemented.
9-50
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Table 9-28. Additional areas of operable impoundments to be
controlled to achieve average radon-222 flux of 20
pCi/m2/s.
Additional
Current Conditions (acres) Area to be
Controlled
State/Impoundment Total Covered Ponded Wet Dry (acres)
Colorado
Canon City -
New Mexico
Ambrosia
Ambrosia
Ambrosia
Lake
Lake
Lake
Total
- Secondary
- Evap . Ponds
- Total
130
121
280
401
0
13
0
13
128
0
162
162
2
0
0
0
0
108
118
226
0
75
118
193
Homestake - Total 210 40 100 0 70 5*
Texas
Panna Maria 160 80 40 40 0 0
Utah
White Mesa
Rio Algom - Lower
Shootaring
Washington
130
47
7
0
0
0
55
18
2
70
29
1
5
0
4
2.4
0
3.5
Sherwood
Wyoming
Lucky Me - Pile 1-3
Lucky Me - Evap . Ponds
Lucky Me - Total
Shirley Basin
Sweetwater
* Based on the reported 65
80
203
104
307
275
37
pCi/g in the
0
108
0
108
0
0
dry exposed
0
35
104
139
179
40
0
0
0
36
30 0
tailings.
40
60
0
60
60
7
32
32
0
32
34
4.4
9-51
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9.5 REFERENCES
EC88 Enviro-Clear Division of Amstar Corporatioii, communication
with sales representative, August 30, 1988.
EPA82 U.S. Environmental Protection Agency, "Final Environmental
Impact Statement for Remedial Action Standards for
Inactive Uranium Processing Sites (40 CFR 192)," Vol.1,
EPA 520/4-82-013-1, Office of Radiation Programs,
Washington, DC, October 1982.
EPA83 U.S. Environmental Protection Agency, "Final Environmental
Impact Statement for Standards for the Control of By-
product Materials from Uranium Ore Processing (40 CFR
192)," Vol.1, EPA 520/1-83-008-1, Office of Radiation
Programs, Washington, DC, 1983
EPA86 U.S. Environmental Protection Agency, "Final Rule for
Radon-222 Emissions from Licensed Uranium Mill Tailings,"
EPA 520/1-86-009, Office of Radiation Programs,
Washington, DC, August 1986.
MA83 Marline Uranium Corp. and Union Carbide Corp., "An
Evaluation of Uranium Development in Pittsylvania County,
Virginia," October 15,1983.
NRC80 U.S. Nuclear Regulatory Commission, "Final Generic
Environmental Impact Statement on Uranium Milling," NUREG-
0706, Washington DC, September 1980.
PNL84 Pacific Northwest Laboratory, "Estimated Population Near
Uranium Tailings," PNL-4959, WC-70, Richland, WA, January
1984.
Ro78 Robinsky, E.I., "Tailing Disposal by the Thickened
Discharge Method for Improved Economy and Environmental
Control," in Volume 2, Proceedings of the Second
International Tailings Symposium. Denver, CO, May 1978.
Ro84 Rogers, V.C.; Neilson, K.K.; and Kalkwarf, D.R., "Radon
Attenuation Handbook for Uranium Mill Tailings Cover
Design," NUREG/CR-3533, prepared for the U.S. Nuclear
Regulatory Commission, Washington, DC, April 1984.
9-52
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10. DEPARTMENT OF ENERGY RADON SITES
The Department of Energy (DOE) radon source category
comprises sites owned or controlled by the Federal government and
operated or maintained under the authority of the DOE where
significant quantities of radium-bearing wastes are located.
These wastes, which include pitchblende residues, uranium and
thorium wastes, contaminated soils, and uranium mill tailings,
release radon-222 and radon-220 to the atmosphere.
Five DOE radon sites are known: (1) the Feed Materials
Production Center (FMPC), Fernald, Ohio; (2) the Niagara Falls
Storage Site (NFSS), Lewiston, New York; (3) the Weldon Spring
Site (WSS), Weldon Spring, Missouri; (4) the Middlesex Sampling
Plant (MSP), Middlesex, New Jersey; and (5) the Monticello
Uranium Mill Tailings Pile (MUMT), Monticello, Utah.
EPA characterized these five sites in 1984 in support of
the previous radionuclide NESHAPS rulemaking (EPA84). Since that
time, DOE has taken extensive interim remedial actions and has
begun an ongoing remedial action and long-term stabilization
program. The information presented in this chapter is based on
recent environmental monitoring, radiological surveys, hazard
characterizations, engineering evaluations, environmental
assessment reports, safety analysis reports, environmental
statements, and remedial investigation/feasibility studies
prepared for the DOE facilities. In addition, cognizant DOE
personnel clarified and confirmed the current status of remedial
actions.
Remedial actions and long-term stabilization programs
currently being planned or implemented comply with the design
standard of 20 pCi/m2/s in 40 CFR Part 192. Since many of
these remedial actions are scheduled for completion in the near
future, in addition to an assessment of the risks from the
current radon emission rates, an assessment is presented for
post-remediation emission rates. Post-remediation emission rates
are assumed to be the lesser of either 20 pCi/m2/s or the
current emission rate.
10.1 SITE DESCRIPTIONS
10.1.1 The Feed Materials Production Center
The FMPC, near Fernald, Ohio, is a prime contractor site
operated by Westinghouse Materials Company of Ohio for the DOE.
The primary mission at the FMPC is to produce purified 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. Emissions
from these processes are addressed in Chapter 2.
10-1
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The primary source of radon emissions at the FMPC is
pitchblende residues stored in two concrete storage tanks,
referred to as silos. These residues resulted from the recovery
of uranium from pitchblende ores during World War II. The
storage silos are located on the western portion of the site,
south of the chemical waste pits and approximately 325 m from the
western site boundary (We87, We88).
The residues are estimated to have a radium concentration of
0.2 ppm, equivalent to about 200,000 pCi/g of radium-226. The
estimated 11,200 kg of residues contain about 1,760 Ci of radium.
The two concrete storage silos were constructed in 1951 and 1952.
In 1964, the silos were repaired, and an earth embankment was
erected around the silos to provide structural integrity and
weather protection, as well as to reduce the radon emissions and
the direct radiation from the silos. In 1979, the vents on the
silos were sealed to further reduce radon emissions. In 1983,
the earth embankment was enlarged.
On July 18, 1986, the DOE and EPA jointly signed a Federal
Facility Compliance Agreement (FFCA) (We88). The state of Ohio
has also been actively involved in the project effort. In
response to the FFCA, the FMPC took action to stabilize the two
K-65 waste storage silos by adding temporary 9.14 m diameter
domes. A foam covering was added on top of the domes to seal the
surface from the weather, insulate the silos from thermal
fluctuations, provide more structural integrity, and further
prevent radon releases (Bo87, DOE86b, DOE87b).
In 1987, the FMPC prepared a report entitled "Feasibility
Investigation for Control of Radon Emission from the K-65 silos"
(Gr87), to evaluate alternatives for the control of radon
emissions in response to CERCLA issues. The report determined
that the FMPC is within the DOE and EPA guidelines and
regulations for the emission of radon, but that additional radon
control would be needed if the silos were to crack. The report
recommended that the void space in the silos be filled with foam
and that weatherproofing be completed after the silos are filled.
The current schedule for Remedial Investigation/Feasibility Study
(RI/FS) activities calls for the Record of Decision (ROD) to be
issued in September of 1990.
The void space has not yet been filled with foam, and the
risk estimates presented here do not account for the foam. When
the foam is inserted in the dome, the radon emissions will be
further reduced, and the risk estimates will be lower (Gr88).
10.1.2 The Niagara Falls Storage Site
The NFSS in Lewiston, New York, is a DOE surplus facility,
operated by Bechtel National, Inc. The 77-ha site, part of the
former Lake Ontario Ordnance Works, is used solely to store
uranium and pitchblende residues.
10-2
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The residues, which were previously contained in buildings
on the site, were consolidated in the Interim Waste Containment
Facility (IWCF) at the end of 1986 (Jo87). Details of the
consolidation are given in the Annual Environmental Reports
(Be87b). The IWCF structure comprises the short-term closure
system for the wastes until the long-term management plan is
completed.
The IWCF occupies 4 ha of the site, measuring 274 m by
137 m. The structure's outer perimeter is formed by a dike and
cutoff wall, each constructed of compacted clay and incorporated
into the finished structure. An engineered, compacted clay cover
placed immediately over the wastes extends beyond the
perimeter dike, completely enclosing the containment structure.
This cover is the principal barrier against moisture intrusion
and radon emanation. The 0.9-m clay layer is covered with 0.3 m
of general soil and 0.15 m of topsoil.
The DOE Record of Decision on long-term disposition of the
NFSS was issued in August 1986. The plan selected is long-term,
in-place management consistent with the guidance provided in the
EPA's regulations governing uranium mill tailings. The plan is
described in the Final Environmental Impact Statement, published
in April 1986 (DOE86a).
The radon level measurements at the site boundary have
decreased over the past few years as a result of remedial actions.
The locations monitored in 1986 read between 0.17 and 0.36 pCi/1
(average 0.26 pCi/1), including background. The background
location averaged 0.31 pCi/1. Mound Labs performed supplemental
radon monitoring in 1986 at the site boundary. These values
ranged between 0.21 and 0.31 pCi/1 (average 0.27 pCi/1),
including background. The background location had a reading of
0.22 pCi/1. These values show good agreement with the values
obtained by the site. Radon monitoring was also performed beyond
the site boundary. The values ranged between 0.20 and 0.35 pCi/1
(average 0.25 pCi/1), including background. Background was
0.22 pCi/1. The current radon levels should be lower due to the
capping of the IWCF, completed in late 1986 (Be87b)
10.1.3 The Weldon Spring Site
The WSS, near Weldon Spring, Missouri, is a DOE surplus
facility The site consists of two physically separate areas,
the 89-ha Weldon Spring Chemical Plant (WSCP) and the Weldon
Spring Raffinate Pits (WSRP) area, and the 3.6-ha Weldon Spring
Quarry (WSQ) area.
The DOE was directed by the Office of Management and Budget
to assume custody and accountability for the WSCP from the
Department of the Army in November 1984. The control and
decontamination of the WSCP, WSRP, and WSQ was designated as a
major project by DOE Order 4240.IE dated May 14, 1985.
Mk-Ferguson Company assumed control as Project Management
10-3
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Contractor for the WSS Remedial Action Project on October 1,
1986. Remediation at this site is being pursued under the
requirements of CERCLA. The DOE has entered into an agreement
with the EPA. A Remedial Investigation/Feasibility Study is in
progress, and the Record of Decision is scheduled for 1991.
Like the NFSS, the Weldon Spring Site is used for the
storage of uranium and thorium wastes. The raffinate pits area
is a remnant of the Weldon Spring Chemical Plant. During the
period that the chemical plant was operated for the Atomic Energy
Commission, the four raffinate pits, occupying 21 ha of the WSCP
and WSRP area, received residues and waste streams from the
uranium and thorium processes conducted at the facility. Pits 1
and 2 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 10-1 (MK86). Surface water (varying in depth with the
seasons) always covers the residues in pits 3 and 4. Pits 1 and
2 are usually covered by water as well, but evaporation during
the summer months can leave these residues exposed.
The quarry site, located about 6 km 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 is a closed basin with surface water
within the rims flowing to the quarry floor and to the sump pond.
The level of water in the pond varies with precipitation and
temperature. There is a storage shed and sampling platform in
the sump area. The site is surrounded by a locked 2.1 m fence
topped with wire.
The quarry was first used to dispose of radioactive wastes
in 1959, when the AEC deposited thorium residues in drums.
During 1963 and 1964, approximately 32,000 cubic meters 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 WSCP. 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
Army's decontamination of the chemical plant generated approxi-
mately 4,600 cubic meters of contaminated equipment and rubble.
Table 10-2 summarizes the radioactive wastes stored in the quarry
(MK86).
The environmental monitoring program for radon consists of
6 locations in the WSRP area, 15 locations in the WSCP area,
10-4
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6 locations at the WSQ, and 4 offsite locations for background
readings. The "1986 Annual Environmental Monitoring Report11
(MK86) indicates that the site boundary radon monitors at WSCP
(which includes the raffinate pits area) read between 0.18 and
0.49 pCi/1 (average 0.32) including background. The background
location read 0.47. The offsite monitors north of the pits and
closer than the other background monitor read between 0.22 and
0.36 pCi/1 (average 0.29). The onsite monitors at the raffinate
pits read between 0.31 and 0.64 pCi/1 (average 0.46). The onsite
monitors at the quarry read between 0.24 and 1.86 pCi/1 (average
0.87) (MK86).
Table 10-1. Characteristics of the four raffinate pits and
activity levels of major radionuclides in the
currently stored materials.
Characteristic
Pit 1
Pit 2
Pit 3
Pit 4
Year Constructed
Surface Area, ha
Pit Volume, m3
Waste Volume, m3
Radionuclide
U-238
U-234
Th-232
Th-230
Ra-228
Ra-226
1958
0.5
14,060
13,224
710
810
100
24,000
850
430
1958
0.5
14,060
13,224
Activity
470
560
120
24,000
200
440
1959
3.4
126,692
98,490
(pci/g)
520
570
120
14,000
100
460
1964
6.
337,
42,
620
610
120
1,600
60
11
1
744
256
10-5
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Table 10-2,
Type of
Waste
Estimated volumes of radioactive wastes stored in
the Weldon Spring Quarry.
Date
Deposited
Volume
Comments
3.8 Percent
Thorium
Residues
1959
140.1
Destrehan St.
Plant
Demolition
Rubble
1963-1964
38,000
3 Percent
Thorium
Residues
Weldon Spring
Feed Materials
Flant Rubble
1966
422
1968-1969
4,222
Drummed residues; volume
estimated; most of the
residues under water;
principal source of
radioactivity is
Th-232 decay series.
Contaminated equipment,
building rubble; estimate
of uranium and thorium
content not available;
principal source of
radioactivity is
U-238 decay series.
Drummed residues; volume
estimated; stored above
water level; principal
source of radioactivity
is Th-232 decay series.
Contaminated equipment,
building rubble; uranium
and thorium content and
radioactivity not avail-
able; principal sources
of radioactivity are
U-238 and Th-232
decay series.
Total
42,784
10-6
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10.1.4 The Middlesex Sampling Plant
The MSP, Middlesex, New Jersey, was used by the Manhattan
Engineering District and the Atomic Energy Commission between
1943 and 1967 for sampling, weighing, assaying, and storing
uranium and thorium ores. The site covers 3.9 ha. After
termination of operations in 1967, the site was decontaminated
and released to the U.S. Marine Corps 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. Both the
Middlesex Municipal Landfill (MML), located 0.8 km north-
northwest of the MSP, and the MSP were designated for remedial
action under FUSRAP.
The cleanup of the MSP, which was completed in 1982,
consisted of recovering contaminated soils from offsite
properties and removing contaminated soil areas from the site.
All materials were consolidated in a storage pile on the southern
portion of the site (Fo79).
In 1984, contaminated soils were transported from the MML to
MSP for interim storage. The storage pad at MSP was enlarged to
accommodate these soils, which were placed on a second pile.
Together, the two storage piles occupy about 2.2 ha, or over half
of the site. Concrete curbing surrounds the pad to prevent
migration of the materials. The top of the storage pile is also
covered with a hypalon material to prevent movement of the
materials (Be85). In 1986, the remedial actions were completed
for the landfill. The volumes of contaminated soils on the MSP
storage pads are given in Table 10-3. The concentration of
radium-226 in the piles is estimated to be 40 pCi/g (Fr88).
Table 10-3. Volumes of contaminated soil on the MSP storage pads.
Date and Source Volume
1980 (Phase I) MSP Cleanup 7,160
1981 (Phase II) MSP Cleanup 19,564
1984 MML Cleanup
(Second Storage Pad) 11,400
1986 MML Cleanup
(Extended Second Storage Pad) 12,234
•Total on Storage Pads 50,358
10-7
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Environmental monitoring at the MML site was discontinued
after 1987. The certification docket releasing the site for
unrestricted use was published in May 1989 (Be89). Environmen-
tal monitoring, maintenance, and surveillance will continue at
the MSP until all remedial activities are completed. The sched-
ule for remediation of the MSP site calls for site surveillance
through 1991, planning, NEPA/CERCLA and design efforts through
1993, and completion of remedial action (excluding certification
docket) by the end of 1996.
The environmental monitoring program for radon consists of
20 locations at the MSP. The detectors are located at site and
on the site boundary. One detector is located about 16 km from
the MSP to measure background levels. The "1986 Annual
Environmental Monitoring Report" (Be87a) indicates the site
boundary radon monitors read between 0.3 and 1.2 pCi/1, including
background, at the MSP. The offsite rate was 2.0 pCi/1. (The
offsite location is apparently at a higher radiation level than
the site itself.) All levels in 1986, including background, were
three times those in 1985. This was observed at other sites in
New Jersey and is believed to be due to drier climatic
conditions. In a nine-month radon survey conducted by Mound Labs
at MSP in 1986 the site boundary detectors ranged between 0.2 and
0.3 pCi/1 (Be87a). The off-site background detector averaged
0.2 pCi/1.
10.1.5 The Monticello Uranium Mill Tailings Pile
The MUMT, located at Monticello, Utah, has been inactive
since 1960. About 817,000 MT of uranium mill tailings were
impounded in four separate areas covering a total of about
18.6 ha. 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 shut down. The government owns the
tailings site. Uranium ore was processed by both acid and
carbonate leaching, and thus the tailings exhibit properties of
both of these processes.
The tailings were stabilized in 1961 by grading and leveling
and the dikes were made of tailings. The tailings were then
covered with about 0.3 m of pit run gravel and dirt, followed by
0.3 m of top soil that was seeded with local vegetation.
Currently, there is about 0.15 m of soil on some areas of the
pile, and the grass cover is not good. Additional demolition
and decontamination activities were conducted in 1974 and 1975 to
reduce radiation levels at the site and improve its appearance.
The mill site was accepted into the Surplus Facilities
Management Program (SFMP) in 1980. The Monticello Remedial
Action Project (MRAP) is specific to the mill site and
contaminated peripheral properties. Areas contaminated outside
those covered by the MRAP are included under the Monticello
Vicinity Properties (MVP) Project.
10-8
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The DOE has completed the Hazard Ranking System evaluation
(score =52.0). A draft RI/FS was completed in January 1988 for
the mill site. Although the mill site is not on the National
Priorities List, guidance from the DOE and EPA mandates that
contractors are to comply with the requirements of CERCLA and
SARA. The DOE, EPA, and the State of Utah have entered into
negotiations for an Interagency Agreement under CERCLA Section
120. A Draft Work Plan is undergoing comment. The MRAP Draft
Work Plan indicates planned completion of the RI/FS by early 1990
with the ROD to follow shortly thereafter. Remedial activities
are expected to begin in 1990 with completion and certification
scheduled for around 1995.
The "1986 Environmental Monitoring Report" (Se87) refers to
the "Draft Environmental Assessment of Remedial Action - 1985"
(Ben85, UN88) as containing onsite and offsite measurements that
represent current conditions. Only minor additions of ore have
since been made to the pile. The report (Ben85) presents several
onsite radon flux measurements and concludes that the EPA
standard for flux of 20 pCi/m2/s is exceeded at each of the
four tailings piles.
10.2 BASIS OF THE RISK ASSESSMENT
10.2.1 Emissions
10.2.1.1 Radon Emission Estimates for the FMPC
There is no current information on the flux of radon-222
from the silos at FMPC. Measurements made by Monsanto-Mound in
1984 and 1985 are no longer valid because of the significant
changes made to the silos since then (Gr87, We87) The radon
releases from the silos were calculated before the 0.1-m foam
covering was placed on top of the domes; thus, these calculations
are also no longer valid. The latter calculation predicted that
650 Ci of radon-222 would be released each year. Radon
concentrations have been measured outside the silos, but the
information needed to develop the actual radon emissions from the
silos is insufficient (We87, We88, DOE87b).
The current radon source term is estimated, based on the
radium content of the residues, the reported areas of the silos,
and the calculated radon flux through the concrete domes. This
latter estimate was based on relationships presented in
Atmospheric Environment (Na85). The radon-222 emissions, after
foaming the exterior of the dome, are estimated to be about
2.5 Ci/y. The current estimated emission rate is 85 pCi/m2/s.
Assuming that remedial activities reduce the radon emission rate
to 20 pCi/m2/s, the emissions would be reduced to 0.6 Ci/y.
10.2.1.2 Radon Emission Estimates for the NFSS
Radon emission estimates are based on the estimated releases
presented in the "Closure/Post-Closure Plan for the Interim Waste
10-9
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Containment Facility at the Niagara Falls Storage Site" (Be86),
and the "Final Environmental Impact Statement" (DOE86a). The
estimated releases from the current storage facility are
0.25 Ci/y (Be86). This corresponds to a radon emission rate of
0.06 pCi/m2/s, which is well below the 20 pCi/m2/s design
standard in 40 CFR Part 192.
The releases of radon from the IWCF are not available in
terms of flux from the pile. Also, the site boundary data as
summarized in Section 10.1.2 are not usable for estimating
releases because they are nearly indistinguishable from the
background data.
10.2.1.3 Radon Emission Estimates for the WSCP and WSQ
Radon emission estimates are based on DOE's estimated
releases (DOE87a). The estimated releases for the current
situation (described as Alternative 4, "No Action," in DOE87a)
are 29 Ci/y of radon-222 for the WSCP, and 14 Ci/y of radon-222
for the WSQ. The current radon emission rates for both sites,
estimated at 2.7 and 3.7 pCi/m2/s for the WSCP and WSQ
respectively, are below the 20 pCi/m2/s design standard in
40 CFR Part 192.
Measured releases of radon from the WSCP and WSQ are not
available in terms of flux from the pits and quarry.
10.2.1.4 Radon Emission Estimates for the MSP
DOE sampled the wastes in the piles in April 1985 and July
1986 (Wa88). The results of these samplings, as noted above,
indicate an average of 40 pCi/g of radium-226 (Fr88). Assuming
that 1 pCi/g radium-226 results in 1 pCi/m2/s radon-222 the
estimated flux rate is 40 pCi/m2/s. Given the dimensions of the
waste piles, the radon source term is estimated at 25 Ci/y. This
estimate gives no credit for any radon attenuation by the hypalon
cover over the wastes (Be85). Reduction of the emission rate to
20 pCi/m2/s would result in a release rate of 13 Ci/y.
The releases of radon at MSP are not available in terms of
flux from the interim storage piles.
10.2.1.5 Radon Emission Estimates for the MUMT
Radiation measurements at the site have been made primarily
to determine external gamma radiation levels. These levels were
reduced by stabilization to a range of 2 to 3 above background
levels (author's observation). Radon emission measurements range
from 133 to 765 pCi/m2/s for the four tailings piles, according
to the "Draft Environmental Assessment of Remedial Action - 1985"
(Ben85) (see Table 10-4). Part of the pile has migrated up to
500 m offsite along Montezuma Creek. The average flux rate of
this material is 40 pCi/m2/s, or 37 Ci/y. DOE estimates the
total radon-222 release to be 1,595 Ci/y (Ben85). This emission
10-10
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rate was assumed to occur from an area of 2.17E+5 m2 (Fr88).
When averaged over all the piles, the current radon emission rate
is 228 pCi/m2/s.
Table 10-4. Radon source strength, areas, and radon flux rates
at the MUMT.
Tailings Pile
Acid Pile
Carbonate Pile
Vanadium Pile
East Pile
Montezuma Creek
West, East, & Central
Radon
Release
(Ci/y)
500
570
88
400
37
Area
(m2)
52,070
23,657
16,216
95,746
29,000(a)
Weighted-Average
Area Radon Flux
(pCi/m2 sec)
312
765
173
133
40
Total 1,595
(a) Estimated based on total area of 2.17E+5 m2 (Fr88).
10.2.2 Other Assumptions Used in the Assessment
Meteorological data for each of these sites were obtained
from nearby weather stations. Nearby population figures were
obtained from DOE reports (Ab84, Gr87, We87, Fo79, DOE87a), and
the regional populations were generated from U.S. census tract
data from 1980 using the computer code SECPOP. All of the sites
were treated as ground-level area sources.
10.3 RESULTS OF THE RISK ASSESSMENT
Exposures and risks to nearby individuals and risks to the
regional population were estimated for both pre- and post-
remediation radon emission rates. A post-remediation emission
rate of the lesser of either 20 pCi/m2/s radon or the current
emission rate was assumed.
10.3.1 Exposures and Risks to Nearby Individuals
The pre-remediation exposures received by individuals living
near these sites and their lifetime fatal cancer risks are
summarized in Table 10-5. The highest risks are associated with
the MUMT, where nearby individuals are estimated to have a
0.1 percent lifetime fatal cancer risk. For the MSP and the
WSCP, nearby individuals are estimated to have a lifetime fatal
cancer risk of l and 2 in 10,000, respectively. At the FMPC and
•WSQ, the nearby individuals have a risk of less than 1 in 10,000,
while at the NFSS the maximum estimated risk is less than 1 in
1 million.
10-11
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Table 10-5. Estimated exposures and risks to individuals
living near DOE radon sites assuming current
radon emission rates.
Maximum Maximum
Initial Rn-222 Maximum Lifetime Fatal
Flux Rate Concentration Exposure Cancer Risk Distance
(pCi/m2/s) (pCi/1) (WL) to Individual (m)
FMPC
NFSS
WSCP
WSQ
MSP
MUMT
85
6E-2
2.7
3.7
40
40
5E-4
6E-5
5E-2
2E-2
4E-2
3E-1
1.5E-6
1.8E-7
1.3E-4
5.6E-5
l.OE-4
9.7E-4
2E-6
3E-7
2E-4
8E-5
1E-4
1E-3
800
500
300
300
400
900
The post-remediation exposures received by individuals living
near these sites and their lifetime fatal cancer risks are
summarized in Table 10-6. The radon emission rate for the NFSS,
WSCP, and WSQ are currently below 20 pCi/m2/s; therefore, the
risks to individuals near these facilities are not shown to
change. At the MUMT and the MSP, the nearby individuals have a
risk of 1 and 0.8 in 10,000, respectively, while at the FMPC the
maximum estimated risk is less than 1 in 1 million.
10.3.2 Risks to the Regional (0-80 km) Populations
The estimated fatal cancers per year in the populations
around DOE radon sites, as a result of current emissions and
post-remediation emissions, are summarized in Table 10-7, along
with the numbers of persons in the population around each site.
The emissions from the MSP result in a greater number of fatal
cancers per year, even though the releases from the MUMT are a
factor of 64 greater than those at the MSP. This is due to the
great disparity in the numbers of persons within 80 km of each
site. Based on current emissions, the estimated total deaths per
year are 7E-2. This is equivalent to one death every 14 years.
The estimated post-remediation total deaths per year are 4E-2.
This is equivalent to one death every 25 years.
10-12
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Table 10-6. Estimated exposures and risks to individuals
living near DOE radon sites assuming
post-remediation radon emission rates.
Maximum Maximum
Rn-222 Maximum Lifetime Fatal
Concentration Exposure Cancer Risk
(pCi/1) (WL) to Individual
FMPc(a)
NFSS(b)
WSCP(C)
WSQ(d)
MSP(a)
MUMT(a)
1E-4
6E-5
5E-2
2E-2
2E-2
3E-2
3.6E-7
1.8E-7
1.3E-4
5.6E-5
5.4E-5
8.5E-5
5E-7
3E-7
2E-4
7E-5
8E-5
1E-4
Distance
(m)
800
500
300
300
400
900
(a) Based on 20 pCi/m2/s.
(b) Based on 6E-2 pCi/m2/s.
(c) Based on 2.7 pCi/m2/s.
(d) Based on 3.7 pCi/m2/s.
Table 10-7,
Estimated fatal cancers/year to the regional
(0-80 km) populations around DOE radon sites
for current radon emission rates.
Fatal Cancers Per Year
Facility
Feed Material
Production Center
Niagara Falls
Storage Site
Weldon Springs
Pits & Quarry
Middlesex
Sampling Plant
Monticello Uranium
Mill Tailings
Totals (a)
(a) Totals may not add
Population Current Post-Remediation
3,200,000
3,800,000
2,300,000
16,000,000
19,000
25,300,000
due to independent
6E-4
4E-5
1E-2
5E-2
8E-3
7E-2
rounding.
1E-4
4E-5
1E-2
3E-2
7E-4
4E-2
10-13
-------
10.3.3 Distribution of the Fatal Cancer Risk
Tables 10-8 through 10-13 show the distribution of fatal
cancer risk in the regional populations around each site for
current radon emission rates. Tables 10-14 through 10-16 show
the distribution of fatal cancer risk in the regional populations
around the FMPC, MSP, and MUMT sites for post-remediation radon
emission rates of 20 pCi/m2/s. Post-remediation emission rates
are not shown for the NFSS, WSCP, and WSQ sites since their
current radon emissions are already less than 20 pCi/m2/s.
Tables 10-17 and 10-18 summarize this information for the
entire DOE radon site source category for current and post-
remediation emissions, respectively. It should be noted that all
of the individuals estimated to have a lifetime fatal cancer risk
greater than 0.1 percent reside in the area around the MUMT.
10-14
-------
Table 10-8. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the FMPC
for current radon emission rates.
Number
Risk Interval of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
0
0
38
3,300,000
0
0
0
0
0
9E-7
6E-4
Totals(a) 3,300,000 6E-4
(a) Totals may not add due to independent rounding.
Table 10-9. Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the NFSS
for current radon emission rates.
Number
Risk Interval of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
0
0
0
3,800,000
0
0
0
0
0
0
4E-5
Totals(a) 3,800,000 4E-5
(a) Totals may not add due to independent rounding.
10-15
-------
Table 10-10.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the WSCP
for current radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
70
400
27,000
2,300,000
0
0
0
1E-4
1E-4
6E-4
8E-3
TotalsC3) 2,300,000 9E-3
(a) Totals may not add due to independent rounding.
Table 10-11.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the WSQ
for current radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
0
200
4,000
2,300,000
0
0
0
0
8E-5
1E-4
4E-3
Totals(a) 2,300,000
(a) Totals may not add due to independent rounding.
4E-3
10-16
-------
Table 10-12.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the MSP
for current radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
200
4,000
310,000
16,000,000
0
0
0
4E-4
2E-3
7E-3
4E-2
Totals(a) 16,000,000 5E-2
(a) Totals may not add due to independent rounding.
Table 10-13.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the MUMT
for current radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
30
1,700
3,300
14,000
180
0
0
6E-4
5E-3
1E-3
9E-4
2E-6
Totals(a) 19,000
(a) Totals may not add due to independent rounding.
8E-3
10-17
-------
Table 10-14.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the FMPC
for post-remediation radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
Totals (a)
0
0
0
0
0
0
3,300,000
3,300,000
0
0
0
0
0
0
1E-4
1E-4
(a) Totals may not add due to independent rounding.
Table 10-15.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the MSP
for post-remediation radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
0
2,000
60,000
16,000,000
0
0
0
0
9E-4
2E-3
2E-2
Totals(a) 16,000,000
(a) Totals may not add due to independent rounding.
3E-2
10-18
-------
Table 10-16.
Estimated distribution of the fatal cancer risk to
the regional (0-80 km) population around the MUMT
for post-remediation radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
30
1,000
14,000
0
0
0
0
5E-5
4E-4
2E-4
8E-5
Totals(a) 19,000 7E-4
(a) Totals may not add due to independent rounding.
Table 10-17.
Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around all DOE radon sites for current
radon emission rates.
Risk Interval
Number
of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< IE- 6
0
0
32
2,000
8,000
360,000
28,000,000
0
0
6E-4
6E-3
3E-3
9E-3
5E-2
Totals(a)
28,000,000
7E-2
(a) Totals may not add due to independent rounding.
10-19
-------
Table 10-18. Estimated distribution of the fatal cancer
risk to the regional (0-80 km) population
around all DOE radon sites for post-remediation
radon emission rates.
Number
Risk Interval of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
100
4,000
92,000
28,000,000
0
0
0
2E-4
1E-3
3E-3
4E-2
Totals(a) 28,000,000 4E-2
(a) Totals may not add due to independent rounding.
10.4 SUPPLEMENTARY CONTROL OPTIONS AND COST
For each of the five sites discussed in this chapter, three
similar supplementary control options required to reduce the
radon emissions to levels of 20, 6, and 2 pCi/m2/s, and their
associated costs, were evaluated. At each site, the current
storage configuration was assumed (e.g., the four mill tailings
piles at Monticello were not moved into one larger pile). The
depth of earth required to reduce radon emissions to the three
levels mentioned above, and the associated costs, were then
calculated using the equations and unit cost for earth covers
presented in Appendix B. It should be noted that the wastes at
the NFSS and the FMPC might require disposal as high-level wastes
at a facility such as the WIPP. However, for this evaluation, it
is assumed that these wastes'remain at-the current sites.
10.4.1 The Feed Materials Production Center
The radon emission rate from the two silos, using the
estimated 2.5 Ci/y source term, is calculated to be 85 pCi/m2/s.
The depths of earth required to reduce the emissions to 20, 6,
and 2 pCi/m2/s are 2.1, 2.3, and 3.3 m, respectively. Based on
the current configuration, it was assumed that only the exposed
domes would have to be covered, and a 3:1 slope was used. The
estimated costs of the coverings are $56,000, $79,000, and
$83,000, to meet the levels of 20, 6, and 2 pCi/m2/s.
10.4.2 The Niagara Falls Storage Site
The current radon emission rate from the IWCF is 0.25 Ci/y,
equivalent to a radon flux of 0.6 pCi/m2/s. Since the current
10-20
-------
emission rate is below all of the proposed options, there are no
costs associated with meeting any of the alternatives.
10.4.3 The Weldon Spring Site
The radon emission flux from the present raffinate pits at
the WSCP is 2.7 pCi/m2/s, while the flux from the WSQ is
3.7 pCi/m2/s. Both the pits and quarry are covered with
water, at various levels depending upon the season and variations
in the rainfall rate.. For the purpose of determining the costs
of achieving the alternative levels, it was assumed that both the
pits and the quarry would be dry. The estimated radon flux from
the dry pits was calculated based on the information presented in
Table 10-1. For pits 1, 2, and 3, the estimated flux is
460 pCi/m2/s, while for pit 4, it is 11 pCi/m2/s.
The depths of earth required to reduce the emission rates to
20, 6, and 2 pCi/m2/s for pits 1, 2, and 3 are 1.6, 2.3, and
2.8 m, respectively. For pit 4, no cover is needed to achieve
20 pCi/m2/s, while 0.3 and 0.9 m would be required to meet the
two lower options. The estimated costs for all four .pits is
$1.73 million to achieve 20 pCi/m2/s, $2.96 million to achieve
6 pCi/m2/s, and $4.26 million to achieve 2 pCi/m2/s.
At this time, insufficient information is available to
develop the costs of achieving the alternative levels for the
WSQ.
•
10.4.4 The Middlesex Sampling Plant
The radon emission rate from the interim storage facility is
estimated to be 40 pCi/m2/s. The depths of earth required to
reduce this to 20, 6, and 2 pCi/m2/s are 0.8, 1.4, and
2.1 meters, respectively. The estimated costs of the earthen
covers are $419,000, $720,000, and $997,000, respectively.
10.4.5 The Monticello Uranium Mill Tailings Piles
The current radon emission rate at the MUMT, averaged over
all of the piles, is 228 pCi/m2/s. The depths of earth
required to reduce the radon flux to 20, 6, and 2 pCi/m2/s are
2.4, 3.4, and 4.4 m, respectively. The costs to achieve these
levels are estimated to be $26.8 million, $39.2 million, and
$50.2 million, respectively. Included in these estimates is the
cost of rip-rap, needed to provide long-term erosion control and
to prevent misuse of the tailings.
The costs to reduce the radon flux to 20, 6, and 2 pCi/m2/s
at all the DOE radon sites are summarized in Table 10-19.
10-21
-------
10.4.6 Effectiveness of the Control Options
Covering the DOE radon sources to reduce the current
emissions to 20, 6, and 2 pCi/m2/s reduces the maximum
individual risk from 1E-3 to 2E-4, 2E-4, and 1E-4, respectively.
It will also reduce the deaths per year estimates to the regional
populations within 80 km from 7E-2 to 4E-2, 2E-2, and 1E-2,
respectively.
Table 10-19,
Site
Summary of capital costs to reduce radon emissions
from DOE radon sites.
Capital Cost ($ 1988 million)
Radon Flux
20 pCi/m2/s
Radon Flux
6 pCi/m2/s
Radon Flux
2 pCi/m2/s
FMPC
NFSS
WSQ
MSP
MUMT
0.056
0
1.7
0.42
27
0.079
0
3.0
0.72
39
0
0
4
1
50
.083
.3
.0
10-22
-------
10.5 REFERENCES
Ab84 Abramiumk, I.N., et al., "Monticello Remedial Action
Project Site Analysis Report," GJ-10, Bendix Field
Engineering Corporation, Grand Junction, CO, December
1984.
Be85 Bechtel National, Inc., "Technical Specification for
Furnishing and Installing Stockpile Cover, Middlesex
Sampling Plant, Middlesex, New Jersey," Specification
17-14-C-05, September 5, 1985.
Be86 Bechtel National, Inc., "Closure/Post-Closure Plan for
the Interim Waste Containment Facility at the Niagara
Falls Storage Site," DOE/OR/20722-85, Oak Ridge, TN, May
1986.
Be87a Bechtel National, Inc., "Middlesex Sampling Plant and
Middlesex Municipal Landfill, Annual Site Environmental
Report, Calendar Year 1986," DOE/OR/20722-149, Oak
Ridge, TN, May 1987.
Be87b Bechtel National, Inc., "Niagara Falls Storage Site,
Annual Site Environmental Report, Calendar Year 1986,"
DOE/OR/20722-150, Oak Ridge, TN, June 1987.
Be89 Bechtel National, Inc., "Certification Docket for the
Remedial Action Performed at the Middlesex Sampling Plant
in Middlesex, New Jersey in 1984 and 1986," May 1989.
Ben85 Bendix Field Engineering Corp., "Draft Environmental
Assessment of Remedial Action at the Monticello Uranium
Mill Tailings Site, Monticello, Utah," DOE-EA, Grand
Junction, CO, 1985.
Bo87 Boback, M.W., et al., "History of FMPC Radionuclide
Discharges," FMPC-2082, Special, UC-11, Feed Materials
Production Center, Westinghouse Materials Company of
Ohio, Cincinnati, OH, May 1987.
DOE86a U.S. Department of Energy, "Final Environmental Impact
Statement, Long-Term Management of the Existing
Radioactive Wastes and Residues at the Niagara Falls
Storage Site," DOE/EIS-0109F, April 1986.
DOE86b U.S. Department of Energy, "Investigation of April 25,
1986, Radon Gas Release from Feed Materials Production
Center, K-65 Silos, by DOE Incident Investigation Board,"
DOE/OR-877, June 27, 1986.
DOE87a U.S. Department of Energy, "Draft Environmental Impact
Statement, Remedial Action at the Weldon Spring Site,"
DOE/EIS-0117D, February 1987.
10-23
-------
DOE87b
EPA84
F079
Fr88
Gr87
Gr88
J087
MK86
Na85
Re8 8
Se87
U.S. Department of Energy, Office of Environment, Safety,
and Health, and Office of Environmental Audit,
"Environmental Survey Preliminary Report, Feed Materials
Production Center, Fernald, Ohio," March 1987.
U.S. Environmental Protection Agency, "Background
Information Document for Final Rules, Volume II, Appendix
C, Radon Emissions from Department of Energy- and Nuclear
Regulatory Commission-Licensed Facilities,"
EPA 520/1-84-022-2, Washington, DC, October 1984.
Ford, Bacon & Davis Utah, Inc., "Engineering Evaluation
of the Former Middlesex Sampling Plant and Associated
Properties, Middlesex, New Jersey," FBDU 230-001 and
FBDU 230-005, Salt Lake City, UT, July 1979.
Frangos, T.G., U.S. Department of Energy, attachment 4 of
written communication to T. McLaughlin, USEPA, presenting
recommendations concerning the 40 CFR Part 61 Subpart H
rulemaking, December 19, 1988.
Grumski, J.T., "Feasibility Investigation for Control of
Radon Emissions from the K-65 Silos, Feed Materials
Production Center, Westinghouse Materials Company of
Ohio," July 30, 1987.
Grumski, J.T., and Shanks, P.A., "Completion Report, K-65
Interim Stabilization Project, Exterior Foam
Application/Radon Treatment System Operation, Revision
1," April 1988.
Jones, M.G., et al., "Performance Monitoring Report for
the Niagara Falls Storage Site Waste Containment
Structure," DOE/OR/20722-159, prepared for the Department
of Energy, Bechtel National, Inc., Oak Ridge, TN, July
1987.
MK-Ferguson Company aftd Jacobs.Engineering Group, Inc.,
"Weldon Spring Site, Annual Environmental Monitoring
Report, Calendar Year 1986," St. Charles, MO.
Nazaroff, W.W, et al., "Radon Transport Into a Detached
One-Story House With a Basement," Atmospheric Environment.
Vol. 19, #1, pp. 31-46, Great Britain, 1985.
Reafsnyder, J.A., Department of Energy, Oak Ridge
Operations, written communication to W. Britz, SC&A,
Inc., June 21, 1988.
Sewell, M., and Spencer, L., "Environmental Monitoring
Report on Department of Energy Facilities at Grand
Junction, Colorado, and Monticello, Utah, Calendar Year
1986," UNC/GJ-HMWP-2, UNC, Grand Junction, CO, March
1987.
10-24
-------
UN88 UNC Technical Services, Inc., Grand Junction, CO, written
communication to W. Britz, SC&A, Inc., March 9, 1988.
Wa88 Wallo, A. Ill, Department of Energy, Office of Nuclear
Energy, written communication to W. Britz, SC&A, Inc.,
June 16, 1988.
We87 Westinghouse Materials Company of Ohio, "Feed Materials
Production Center, Environmental Monitoring Annual Report
for 1986," FMPC-2076, Special, UC-41, Cincinnati, OH,
April 30, 1987.
We88 Westinghouse Materials Company of Ohio, "Feed Materials
Production Center, Environmental Monitoring Annual Report
for 1987," FMPC-2135, Special, UC-41, Cincinnati, OH,
April 30, 1988.
10-25
-------
-------
11. UNDERGROUND URANIUM MINES
11.1 GENERAL DESCRIPTION
In conventional uranium mining operations, ore is removed
from the ground in concentrations of 0.1 to 0.2 percent U^Os or
280 to 560 /iCi of uranium-238 per metric ton of ore. Since the
uranium-238 in the ore is normally present in near secular
equilibrium with its decay products, these ores also contain
about 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 and produce the product UsOg. Radioactive
emissions to air from uranium mines and mills consist of radio-
nuclide-bearing dust and radon-222 gas.
Conventional uranium mining operations include both
underground and open pit mines. In 1987, conventional mining
techniques accounted for about 63 percent of total U.S. uranium
production (Pi88a). (The health impact of open pit mines is
assessed in Chapter 12.)
In 1982, 139 underground mines were operating in the United
States (DOES3). However, during the past six years, uranium
production and the number of uranium mines in the United States
have declined sharply. Currently, only 13 underground uranium
mines are producing ore (Section 23, Mt. Taylor, eight UMETCO
Minerals Corporation mines, and three breccia-pipe mines). In
addition, two underground mines (Sheep Mountain No. 1 and
Schwartzwalder) are on standby. The production of ^Og by
conventional mining methods fell from 20.6 million pounds in 1982
to only 7.8 million pounds in 1987 (Pi88a). The principal causes
of this reduction were a decline in the price of ^Og ($40 per
pound in 1980 to the current $15 per pound) and the increasing
competition from foreign suppliers (EPA83a, Pi88a).
A list of the currently operating mines is presented in
Table 11-1. Although on standby status, the Schwartzwalder mine
is included because it continues to operate its ventilation
system during exploration activities and releases radon-222 to
the air. If the outcome of the current explorations is
favorable, it will resume production. Also, Sheep Mountain No. 1
may be expected to reopen if there is a sufficient increase in
the market price of ^Og (Pi89). The expected life of these
mines and their ore production rates are included in the table.
Only the Mt. Taylor mine in New Mexico is expected to operate
over an extended period. The three breccia-pipe mines are not
expected to operate for more than about six years (Pi88a). Thus,
underground uranium mines are present in five western states,
but it is likely that uranium mining will be conducted in fewer
states during the next decade.
11-1
-------
Table 11-1. Currently operating underground uranium mines in the United States.
I
to
State
Arizona
Arizona
Arizona
Colorado
Colorado
Colorado
Colorado
Colorado
Mine Name
Kanab North
Pigeon
Pinenut
Calllham
Deremo-Snyder
King Solomon
Nil
Schwartzwalder
Company
Energy Fuels Nuclear, Inc.
Energy Fuels Nuclear, Inc.
Energy Fuels Nuclear, Inc.
UMETCO Minerals Corp.
UMETCO Minerals Corp.
UMETCO Minerals Corp.
UMETCO Minerals Corp.
Cotter Corp.
Type
Breccia-pipe
Breccia-pipe
Breccia-pipe
Modified Room
and Piilar
Modified Room
and Pillar
Modified Room
and Pillar
Modified Room
and Pillar
Modified Room
and Pillar with
Vein Structure
Expected Current Ore
Life Production Rate
(y) (MT/d)0>>
6 270-360
6 270-360(0
3 270-360
(e) (e)
(e) 280
(e) 350(0
(e) 50
Standby(8) 0
(a) The types of underground mines are discussed in Section 11.1.1.
(b) MT/d - metric tons per day; 1 short ton - 0.907 metric ton.
(c) Predicted production.
(d) Assumed but unconfirmed.
(e) Information not available.
(f) Based on Jo89 and 260 production days per year. In some cases, quantities may reflect earlier
rather than the current production rates.
(g) Exploration is in progress.
(h) Mine placed on standby in April 1989. Ore production prior to closing was based on producing
110,000 Ibs U30g from 0.21X grade ore during the five months prior to the mine closing (P189).
rates
Source: Pi88a
-------
Table 11-1. Currently operating underground uranium mines in the United States (continued).
State Mine Name Company
Colorado Sunday UMETCO Minerals Corp.
Colorado Wilson-Silverbell UMETCO Minerals Corp.
New Mexico Mt. Taylor Chevron Resources Co.
New Mexico Section 23 Homestake Mining Co.
£ Utah La Sal UMETCO Minerals -Corp .
1
U)
Utah Snowball -Pandora UMETCO Minerals Corp.
Wyoming Sheep Mountain #1 U.S. Energy Co.
Type
Modified Room
and Pillar
Modified Room
and Pillar
Modified Room
and Pillar
Modified Room
and Pillar
Modified Room
and Pillar
Random Drifting
Expected Current Ore
Life Production Rate
)
(e)
(e)
20
1.25
(e)
(e)
5
200
-------
11.1.1 Process Description
Fifteen underground uranium mines are included in this
assessment. Included are the Mt. Taylor and Section 23 mines
which utilize the modified room and pillar method of underground
mining; the Schwartzwalder mine which uses the modified room and
pillar method in conjunction with vein structure mining; the
Sheep Mountain No. 1 mine which uses random drifting with short
cross-cut drifts; and the Pigeon, Kanab North, and Pinenut mines
which apply a different mining technique to recover the vertical
breccia-pipe deposits. Although unconfirmed, UMETCO Minerals
Corporation is believed to use the modified room and pillar
method to remove ore from their mines. Irrespective of the
mining method, the principal radioactive effluent in the mine
ventilation air is radon-222 which is released during mining
operations.
11.1.1.1 The Modified Room and Pillar Method
In this method, a large diameter main entry shaft is drilled
to a level below the ore body. A haulage way is then established
underneath the ore body. Vertical raises are driven up from the
haulage way to the ore body. Development drifts are driven along
the base of the ore body connecting with the vertical raises.
Mined ore is hauled along the development drifts to the vertical
raises and gravity fed to the haulage way for transport to the
main shaft and hoisting to the surface.
Ventilation air generally enters the mine through the main
shaft and is vented through one or more shafts installed at
appropriate distances along the ore body. Typical ventilation
flow rates are on the order of 200,000 to 400,000 cfm.
11.1.1.2 Vein Structure Mining
When ore deposits follow faults, vein structure mining is
often applied, as at the Schwartzwalder mine. This involves a
combination of methods including shrinkage and sublevel stoping
for vertical veins and open stoping with random pillars for
inclined and horizontal veins. Ore, broken by drilling and
blasting, is gravity fed through draw cones to the haulage level
and moved out through the shaft or horizontal adits. Most of the
mined-out stopes are interconnected; however, bulkheads and air
doors are extensively used throughout the mine to control air
flow.
11.1.1.3 Breccia-Pipe Mining
Breccia-pipe deposits of uranium ore are circular, chimney-
like masses of highly fragmented rock mineralized at various
levels from solutions precipitating uranium and other minerals.
Each breccia-pipe is separate and discrete and when exploited,
becomes an individual mine.
11-4
-------
A single, large shaft is driven vertically outside the
breccia-pipe to a depth exceeding the deposit. A horizontal
haulage drift extends from the shaft to beneath the breccia-pipe.
Ore, broken by drilling and blasting, falls to the haulage drift
below where it can be removed. Horizontal drifts are constructed
at regular intervals from the shaft to the ore deposit to provide
access to the ore and ventilation. A large, ovate, chimney-like
void extending hundreds of feet high remains after mining is
completed.
Mine ventilation air is forced by surface fans through lined
boreholes to the bottom level of the mine and then diverted to
each level. Exhaust air at the Pigeon and Pinenut mines is
diverted out a 2.6-m diameter horizontal duct centered 1.5 m
above ground surface. Due to the proximity of the Kanab North
ore body to the Kanab Creek Canyon, all mine exhaust is
discharged through a 3.05 m X 4.58 m horizontal adit in the
canyon wall, about 280 m below the canyon rim.
11.1.2 Existing Control Technology
The only technology presently in use to control the
emissions of radon-222 from underground mines is the bulkheading
of mined-out areas. Permanently installed bulkheads are
presently used in all operating mines except the breccia-pipe
mines. This technology was initially used in reducing radon and
radon progeny in the mine atmosphere and, thus, exposure to
miners. Regulations delineating the requirements for bulkheading
were promulgated under 40 CFR 61, Subpart B, on April 17, 1985.
However, the effectiveness of bulkheads in reducing radon
emissions from underground mines is much less than earlier
estimates projected (EPA85). It is now believed unlikely that
any of the operating mines can achieve any significant additional
reduction in radon-222 emissions by the use of bulkheads (see
Section 11.4.1).
11.2 BASIS OF THE EXPOSURE AND RISK ASSESSMENT
11.2.1 Radon-222 Emissions
Radon-222 is the radionuclide emitted from underground
uranium mines that causes the greatest health risk. The major
source of radon-222 emissions to air is the mine vents through
which the ventilation air is exhausted. 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.
In addition to the mine vents, radon-222 is emitted to air
from several aboveground sources at an underground uranium mining
operation. These sources are the ore, sub-ore, and waste rock
storage piles, as well as the loading and dumping of these
materials. The Pacific Northwest Laboratory has estimated the
11-5
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radon-222 emissions from these sources to be about 2 to 3 percent
of the emissions from the vents (Ja80). The EPA has estimated the
emissions from the aboveground sources to be about 10 percent of
mine vent emissions (see Table 11-2).
The aboveground sources also emit radionuclides to air as
particulates. The particulate emissions result from ore dumping
and loading operations, wind erosion of storage piles, and
vehicular traffic. The 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 times less) than the risks
from radon-222 emissions (EPA83b). Therefore, the health risk
assessment presented in the subsequent sections of this chapter
will be limited to radon-222 emissions from the mine exhaust
vents.
Table 11-2. Estimated annual radon-222 emissions from
underground uranium mining sources (EPA83b).
Average Large Mine(a)
Source (Ci/y)
Underground
Mine vent air 3,400
Abovearound
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
(a) Ore grade =0.1 percent U3Os. Annual production of ore and
sub-ore = 2 X 105 MT, and waste rock = 2.2 X 104 MT.
Table 11-3 presents the parameters describing radon-222
emissions at the 15 assessed underground uranium mines. Measured
radon-222 concentrations in mine ventilation exhaust air were
available for only the Section 23, Mt. Taylor, and Schwartzwalder
mines. Only the radon decay product concentrations, in terms of
working levels (WL), had been measured in ventilation exhaust air
11-6
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Table 11-3. Radon-222 concentrations and annual release rates in mine
ventilation exhaust air.
Mine
Exhaust Exhaust Radon in Exhaust Annual Radon
Vent Rate, cfm Air, pCi/1 Release, Ci/y(a)
Section 23
TOTAL
Mt. Taylor
Schwartzwalder
TOTAL
Kanab North
Pigeon
Pinenut
Sheep Mountain # 1
TOTAL
1
2
3
4
5
6
7
8
9
10
11
12
13
1
1
2
3
4
5
1
1
1
14
173-49
162-60
146-65
146-46
48,381
45,959
44,426
15,950
16,000
36,250
28,640
39,656
17,973
44,528
20,599
Unknown
18,327
376,700
563,000
81,200
85,000
67,100
13,400,
103,200
349,900
200,000
265,000
43,000
13,000
22,000
43,000
43,000
43,000
164,000
8,085
17,541
534
2,968
9,488
3,755
2,173
4,026
3,510
1,928
6,388
30
13,241
260
1,527
1,419
1,268
10
958
550 (b)
650 (b)
550 (b)
205 (b)
100 (b)
15 W
123 (b)
13 (b)
1,728
3,562
105
209
671
601
275
705
279
379
116
50
214
8,894
2,180
1,847
1,796
1,267
2
1,473
6,385
1,640
2,560
350
40
33
10
79
8
170
(a) All mine releases, except those for Section 23 and the UMETOO mines (Jo89),
are based on continuous operation.
(b) Based on WL measurements in exhaust air and an equilibrium fraction of
radon decay products to radon of 0.20.
(c) Obtained from Jo89. Lists total exhaust vents and portals at mine.
(d) Total estimated exhaust rate from all vents at mine (Jo89).
(e) Based on 2080 hours per year operation (Jo89).
(f) Based on 4160 hours per year operation (Jo89).
11-7
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Table 11-3. Radon-222 concentrations and annual release rates in mine
ventilation exhaust air (continued).
Exhaust Exhaust Radon in Exhaust Annual Radon
Mine Vent Rate, cfm Air, pCi/1 Release, Ci/y(a)
King Solomon 13 (c)
Sunday 12 (c)
Deremo-Snyder 11 (c)
Wilson-Silverbell 7 (c)
La Sal 5(°)
Snowball-Pandora 4 (c)
Calliham l(c)
Nil 3 (°)
880, 000 (d)
680, 000 (d)
420, 000 (d)
345, 000 (d)
535, 000 (d)
635, 000 (d)
115, 000 (d)
300, 000 (d)
650 (b)
650 (b)
650 (b)
650 (b)
650 (b)
650 (b)
650 (b)
650 (b)
2,020(e)
3,120(f)
960 (e>
790 (®)
2,460(f)
2,920(f)
260 (e)
690 (e)
(a) All mine releases, except those for Section 23 and the UMETOO mines (Jo89),
are based on continuous operation.
(b) Based on WL measurements in exhaust air and an equilibrium fraction of
radon decay products to radon of 0.20.
(c) Obtained from Jo89. Lists total exhaust vents and portals at mine.
(d) Total estimated exhaust rate from all vents at mine (Jo89).
(e) Based on 2080 hours per year operation (Jo89).
(f) Based on 4160 hours per year operation (Jo89).
at the other 12 mines. These working level concentrations, in
conjunction with information on the radon-radon decay product
equilibria, were used to estimate the radon-222 concentrations in
the mine exhaust air at the mines reporting working-level
concentrations.
The concentration of radon-222 progeny measured in the
exhaust vents at the Pigeon and Kanab North mines were 1.3 WL and
1.1 WL, respectively. Using these concentrations with an assumed
equilibrium fraction of 0.20, believed to be reasonable
considering the ventilation characteristics of these mines and
the half-lives of the radon-222 decay products, radon-222
concentrations of 650 pCi/1 and 550 pCi/1 were estimated for the
exhaust air at the Pigeon and Kanab North mines, respectively.
No radioactivity measurements were available from the Pinenut
mine. For this mine, the radon-222 concentration is assumed to
be equal to that of the Kanab North mine, 550 pCi/1.
Mine exhaust rates and working-level concentrations were not
provided for individual exhaust vents at the eight UMETCO
Minerals Corporation mines. Rather, total mine exhaust
11-8
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parameters were provided by the company (Jo89). Using the
company's estimated working-level concentrations of 1.3 WL for
each mine and assuming an equilibrium fraction of 0.20, the
radon-222 concentration in the exhaust air from each mine was
estimated to be 650 pCi/1.
The measured working-level concentrations listed below were
used with an assumed equilibrium fraction (0.20) to estimate the
radon-222 concentration in the mine air from each exhaust vent
of the Sheep Mountain No. 1 mine.
Vent No. Average WL (Pi89)
14 0.41
173-49 0.20
162-60 0.03
146-65 0.245
146-46 0.025
The annual release rates of radon-222, the source terms, are
given for each mine in the last column of Table 11-3. These
estimated annual emission rates were calculated by multiplying
the concentrations by the annual volume of air exhausted. The
resulting emission rates are expressed in Ci/y. For example, the
radon-222 emission rate for the Mt. Taylor mine is obtained by
the following expression:
Emission Rate (Ci/y) = 260 pCi/1 x 28.316 I/ft3 x 563,000 ft^/min
x 5.26 x 105 min/y x 10~12 Ci/pCi
= 2,180 Ci/y.
The annual release rates at the Schwartzwalder, Section 23,
and Sheep Mountain No. 1 mines are the sums of the release rates
at the 5, 13, and 5 vent clusters, respectively. The annual
radon-222 emissions from underground uranium mines are estimated
to range from about 170 Ci/y to a maximum of 8,900 Ci/yr. The
total emissions from all 15 mines are approximately 35,400 Ci/y.
11.2.2 Health Impact Assessment
This section contains an assessment of the risk of cancer
caused by radon-222 emissions from underground uranium mines. The
health impact assessment addresses the following specific topics:
1. working level exposure and the lifetime fatal cancer
risk to the maximum exposed individual from radon-222 at
each underground uranium mine; and
2. the number of fatal cancers committed per year in the
regional population (the total number of people who
reside within 80 km of a mine) at each underground mine
due to radon-222.
11-9
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All lung cancers resulting from the inhalation of radon-222
progeny are considered fatal.
The location of the maximum exposed individual at each mine
was estimated by analyzing onsite visit reports (Pi88a, Pi89),
company reports (Jo89), and U.S. Geological Survey maps. The
AIRDOS-EPA (Mo79) and DARTAB (Be81) codes were used to estimate
the exposure to radon-222 and the increased chance of lung cancer
for individuals who reside at these selected locations. The
radon-222 decay product equilibrium fractions at these residences
were determined as a function of the distance from the mine
vents.
Collective risks for the regional population due to
radon-222 were calculated from the annual collective exposures
(person WLM) using AIRDOS-EPA (Mo79) and DARTAB (Be81) codes.
The population distribution within 80 km of each mine was deter-
mined using the computer program SECPOP (At74), which uses 1980
census data to compute the population in each annular sector.
Collective exposures to radon-222, expressed in person WLM,
were estimated for each mine by multiplying the estimated
radon-222 progeny concentration (WL) in each annular sector by
the population in that sector and by the conversion factor
51.56 WLM/y per WL. The cumulative WL exposure of each
population segment was adjusted using a radon progeny equilibrium
fraction that is related to the distance from the mine vent to
the population segment. The locations of individual exhaust
vents were not available for the UMETCO Minerals Corporation
mines. For these eight mines, longitude/latitude locations of the
"mine complex" were used to determine these distances (Sa89).
The parameters used in the AIRDOS-EPA code for each underground
mine are listed in Appendix A.
The location of the maximum exposed individual is presented
as the distance, m, from the mine ventilation exhaust vent. A
single discharge point was assumed for the multiple vented mines,
Schwartzwalder, Sheep Mountain No. 1, and Section 23. It was
located approximately in the center of the multiple vents with a
bias toward those with larger emissions. The mine ventilation
exhaust vents are described in Section 11.4.4. Vents that are
horizontally oriented were all assigned 1-m release heights.
11.3 RESULTS OF THE EXPOSURE AND RISK ASSESSMENT
11.3.1 Risks to Nearby Individuals
The highest individual risks for each of the 15 assessed
underground uranium mines are listed in Table 11-4 in the order
of decreasing risk. Included for each mine is the location of
the individual with respect to the distance from the mine
ventilation exhaust vent and the radon-222 concentration and
working-level exposure at that location. Maximum lifetime
individual risks ranged from about 3E-6 at the Pinenut mine near
11-10
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Table 11-4. Estimated exposures and risks to individuals living near
underground uranium mines.
Mine/location
Maximum
Radon Maximum Maximum Lifetime
Oonoentration Exposure Fatal Cancer Risk Distance (a)
(pCi/1) (WL) to Individual (meters)
La Sal -
Near La Sal, UT
Deremo-Snyder -
Near Egnar, CO
Snovfoall-Pandora -
Near La Sal, UT
l.OE+0
4.1E-1
2.6E-1
Schwartzwalder - 2.5E-1
13 km NW Golden, CO
Calliham -
Near Egnar, CO
2.6E-1
Section 23 - 5.0E-2
50 km N Grants, NM
3.1E-3
1.2E-3
9.1E-4
8.3E-4
7.6E-4
3.0E-4
King Solomon -
Near Uravan, CO
Wilson-Silverbell -
Near Egnar, CO
Sunday -
Near Naturita, CO
Nil -
Paradox Valley, CO
6.2E-2
7.0E-2
5.1E-2
1.1E-2
Pigeon - 6.4E-3
24 km S Fredonia, AZ
Mt. Taylor - 4.1E-3
50 km NE Grants, NM
Kanab North - 2.6E-3
30 km SSW Fredonia, AZ
Sheep Mountain No. 1 - 1.1E-3
12 km S Jeffrey City, WY
Pinenut - 2.8E-4
53 km SSW Fredonia, AZ
2.6E-4
2.5E-4.
2.4E-4
5.4E-5
4.5E-5
2.7E-5
1.8E-5
4.7E-6
4E-3
2E-3
1E-3
1E-3
1E-3
4E-4
4E-4
3E-4
3E-4
7E-5
6E-5
4E-5
2E-5
6E-6
800
800
2,000
1,400
500
12,800
4,000
2,000
6,300
6,300
24,000
15,000
30,000
5,200
2.0E-6 3E-6 53,000
(a) Distance from the exhaust vent to the maximum exposed individual.
11-11
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Fredonia, Arazona to a high of 4E-3 at the La Sal mine near La
Sal, Utah. The magnitude of the risk is most often controlled by
either the source term or the distance and direction of the
individual's residence from the mine vent. The larger risks
estimated for some UMETCO Minerals Corporation mines are due not
only to small distances to the nearby individuals, but also to
the arbitrary positioning of the nearby individual in the
predominant downwind direction from the mine. This was done
because of the absence of directional information for nearby
individuals at these mines and probably overestimated the risk to
the maximum exposed individual in most cases.
The individual risks estimated for underground uranium mines
in the 1984 EPA assessment (EPA85) were significantly higher than
those estimated here. The primary reason for this decrease is
the depressed condition of the industry which has resulted in
many mines closing and large numbers of people moving from these
regions. Since many of the people living near the mines moved
away, distances between mines and populations have increased.
For example, at the time of the earlier assessment, many individ-
uals lived within 500 m of a mine vent. Now, only one individual
lives within 500 m of a mine vent, and only four live within
1,000 m. However, one of these individuals (located 700 m SW of
the Mt. Taylor mine) is not at maximum risk due to the height
(20 m) of the exhaust stack, plume buoyancy, and the very low wind
frequency (0.9 percent) in the direction of that individual.
11.3.2 Risks to the Regional Populations
The collective risks of fatal lung cancer resulting from
radon-222 emissions occurring in the regional 80-km population
around each underground uranium mine are listed in Table 11-5 in
the order of decreasing risk. Also listed are the 1980 census
populations within the 80-km regions. The highest collective
risk occurs in the densely populated Denver/Golden, Colorado,
area where it is estimated that a fatal lung cancer will occur
about every year due to the radon-222 emissions from the
Schwartzwalder mine. The collective risks within regional
populations at the other mines are much lower, primarily because
fewer people live within the 80-km regions. For example, it is
estimated that radon-222 released from the Section 23 and King
Solomon mines, those falling second and third in the ordered
listing, will result in only one fatal lung cancer every 20 and
200 years, respectively.
An additional output of the DARTAB computer code provides
the frequency distribution of lifetime fatal cancer risks around
each mine. It predicts the number of people in each of a series
of lifetime risk intervals and the number of cancer deaths that
occur annually within each risk interval. The individual distri-
butions were combined into an overall distribution of lifetime
fatal cancer risks around all underground uranium mines. The
11-12
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Table 11-5.
Estimated committed fatal cancers per year due to
radon-222 emissions from underground uranium mines.
Committed
Fatal
Mine and
Location
Schwartzwalder
13 km NW Golden, CO
Section 23
50 km N Grants, NM
King Solomon -
Near Uravan, CO
Snowball -Pandora -
Near La Sal, UT
Sunday -
Near Naturita, CO
La Sal -
Near La Sal, UT
Mt. Taylor
50 km NE Grants, NM
Pigeon
24 km S Fredonia, AZ
Nil -
Paradox Valley, CO
Deremo-Snyder -
Near Egnar, CO
Kanab North
30 km SSW Fredonia,
Wilson-Silverbell -
Near Egnar, CO
Calliham -
Near Egnar, CO
Sheep Mountain No. 1 -
12 km S Jeffrey City,
Pinenut
53 km SSW Fredonia,
1980 Population
Within 80 km
1,800,000
65,000
67,000
21,000
24,000
21,000
50,000
7,800
55,000
30,000
11,000
AZ
30,000
30,000
5,200
WY
8,200
AZ
Cancers Per Year
(0-80 km)
7E-1
5E-2
5E-3
4E-3
4E-3
3E-3
3E-3
2E-3
2E-3
1E-3
1E-3
1E-3
4E-4
2E-4
2E-4
11-13
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distribution is shown in Table li-6. The distribution reflects
the number of deaths expected to occur annually within the 0-80
km population listed due to radon-222 emissions from underground
uranium mines. For example, about 2,200,000 people are at risk
within the 15 regions due to their exposure to radon-222 from all
underground uranium mines, and within this population, about 0.8
fatal lung cancers are expected to occur per year. Of the pre-
dicted deaths per year caused by emissions of radon-222 from
underground mines, about 90 percent are attributable to the
Schwartzwalder mine.
Table 11-6. Estimated distribution of the fatal cancer risk
caused by radon-222 emissions from all underground
uranium mines.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
86,
1,600,
450,
51,
0
0
5
000
000
000
000
0
0
1E-4
2E-1
6E-1
3E-2
4E-4
Totals 2,200,000
8E-1
11.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS
A number of methods to control radon emissions from
underground uranium mines have been evaluated. These are: (1)
bulkheading; (2) use of a sealant coating on exposed ore
surfaces; (3) activated carbon adsorption of radon from
contaminated mine air; (4) extending the height of the mine air
exhaust stacks; and (5) other control technologies. Also
considered are the design and development of new underground
mines in a way that will limit the diffusion of radon into the
mine air. Brief descriptions of these technologies and their
effectiveness with costs, in 1988 dollars, are presented below.
11.4.1 Bulkheading
This method reduces radon emissions by sealing off
(bulkheading) openings to worked-out areas of the mine. The radon
emanating from these areas of the mine will decay in the sealed-
off area rather than be discharged into the outside air. A
bulkhead is an air-restraining barrier, usually consisting of a
timber or metal stud frame covered with timber, expanded metal
lath, plywood, or other sheet products. Concrete or cinder blocks
are also sometimes used. A sealant (polyurethane, shotcrete,
11-14
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etc.) is usually applied to the structure and to the joints
between the structure and the rock to form a continuous seal.
Airtight bulkheads can seldom be achieved. Most bulkheads
leak to some extent because the mine is under a negative pressure
causing air flow through the bulkhead and the fractured and
porous rock near the bulkhead. Since the radon in the sealed
area behind a bulkhead will build up to relatively high
concentrations (i.e., tens of thousands of picocuries per liter),
it is necessary to prevent or minimize any leakage of air from
behind the bulkhead into the working areas of the mine. Any such
leakage could significantly increase the radon decay product
concentration to which the miners are exposed. Therefore, it is
often necessary to maintain a negative differential pressure
behind the bulkhead to prevent leakage of contaminated air into
the active mine airways. This negative pressure is achieved by
bleeding (i.e., removing) air from behind the bulkhead into an
exhaust airway. For bulkheads to be effective in reducing radon
emissions to aboveground air, however, the amount of air bleed
necessary to maintain an adequate differential across the
bulkhead must be managed. The smaller the air bleed, the more
radon will decay behind the bulkhead rather than being released
above ground.
Several theoretical evaluations of the effectiveness of
bulkheads in reducing radon emissions to air from active
underground uranium mines have been conducted (Ko80, B184). One
study of a model mine (Ko80) estimated that bulkheading would
achieve a 14 percent reduction in radon emissions at a cost of
$0.15 per pound of uranium oxide ($0.45 per ton of 0.15 percent
uranium ore). In this study,'each stope is bulkheaded upon
completion of the mining activity with a 50 percent daily air
volume bleed. Another study of 13 case mines (B184) estimated
that bulkheading could reduce radon emissions by about 60 percent
for a few cents per pound of uranium oxide. In this study,
80 percent of the surface area is sealed off with a 10 percent
daily air volume bleed.
Both of these studies are based on extensive bulkheading of
the mines and a controlled air bleed behind the bulkhead. None
of the existing mines can meet the conditions needed to achieve
radon emission reductions through the use of bulkheads. Some of
the factors involved are the following:
1. many worked-out areas of the mines are used as
ventilation passageways or emergency escapeways and
cannot be sealed off;
2. many worked-out areas are not accessible for bulkhead
installation and maintenance because of safety hazards;
3. for the breccia-pipe mines, the mining method precludes
the use of bulkheads; and
11-15
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4. for all of these mines, limiting the amount of air
removed from behind the bulkheads is not practical.
11.4.2 Sealant Coatings
This method reduces radon emissions by preventing the radon
from entering mine air by sealing the exposed mine surfaces.
These sealants include a large group of industrial polymer
chemical products which form thick adhesive coatings. A two- or
three-layer system has been shown to produce the most favorable
results, with shotcrete as the base coating (B184, Ko80, FrSla).
Laboratory studies have shown these sealants to be very
effective in reducing radon emanations from uranium ore surfaces.
However, the presence of pinholes and the difficulty in applying
a perfect coating on the surface considerably reduce the
effectiveness of the sealants.
Field studies in inactive test mines have demonstrated that
some rock surfaces can be sealed to reduce radon emanations by up
to 75 percent (Ko80). No field studies have been conducted to
measure the effectiveness of sealants in reducing radon emissions
in active mines.
Several theoretical studies of the effectiveness of sealants
in reducing radon emissions from active uranium mines have been
conducted (Ko80, B184). One study of a model underground uranium
mine (Ko80) estimated that sealants could achieve about an
11 percent reduction in radon emissions at a cost of $0.63 per
pound of uranium oxide ($1.90 per ton of 0.15 percent uranium
ore). Only the development drifts are sealed in this model mine
at a unit cost of about $0.88 per ft2.
In another study of 13 case mines (B184), it was estimated
that sealants could achieve about a 56 percent reduction in the
radon emissions at a cost of $0.53- $3.75 per pound of uranium
oxide. In this study, 80 percent of the mine surface was consid-
ered to be sealed at a unit cost of $1.03 per ft2.
For reasons discussed below, it was not practical for any of
the existing mines to apply sealants to 80 percent of the mine
surfaces. The first study (Ko80) is believed to provide a more
realistic estimate of the potential radon emission reductions
achievable in some mines by applying sealants.
Although sealants have been shown to reduce radon emanations
from rock surfaces under experimental conditions, the use of this
technique to reduce radon emissions from active underground
uranium mines is significantly limited for these reasons:
1. Sealants cannot be applied to many areas of the existing
mines because:
(a) active drifts or stopes cannot be sealed due to the
mining activities;
11-16
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(b) most mined-out areas cannot be entered due to safety
hazards; and
(c) floors, haulageways, and areas with significant
vehicular traffic cannot be effectively sealed.
2. Pinholes in the sealant will act as a conduit for the
radon and reduce some of the effectiveness of the
sealant. Perfect bonding cannot be assured, and radon
will migrate behind the skin of the sealant and escape
through a pinhole or along the rib flow junction.
3. Application on rock surfaces is limited because of hot
rock surfaces and water inflow through the rock
surfaces.
4. Geological conditions are not conducive to good sealant
application. Rock stress is often high, causing the rock
to crack and slabs to break away from roof and sides.
For these reasons, the use of sealants to reduce radon in
existing underground uranium mines is not widely applicable. The
method is extremely limited and can achieve only small radon
reductions.
11.4.3 Adsorption on Activated Carbon
The bleeder pipes used to achieve negative pressure behind
bulkheads (see Section 11.4.1) release significant quantities of
radon into the exhaust ventilation system of a mine. A possible
solution to the problem is to integrate an activated carbon trap
into the system that removes the radon before it enters the mine
ventilating system.
Several activated carbon systems have been investigated
(Ko80, B184). In general, air from the bleeder pipe is first
filtered to remove dust particles and radon decay products and
then passed through an activated carbon trap. A dehumidifier can
be placed in the system before the carbon trap if the humidity in
the mine is high. The carbon trap is periodically regenerated by
passing hot air through the trap, collecting the eluted radon in
a second carbon trap. The efficiency of the system is very
dependent on moisture, temperature, and the flow rate of air
through the trap. About 100 CFM is generally considered an upper
flow rate limit (Ko80).
A theoretical evaluation of the effectiveness of activated
carbon systems in reducing radon emissions from underground
uranium mines has been conducted (Ko80). The study, based on a
model mine, estimated that the use of activated carbon traps on
bulkhead bleeder pipes would achieve a 35 percent reduction in
the radon emissions from the mine at a cost of $1.92 per pound of
uranium oxide ($5.75 per ton of 0.15 percent uranium ore). In
this case study, 12.5 carbon systems were operated, each treating
100 CFM of air.
11-17
-------
The use of activated carbon systems to remove radon from all
air exhausted from an underground uranium mine was also theoreti-
cally evaluated (B184). The study assumed that seventy-two
5,000 CFM carbon adsorption units would be required to accommo-
date a mine ventilation rate of 360,000 CFM. It was estimated
that these systems would result in an annual cost of over
$55 million or about $100 per pound of uranium oxide ($300 per
ton of 0.15 percent uranium ore). The enormous size of this
system, the radiation potential resulting from the buildup of
radon and its decay products on the traps, and the costs render
this approach infeasible.
Although activated carbon adsorption applied to bulkhead
bleeder pipes appears technically feasible, none of the existing
mines are using these systems to reduce radon emissions. These
systems have not been employed because of the numerous disadvan-
tages associated with them. Some of these disadvantages are:
1. The systems require continuous attention by trained
personnel.
2. Skilled operators, usually not available in mining
communities, are required to operate and maintain the
systems.
3. High humidity in mine atmospheres significantly reduces
the effectiveness of the carbon systems.
4. Radiation hazards may be caused by the decay of radon
and its progeny that is adsorbed on the charcoal.
5. Safety problems due to the interruption of electrical
service or system malfunction can increase the radon
concentration in the mine air.
6. No commercial units applicable to mine atmospheres are
available, and further development work on the systems
is required.
Although activated carbon adsorption systems may be a feasi-
ble technology for removing radon from bulkhead bleeder tubes,
the systems have not been shown to be practical in an underground
mine atmosphere for technical, safety, and economic reasons.
11.4.4 Mine Ventilation Exhaust Stacks
Increasing the vertical heights of mine ventilation exhaust
stacks will reduce the ground-level radon concentration near the
stacks (Dr80). The exposure to radon and, therefore, risk
to people living relatively near the exhaust vents can be reduced
by increasing the height of the ventilation exhaust stacks. The
ventilation exhaust vents at the 15 assessed underground mines
are described in Table 11-7. Except for the Mt. Taylor mine,
which has a 20-m stack, mines presently release emissions at
11-18
-------
about 1 to 2 m. In order to implement this control technology,
mines with multiple stacks must consider extending more than one
stack. Also, mines that vent horizontally to a canyon wall have
additional problems in extending their exhaust stacks vertically.
Table 11-7. Current mine ventilation exhaust vents.(a)
Mine
Pigeon
Pinenut
Kanab North
Section 23
Mt. Taylor
Schwartzwalder
Sheep Mountain #
King Solomon
Sunday
Deremo-Snyder
Wilson-Silverbell
Calliham
Nil
La Sal
Snowball -Pandora
Number of Vent
Exhaust Orientation
Vents
1 Horizontal
1 Horizontal
1 Horizontal
4 Vertical
9 Vertical
1 Vertical
4 Horizontal
1 Vertical
1 5 Horizontal
13 Vertical
12 Vertical
11 Vertical
7 Vertical
1 Vertical
3 Vertical
5 Vertical
Diameter Approximate
of Vent Vent Height
(m) (m)
2.44 1.5
2.44 1.5
2.44(b) Canyon Wall
1.22 2.3
1.83 2.3
7.32 20
2.44(b) l-2(c)
2.44 1-2
0.91-1.52 2.4
2.44 1.2
2.44 1.2
2.44 1.2
2.44 1.2
2.44 1.2
1.83 1.2
2.14 1.2
4 Vertical 2.14 1.2
(a) Exhaust vent data from Pi88a, Pi89, Jo89, and Sa89.
(b) These are actually rectangular vents having an effective area
approximately the same as a 2.44 diameter opening.
(c) Two vents exhaust to a canyon wall.
11-19
-------
To determine the benefit of higher emission release heights,
the reduction in the radon concentration and risk was evaluated
at the location of the maximum exposed individual at each operat-
ing underground uranium mine for exhaust stack heights of 10 m,
20 m, 30 m, and 60 m. The results of this study are presented in
Table 11-8. Also listed in the table are the radon concentra-
tions and risks estimated using the current (baseline) stack
heights (see Table 11-7) and assuming all are vented vertically.
The percent reductions in the radon concentration and life-
time individual risk at each release height are similar at each
mine, except at the Sheep Mountain No. 1 mine. The maximum
exposed individual was located at a much greater distance at the
higher release heights which allowed for more dilution (reducing
the radon concentration) but provided a longer time, producing a
higher progeny/radon equilibrium fraction. Increasing the re-
lease heights had only a limited effect on the cumulative risks
to the regional populations. The percent reduction in the radon
concentration and lifetime individual risk with increasing re-
lease height was greatest when the distance from the mine to the
maximum exposed individual was small and least when the distance
was large.
To illustrate this, the range and average percent reduction
determined for each release height at 14 of the underground mines
are shown in Table 11-9. The mines are divided into three
categories depending upon the distance from the mines to the
maximum exposed individual: small distances (1,400 m or less);
long distances (24,000 m and greater); and intermediate dis-
tances. The Mt. Taylor mine was not included in this summary
since the mine currently operates with a 20-m stack height.
These results show that increasing the height of the mine exhaust
stack is very effective in reducing the radon concentration and
risk when small distances exist between the mine and the individ-
ual. However, the effectiveness decreases with distance and
becomes of marginal value at long distances.
11-20
-------
Table 11-8.
Estimated lifetime fatal cancer risk to the maximum
exposed individual and the committed fatal cancers
per year due to radon-222 emissions from underground
uranium mines as a function of vent stack height.
Stack Location of Concentration of
Height, m Individual, m Radon-222, pCi/1
Baseline (a)
10
20
30
60
Baseline (b)
10
20
30
60
Basel ine(a)
10
20
30
60
Baseline (a)
10
20
30
60
Baseline (°)
10
20
30
60
(a) Baseline
(b) Baseline
(c) Baseline
1,400
1,400
1,400
1,400
1,400
12,800
12,800
12,800
12,800
12,800
24,000
24,000
24,000
24,000
24,000
30,000
30,000
30,000
30,000
30,000
15,000
15,000
15,000
15,000
15,000
height -
height -
height -
Schwartzwalder Mine
2.5E-1
2.1E-1
1.4E-1
8.7E-2
3 . 1E-2
Section 23 Mine
4.9E-2
4.5E-2
3.9E-2
3.2E-2
1.5E-2
Pigeon Mine
6.4E-3
6.3E-3
5.9E-3
5.3E-3
3.2E-3
Kanab North Mine
2.6E-3
2.6E-3
2.4E-3
2.2E-3
1.3E-3
Mt. Tavlor Mine
4.1E-3
5.4E-3
4.1E-3
3.1E-3
1.5E-3
1 . 0 meters
2 . 0 meters
20 meters
Lifetime
Risk to
Individual
1.2E-3
9.6E-4
6.4E-4
4.0E-4
1.4E-4
4.1E-4
3.8E-4
3.2E-4
2.6E-4
1.2E-4
6.1E-5
5.9E-5
5.6E-5
5.0E-5
3.0E-5
2.4E-5
2.4E-5
2.3E-5
2.0E-5
1.2E-5
3.6E-5
4.8E-5
3.6E-5
2.7E-5
1.3E-5
Committed
Fatal
Cancers
Per Year
(0-80 km)
7.1E-1
6.9E-1
6.5E-1
5.9E-1
3.9E-1
4.7E-2
4.4E-2
3.8E-2
3.2E-2
1.7E-2
2.2E-3
2.1E-3
2.0E-3
1.8E-3
1.2E-3
1.3E-3
1.2E-3
1.2E-3
1.1E-3
6.8E-4
3.1E-3
4.0E-3
3.1E-3
2.5E-3
1.4E-3
11-21
-------
Table 11-8.
Estimated lifetime fatal cancer risk to the maximum
exposed individual and the committed fatal cancers
per year due to radon-222 emissions from underground
uranium mines as a function of vent stack height
(continued).
Stack Location of Concentration of
Height, m Individual, m Radon-222, pCi/1
Basel ine(a)
10
20
30
60
Baseline (b)
10
20
30
60
Baseline (a)
10
20
30
60
Baseline (a)
10
20
30
60
Baseline (a)
10
20
30
60
(a) Baseline
(b) Baseline
(c) Baseline
53', 000
53,000
53,000
53,000
53,000
5,200
12,650
12,650
12,650
12,650
4,000
4,000
4,000
4,000
4,000
6,300
6,300
6,300
6,300
6,300
800
800
800
800
800
height -
height -
height -
Pinenut Mine
2.8E-4
2.8E-4
2.6E-4
2.4E-4
1.5E-4
Sheep Mountain No. 1
1.1E-3
7.6E-4
7.1E-4
6.3E-4
3.6E-4
Kincj Solomon
6.2E-2
5.9E-2
5.1E-2
4.1E-2
1.6E-2
Sunday
5.1E-2
4.9E-2
4.4E-2
3.7E-2
1.7E-2
Deremo-Snyder
4.1E-1
2.9E-1
1.3E-1
6.0E-2
1.4E-2
1 . 0 meters
2.0 meters
20 meters
Lifetime
Risk to
Individual
2.7E-6
2.6E-6
2.5E-6
2.3E-6
1.4E-6
6.5E-6
6.3E-6
5.9E-6
5.2E-6
3.0E-6
3.5E-4
3.4E-4
2.9E-4
2.3E-4
8.9E-5
3.3E-4
3.2E-4
2.9E-4
2.4E-4
1.1E-4
1.7E-3
1.2E-3
5.4E-4
2.5E-4
6.0E-5
Committed
Fatal
Cancers
Per Year
(0-80 km)
1.7E-4
1.6E-4
1.5E-4
1.4E-4
9.1E-5
1.7E-4
6.4E-4
1.5E-4
1.4E-4
7.8E-5
5.4E-3
5.3E-3
5. OE-3
4.6E-3
2.9E-3
3.5E-3
3.4E-3
3.3E-3
3. OE-3
1.9E-3
1.3E-3
1.3E-3
1.2E-3
1.1E-3
6.8E-4
11-22
-------
Table 11-8.
Estimated lifetime fatal cancer risk to the maximum
exposed individual and the committed fatal cancers
per year due to radon-222 emissions from underground
uranium mines as a function of vent stack height
(continued).
Stack Location of Concentration of
Height, m Individual, m Radon-222, pCi/1
Lifetime
Risk to
Individual
Committed
Fatal
Cancers
Per Year
(0-80 km)
Wilson-Silverbell
Basel ine(a)
10
20
30
60
2
2
2
2
2
,000
,000
,000
,000
,000
7
6
4
3
9
.OE-2
.4E-2
.8E-2
.2E-2
.1E-3
3.
3.
2.
1.
4.
4E-4
1E-4
3E-4
5E-4
4E-5
1.
1.
9.
9.
5.
1E-3
OE-3
9E-4
OE-4
6E-4
Calliham
Baseline (a)
10
20
30
60
Baseline (a)
10
20
30
60
Baseline (a)
10
20
30
60
Baseline (a)
10
20
30
60
(a) Baseline
(b) Baseline
(c) Baseline
500
500
500
500
500
2
1
3
1
3
. 6E-1
. 3E-1
.9E-2
.9E-2
.7E-3
1.
5.
1.
7.
1.
1E-3
2E-4
6E-4
5E-5
5E-5
3.
3.
3.
3.
1.
6E-4
5E-4
3E-4
OE-4
8E-4
Nil
6
6
6
6
6
2
2
2
2
2
,300
,300
,300
,300
,300
800
800
800
800
800
,000
,000
,000
,000
,000
height -
height -
height -
1
1
9
8
3
La
1
7
3
1
3
Snowball
2
2
1
1
3
1.0 meters
2.0 meters
20 meters
. 1E-2
. 1E-2
.8E-3
. 3E-3
.8E-3
Sal
.OE+0
.4E-1
. 3E-1
.5E-1
. 6E-2
.-Pandora
.6E-1
. 4E-1
.8E-1
.2E-1
.4E-2
7.
7.
6.
5.
2.
4.
3.
1.
6.
1.
1.
1.
8.
5.
1.
3E-5
1E-5
4E-5
4E-5
5E-5
4E-3
1E-3
4E-3
5E-4
5E-4
3E-3
1E-3
6E-4
7E-4
6E-4
1.
1.
1.
1.
1.
3.
3.
3.
2.
1.
4.
3.
3.
3.
2.
8E-3
8E-3
7E-3
6E-3
OE-3
4E-3
3E-3
1E-3
8E-3
8E-3
OE-3
9E-3
7E-3
4E-3
2E-3
11-23
-------
Table 11-9. Effectiveness of various stack heights.
BaseLine
Height to:
Percent Reduction in
Radon Concentration
Range Average
Percent Reduction in the
Individual Lifetime Risk
Range Average
Less Than or Equal to 1.400 m to Maximum Exposed Individual(a)
10 m
20 m
30 m
60 m
16-50
44-85
65-93
88-99
30
66
82
95
20-53
47-85
67-93
88-99
33
67
82
95
Between 1.400 m and 24.000 m to Maximum Exposed Individual(*>)
10 m
20 m
30 m
60 m
0-31
11-36
25-54
65-87
9
23
39
74
3-15
9-34
20-56
54-88
6
20
37
72
Greater Than or Equal to 24.000 m to Maximum Exposed Individual(c)
10 m
20 m
30 m
60 m
0- 2
7- 8
14-17
46-50
1
8
15
49
0- 4
4- 8
15-18
48-51
2
6
17
50
(a) Includes the Deremo-Snyder, Calliham, La Sal, and
Schwartzwalder mines.
(b) Includes all other mines except the Mt. Taylor mine.
(c) Includes the Pigeon, Kanab North, and Pinenut mines.
11-24
-------
The costs to extend the mine ventilation exhaust stacks at
the 15 underground uranium mines to heights of 10 m, 20 m, 30 m,
and 60 m have been estimated (Pi88b). The cost estimates, based
on the general framing plan shown in Figure 11-1, include rolled
steel plates to be used as ventilation duct extensions, structur-
al steel shapes for supports, and concrete for the foundations.
Composite costs used were $1.80 per pound for structural steel,
finished, fabricated, and erected, and $150 per cubic yard for
concrete, delivered and placed. A detailed description of the
basis for the cost estimate is given in Appendix 11-A.
Height
Figure 11-1. General framing plan of a mine ventilation exhaust
stack.
Because each mine has different exhaust characteristics that
affect the costs, primarily the number of stacks and their diame-
ters (see Table 11-7), costing was performed for each individual
mine. The estimated costs, in 1988 dollars, to extend the
heights of the exhaust stacks at each mine are given in
Table 11-10.
11-25
-------
Table 11-10,
Estimated costs (dollars) to extend the heights of
the ventilation exhaust stacks at each underground
uranium mine (Pi88b).
Stack Height
Mine
Section 23
Schwartzwalder(a)
Pigeon
Pinenut
Kanab North (a)
Mt. Taylor
Sheep Mountain
No. 1
King Solomon
Sunday
Deremo-Snyder
Wilson-Silverbell
Calliham
Nil
La Sal
Snowball-Pandora
Totals
10 Meter
222,500
93,900
31,200
31,200
31,200
(b)
70,000
405,600
374,400
343,200
218,400
31,200
55,500
124,300
99,400
2,132,000
20 Meter
507,400
241,500
80,500
80,500
80,500
(b)
159,500
1,046,500
966,000
885,500
563,500
80,500
126,600
306,800
245,400
5,370,700
30 Meter
950,700
439,800
146,600
146,600
146,600
425,500(c)
307,500
1,905,800
1,759,200
1,612,600
1,026,200
146,600
234,900
562,300
449,800
10,260,700
60 Meter
1,890,900
874,200
291,400
291,400
291,400
l,055,200(c)
612,000
3,788,200
3,496,800
3,205,400
2,039,800
291,400
467,100
1,117,500
894,000
20,606,700
I (a) Estimates do not include converting vents that exhaust
horizontally through canyon walls.
,(b) These estimates are not applicable since the current exhaust
stack height is 20 m.
(c) These estimates may be somewhat high if any part of the
present 20-m structure can be used.
11-26
-------
There are two cost items not included in Table 11-10
(Pi88b) The estimates do not include the loss of revenue caused
by the shutdown during the installation of the extended stacks
It is estimated that it would require one to two months for these
conversions, resulting in an additional cost of $0.9 to
???5 million dollars in lost revenue (mining expenses wi11 con-
Jinue near normal during this period). These costs will depend
on the period of shutdown and the production rate of the mine.
The second cost item not included in the above estimate is the
expense of installing larger fans, which may be needed to redis-
tribute the air flow underground.
Although this control alternative does not reduce the emis-
sions of radon from underground uranium mines, it is effective in
redoing the exposure and lung cancer risks to the nearby indi-
viduals from these emissions. It also, to a lesser extent,
riduceS the exposures and cumulative risks to the regional popu-
lations. This control alternative is achievable with current
technology.
11.4.5 other Control Technologies
Backfilling is the practice of filling mined-out areas of an
underground mini with waste rock which provides ground support in
the mine disposal of unwanted material without hoisting it to
the sirrace^and a reduction in the mine ventilation requirements
(FrSlb). Backfilling is practiced at the underground mines,
except at the breccia-pipe mines where the mining method prevents
Its Sse. However, because underground mining methods reduce the
ratio of waste to ore mined (only 5 to 20 percent of the mined
tonnage is available for backfilling), this control alternative
will require that material be obtained from an aboveground source
and transported underground, e.g., clasJified,.mlliTn^1"f °^on
surface sands. In a mine test of one stope, the amount of radon
released from the stope was reduced 84 percent after the stope
was 90 percent backfilled (FrSlb). In a study of 13 case mines
(B184), it was estimated that backfilling with classified mill
tailings and surface sand to the extent that would achieve an
80 Bpercent reduction in radon emissions would cost $0.85 to $9.90
per pound of uranium oxide. Therefore, it was concluded:
(1) backfilling is less cost-effective than bulkheading to reduce
radon emissions from a mine; (2) vast abandoned areas of the mines
are inaccessible to backfilling due to unsafe rock conditions;
(3) many of the worked-out areas are used as ventilation Passage-
ways or emergency escapeways and cannot be backfilled; and (4) the
mining methods used in breccia-pipe mines preclude the use of
backfilling.
Theoretically, a positive mine pressure will force the radon
in mine air through the ore body or surrounding area to the
surface and, if conditions are right, the radon will decay before
reaching the surface (Ko80, FrSla). However, this practice will
not be applicable at all mines, as it is critically dependent on
the surrounding geology. An "air" sink is required to accept the
the surrounding geology.
11-27
-------
radon, and if the rock surrounding the mine is impermeable the
™??™?CGnti£*i0n ln the mine air wil1 ^ickly "turn to'previ-
ous levels. This process has shown limited success in reducina
radon concentrations in a mine atmosphere, but the reduction in
mine emissions was not determined nor have costs for the process
been estimated (Praia). After a thorough review of this pr°CeSS
technology, the Bureau of Mines concluded that a positive
pressure condition is ineffective in reducing radon emissions
from underground uranium mines (B184).
Experiments using strong oxidizing agents to convert radon
to a chemical form that can be absorbed on scrubbers or absorp-
tion beds have been performed (FrSla). However, the corrosive
and toxic nature of these reactants makes their use in mines
impractical and, most likely, unacceptable. Other techniques
such as cryogenic condensation, gas centrifugation, molecular
sieves, and semipermeable membranes have been reviewed as possi-
ble techniques for reducing radon emissions from underground
mines, but were found to be impractical and too costly (Ho84,
0X04 J •
11-4.6 New Underground Mines
The control of radon emissions from mature underground
uranium mines has been only marginally successful, and supplemen-
tary control technologies, as seen above, have not significantly
reduced radon emissions from these mines. The manner in which
these mines were developed and are operated is not optimal for
radon.control. Although it is not likely that new mines will be
starting in appreciable numbers, a positive change in the present
depressed condition of the industry could initiate new mine
development. If this should occur, new mines can be developed
and operated in a way that would minimize, without undue burden
the emission of radon to the atmosphere. '
Extensive pre-operational planning is imperative in order to
minimize radon emissions from new underground mines. Planning is
necessary to insure adequate access to the ore and the achieve-
ment of an efficient arrangement of openings for optimal ventila-
tion distribution simultaneously with a minimal release of radon
into the mine atmosphere. The life cycle of a mine can be divid-
ed into five stages: exploration, construction, underground
development, ore extraction, and abandonment. Procedures to
minimize radon emissions should be considered during each mining
stage. Preplanning should also consider using retreat mining
wherever possible with breccia-pipe, roll-blanket, roll-front,
and vein-type uranium deposits.
11.4.7 Conclusions
Considerable effort has been made to find technologies that
would effectively control the emissions of radon from underground
uranium mines. Numerous alternatives have been reviewed and
tested, but none appear to meet the conditions necessary to
11-28
-------
achieve adequate radon emission reductions. Bulkheads have been
partially successful but cannot be used to reduce radon emissions
further. Extending the height of mine ventilation exhaust
stacks, however, does effectively reduce the exposure and risk to
nearby individuals. Health risks resulting from radon emissions
can be most effectively controlled at future mines by following a
carefully planned program in the development and operation of the
mine.
11-29
-------
11.5 REFERENCES
BeSl
B184
At74 Athey, T.W.; Tell, R.A.; and Janes, D.E., "The Use of an
Automated Data Base in Population Exposure Calculations,"
from Population Exposuresr Health Physics Society
CONF-74018, October 1974.
Begovich, C.L.; Eckerman, K.F.; Schlatter, E.G.; Ohr,
S.Y.; and Chester, R.O., "DARTAB: A Program to Combine
Airborne Radionuclide Environmental Exposure Data with
Dosimetric and Health Effects Data to Generate
Tabulations of Predicted Health Impacts," ORNL-5692, Oak
Ridge National Laboratory, Oak Ridge, TN, August 1981.
Bloomster, C.H.; Jackson, P.O.; Dirks, J.A.; and Reis,
J.W., "Radon Emissions From Underground Uranium Mines,"
Draft Report, Pacific Northwest Laboratory, 1984.
DOE83 Department of Energy, "Statistical Data of the Uranium
Industry," GJO-100(83), Grand Junction, CO, January 1983.
Dr80 Droppo, J.G.; Jackson, P.O.; 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, EPA Contract Report 80-2, EPA, Office of
Radiation Programs, Washington, DC, August 1980.
Droppo, J.G., "Modeled Atmospheric Radon Concentrations
From Uranium Mines," Draft Report, Pacific Northwest
Laboratory, PNL-52-39, September 1984.
Environmental Protection Agency, "Regulatory Impact
Analysis of Final Environmental Standards for Uranium
Mill Tailings at Active Sites," EPA 520/1-83-010, Office
of Radiation Programs, Washington, DC, September 1983.
EPA83b Environmental Protection Agency, "Potential Health and
Environmental Hazards of Uranium Mine Wastes," EPA
520/1-83-007, Office of Radiation Programs, Washington,
DC, June 1983.
EPA85 Environmental Protection Agency, "Background Information
Document - Standard for Radon-222 Emissions from
Underground Uranium Mines, " EPA 520/1-85-010, Office of
Radiation Programs, Washington, DC, April 1985.
FrSla 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, CO, October 1981.
Dr84
EPA83a
11-30
-------
FrSlb Franklin, J.C. and Weyerstad, K.D., "Radiation Hazards in
Backfilling with Classified Uranium Mill Tailings,"
Proceedings of the Fifth Annual Uranium Seminar,
Albuquerque, NM, September 20-23, 1981.
Ho84 Hopke, P.K.; Leong, K.H.; and Stukel, J.J., "Mechanisms
for the Removal of Radon from Waste Gas Streams," EPA
Cooperative Agreement CR806819, UILU-ENG-84-0106,
Advanced Environmental Control Technology Research
Center, Urbana, IL, March 1984.
JaSO Jackson, P.O.; Glissmeyer, J.A.; Enderlin, W.I.;
Schwendiman, 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, WA, February 1980.
Jo89 Jones, R.K., Environmental Coordinator, UMETCO Minerals
Corporation, Grand Junction, CO, Comments on Proposed
Radionuclide NESHAPS Standards, Docket No. A-79-11, to
Central Docket Section (A-130), U.S. Environmental
Protection Agency, Washington, DC, May 12, 1989.
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, DC, April 1980.
Mo79 Moore, R.E.; Baes, C.F. Ill; McDowell-Boyer, L.M.;
Watson, A.P.; Hoffman-, F.O.; Pleasant, J.C.; and Miller,
C.W., "AIRDOS-EPA: A Computerized Methodology for
Estimating Environmental Concentrations and Dose to Man
From Airborne Releases of Radionuclides," EPA
520/1-79-009, Oak Ridge National Laboratory for U.S. EPA,
Office of Radiation Programs, Washington, DC, December
1979.
Pi88a Pierce, P.E., Senior Mining Engineer, Grants, NM, written
communication, August 1988.
Pi88b Pierce, P.E., Senior Mining Engineer, Grants, NM, written
communication to R.L. Blanchard, SC&A, Inc., Montgomery,
AL, November 28, 1988.
Pi88c Pierce, P.E., Senior Mining Engineer, Grants, NM, written
communication to D.J. Goldin, SC&A, Inc., McLean, VA,
November 1988.
Pi89 Pierce, P.E., Senior Mining Engineer, Grants, NM, written
communication, May 1989.
11-31
-------
Sa89 Sampson, G., UMETCO Minerals Corporation, Grand Junction,
CO, written communication to Wayne Dolezal, Grants, NM,
May 8, 1989.
Tr79 Travis, C.C.; Cotter, S.J.; Watson, A.P.; Randolph, M.L.;
McDowell-Boyer, L.M.; and Fields, D.E., "A Radiological
Assessment of Radon-222 Released From Uranium Mills and
Other Natural and Technologically Enhanced Sources,"
Prepared by the Health and Safety Research Division, Oak
Ridge National Laboratory for U.S. Nuclear Regulatory
Commission, NUREG/CR-0573, 1979.
11-32
-------
APPENDIX 11-A
Basis of the Cost Estimates for
Exhaust Stack Modifications at Existing
Underground Uranium Mines
11-A-l
-------
The following explains the basis for estimating the costs to
extend the mine air exhaust stacks at the 15 existing underground
uranium mines presented in Section 11.4.4.
A typical mine ventilation exhaust stack will include steel
plate ducting from a vertical shaft to a large, high pressure
fan. Discharge from the fan will pass through a flared duct (an
evase) before release to the atmosphere. The cost estimate has
been prepared for straight line ducting mounted vertically with-
out an evase, transition pieces, or equipment including fans and
airdoors.
Actual steel members were considered as the diameter in-
creased from 4 feet to 24 feet and the height increased from
33 feet (10 meters) to 200 feet (60 meters). Such a distinction
was made so that bogus costs were not generated on structures
that could not possibly be built and utilized in the mining
operation. In the structural calculations, a minimum safety
factor of 8.0 was used. The vertical stack is a tower structure
composed of liner, posts, and cross-braces all integrated into
one unit. The slenderness ratio of classic structural design is
applicable where the length (height) divided by the base dimen-
sion shall not be greater than 50 and the length divided by the
radius of gyration shall not exceed 120.
The first component considered was the stack lining. A
single body of steel plate was considered. Material weight per
vertical foot of stack lining was determined as shown in
Table 11-A-l.
The liner plate weights used are given in Table ll-A-2 for
heights of 30 feet (10 meters), 70 feet (20 meters), 100 feet
(30 meters), and 200 feet (60 meters). Thicknesses of 1/4-inch for
4-foot diameter, 1/4-inch for 6-foot diameter, 3/8-inch for
8-foot diameter, and 1/2-inch for 24-foot diameter stacks were
selected.
Only primary steel members were considered for each struc-
ture. Posts and cross-braces were commonly sized. Secondary
members and connectors should be included considering the degree
of conservatism used in the calculations. All steel weights are
included in Table ll-A-2.
Concrete foundations were included. The quantities increased
as stack liner diameters increased. Concrete, regardless of
stack height, included 4 cubic yards for a 4-foot diameter stack,
5 cubic yards for a 6-foot diameter stack, 12 cubic yards for an
8-foot diameter stack, and 50 cubic yards for a 24-foot diameter
stack.
ll-A-2
-------
Table 11-A-l. Weights of stack liner per vertical foot.
Liner Thickness
Stack Diameter 1/4-Inch 3/8-Inch
4'
6'
8'
24'
Table ll-A-2.
Stack
Height Steel
Meters Member
10
20
30
60
10
20
30
60
10
20
30
60
10
20
30
60
6WF20
6WF20
8WF35
8WF35
6WF20
6WF20
8WF35
8WF35
6WF20
6WF20
8WF35
8WF35
8WF35
8WF35
1OWF49
1OWF49
128.3
192.3
256.3
769.1
Ibs 192.3 Ibs
288.4
384.5
1153.6
Weights of structural steel
Support Steel
Brace Spacing Length Weight
(Feet) (Feet) (Ibs)
4 -foot
10
10
10
5
6-foot
10
10
10
5
8 -foot
10
5
5
5
2 4 -foot
10
5
10
5
Diameter;
180
420
600
1200
Diameter ;
204
476
680
1360
Diameter;
240
840
1200
2400
Diameter;
480
1960
1600
5600
1/4
3
8
21
42
1/4
4
9
23
47
3/8
4
16
42
84
1/2
16
68
78
274
Inch
,600
,400
,000
,000
Inch
,080
,520
,800
,600
Inch
,800
,800
,000
,000
Inch
,800
,600
,400
,400
I/ 2 -Inch
5/8-Inch
256.4 Ibs 320.5 Ibs
384.5 480.7
512.7 640.8
1538.2 1922.7
used.
Casing Total
Weight Weight
(Ibs) (Ibs)
Thick
3,
8,
12,
25,
Thick
5,
13,
19,
38,
Thick
11,
26,
38,
76,
Thick
46,
107,
153,
307,
846
974
820
640
769
461
230
460
535
915
450
900
146
674
820
640
7,
17,
33,
67,
9,
22,
43,
86,
16,
43,
80,
160,
62 ,
176,
232,
582,
,446
,374
,820
,640
,849
,981
,030
,060
,335
,715
,450
,900
,946
,274
,220
,040
ll-A-3
-------
Costs were based on actual past quotations and escalated to
current values as per U.S. Bureau of Labor Statistics, Consumer
Price Index, and other sources. A composite cost of $1.80 per
pound was used for structural steel finished, fabricated, and
erected. A concrete cost of $150 per cubic yard was used for
delivery and placement. Total costs per individual stack are
shown in Table ll-A-3.
Each underground uranium mine has a different set of operat-
ing and ventilating conditions. Thus, the exhaust ports from
each mine were constructed to meet these localized conditions.
The number and size of each exhaust shaft included in the cost
estimate are shown in Table ll-A-4.
Table ll-A-3. Exhaust stack costs (dollars) for individual
stacks.
Stack
Diameter
4 ft
6 ft
8 ft
24 ft
Cost
Component
Steel
Concrete
Total
Steel
Concrete
Total
Steel
Concrete
Total
Steel
Concrete
10
13,400
600
14,000
17,700
750
18,500
29,400
1,800
31,200
113,300
7,500
Stack Hed
20
31,300
600
31,900
41,400
750
42,200
78,700
1,800
80,500
317,300
7,500
dht . meters
30
60,900
600
61,500
77,500
750
78,300
144,800
1,800
146,600
418,000
7,500
60
121,800
600
122,400
154,900
750
155,700
289,600
1,800
291,400
1,047,700
7,500
Total 120,800 324,800 425,500 1,055,200
Note - Costs of the 7-foot diameter stacks were estimated by the
ratio of costs for 6 and 8-foot diameter stacks.
ll-A-4
-------
Table ll-A-4. Number and size of exhaust shafts assumed for cost
estimate.
Mine
Section 23
Schwartzwalder
Pigeon
Pinenut
Kanab North
Mt. Taylor
Sheep Mountain No. 1
King Solomon
Sunday
Deremo-Snyder
Wilson-Silverbell
Calliham
Nil
La Sal
Snowball -Pandora
No. of Vents
4
9
3
1
1
1
1
5
13
12
11
7
1
3
5
4
Diameter
(feet)
4
6
8
*
8
8
8
24
4 (Avg)
8
8
8
8
8
6
7
7
ll-A-5
-------
-------
12. SURFACE URANIUM MINES
12.1 GENERAL DESCRIPTION
Uranium is a silvery-white, radioactive metal that is used
as fuel in nuclear reactors and as a constituent of nuclear
weapons. The uranium is removed from the ore by milling and may
be enriched in the uranium-235 isotope prior to use. The
background concentration of uranium in the earth's crust is
approximately 2 parts per million; it occurs in many rocks as a
minor constituent. In the United States, most of the uranium
resources occur in sandstone host rocks, including coarse and
fine-grained clastic materials.
In surface mining, the topsoil and overburden are excavated
or stripped to expose the uranium ore. Topsoil may be segregated
and saved for reclamation; overburden is piled on unmineralized
land beside the exavation or pit. Low-grade ore encountered in
the stripping may be saved for blending with higher grade ore or
for subsequent heap leaching. It may also be segregated for
later burial or mixed with waste rock and serve in part as the
earthen cover for reclamation. Typically, the pits and
overburden or waste piles will cover over 100 acres each; the
pits, waste piles, and haul roads of a major open-pit mining
operation may cover over 1,000 acres.
Initial excavation may uncover most or all of an ore body,
or mining may progress in phases along the ore zone; this is an
economic consideration determined largely by the size, shape,
depth, and characteristics of the ore zone. Where the stripping
is done in phases, overburden from the subsequent cuts is
backfilled into the earlier mined area, each area being reclaimed
as the mining progresses along the ore zone until the final cut
is completed. When mining is completed, the final cut may be
backfilled, remaining highwalls reduced, waste piles sloped and
graded, topsoil replaced, and the area revegetated. The extent
and success of these efforts depends on applicable regulations or
lease requirements.
12.1.1 Surface Mine Production
Annual uranium ore production from surface mines in the
United States from 1948 through 1986 is presented in Table 12-1.
The data show the cyclical nature of the industry. Production
trends pointed upward during the 1950s and early 1960s, reaching
a peak of about 2.5 million tons in 1961. During the remainder
of the 1960s, production never exceeded the 1961 peak, averaging
only about 1.7 million tons per year. In 1971, production
increased sharply, starting an upward trend that would continue
until the peak of 1980 when more than 10 million tons were
produced. Since 1980, the trend has been sharply downward,
falling to less than 2 million tons in 1984, and below a million
tons by 1986. Since the peak production year of 1980, the number
of active surface mines has declined from 167 to 2.
12-1
-------
Table 12-1. Uranium ore production from surface mines, 1948-1986,
Year Thousand Tons
of Ore
1948
1949
1950
1951
1952
1953
1954
1955
1956
1957
1958
1959
1960
1961
1962
1963
1964
1965
1966
1967
1968
1969
1970
1971
1972
1973
1974
1975
1976
1977
1978
1979
1980
1981
1982
1983
1984
1985
1986
<1
1
23
28
65
179
266
374
1,247
1,613
2,358
2,206
2,393
2,482
1,782
1,879
1,537
1,243
1,333
1,593
2,366
2,173
2,801
3,284
3,887
4 ,.544
4,216
4,247
4,673
5,578
8,237
9,655
10,394
8,436
5,504
(a)
1,968
936
(a)
(a) Data not available.
12-2
-------
Some of the recently idled open-pit mines are being held on
a standby status; the operators hoping for a recovery in the
uranium market. Others, nearing the end of their economic
reserves as the market slumped, have been closed permanently and
either reclaimed or abandoned.
Much of today's uranium production is from underground mines
and alternative sources; this trend is expected to continue for
the foreseeable future. It is expected that present trends
will continue at least through 1995, with uranium mining
concentrated in a dozen or so medium to large underground mines
and a few open-pit mines. Factors that could alter this include
legislative supports favoring the domestic uranium industry or
changes in international conditions, such as a repeat of the
energy crisis of the mid-1970s, leading to renewed interest in
nuclear power generation.
Historically, the principal states in which uranium ores
have been mined are Arizona, Colorado, New Mexico, Texas, Utah,
Washington, and Wyoming; lesser amounts have been produced in
California, Idaho, Montana, Nebraska, Nevada, North Dakota,
Oregon, and South Dakota (DOES6).
Over 1,300 surface uranium mines have been identified in the
United States (EPA83). Of this total, over 1,000 have been
identified as having uranium production under 1,000 tons. These
small mines typically have surface areas ranging from several
hundred to several thousand square feet. The remainder of the
mines, categorized by 1,000 - 100,000 and > 100,000 tons uranium
ore production, are summarized by location in Table 12-2.
Table 12-2. Breakdown by state of surface uranium mines with
> 1,000 tons production.
State 1,000 - 100,000 Tons Greater than 100,000 Tons
Arizona
California
Colorado
Idaho
Montana
Nevada
New Mexico
North Dakota
Oregon
South Dakota
Texas
Utah
Washington
Wyoming
37
l
12
1
1
1
3
10
1
33
19
6
3
66
1
0
4
0
0
0
5
0
1
2
25
0 "
2
31
12-3
-------
The larger production mines typically have features such as
overburden, topsoil, and low grade mineralization (ore)
associated with the actual pit surfaces. All of these features
contribute to radon and particulate emissions, with intensity
determined by uranium content and size.
The 265 mines identified in Table 12-2 accounted for over 99
percent of all surface uranium ore production and, subsequently,
particulate and radon emissions. Of the 265 mines listed, 2 are
actively producing uranium ore; these are the Chevron Resource
Company's Rhode Ranch mine, approximately 110 miles due south of
San Antonio, Texas, and the Pathfinder Mine's Shirley Basin mine
in Carbon County, Wyoming. The remaining 263 mines are closed
and in varying states of reclamation.
12.1.2 Standards and Regulations Applicable to Surface Uranium
Mining
Health, safety, and environmental hazards associated with
uranium mining are regulated by a variety of Federal and state
laws. Passage of the National Environmental Policy Act at the
beginning of 1970 marked the onset of the public's new
environmental awareness; subsequently, especially through the
1970s, there was a rapid succession of increasingly strict
environmental laws affecting mining activities. These laws were
passed at both the Federal and state level.
12.1.2.1 Federal Regulations
Federal laws and regulations applicable, at least in part,
to uranium mining include the Clean Air Act, the Federal Water
Pollution Control Act of 1948, the Safe Drinking Water Act, the
Solid Waste Disposal Act, and the Resource Conservation and
Recovery Act. These provide basic requirements for environmental
protection and require the EPA to establish standards and
guidelines under which the states may issue permits and enforce
the laws. States may establish stricter or more detailed
standards, but their regulations generally parallel those of the
EPA.
Another law that has indirectly affected the surface uranium
mining industry is the Surface Mining Control and Reclamation Act
of 1977. Although this act applies only to coal mining
operations, the environmental and reclamation requirements that
it established have served as models for many western states in
regulating non-coal surface mining operations.
Table 12-3 gives an overview of Federal laws, regulations,
and guidelines applicable to surface uranium mining.
12-4
-------
Table 12-3. Federal laws, regulations, and guidelines for uranium mining.
Environmental Standards
to
l
ui
Department or Agency
Permits and Approvals
Prospecting Mining Reclamation Air Surface Ground Solid Public Health
Quality Water Water Waste and Safety
Department of the Interior:
Bureau of Land Management
Bureau of Indian Affairs
National Park Service
Fish and Wildlife Service
Bureau of Reclamation
Department of Agriculture:
Forest Service
Department of Energy
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
Environmental Protection Agency
Army Corps of Engineers
Department of Labor:
Mine Safety & Health
Administrat ion
Occupational Health &
Safety Administration
X X
X
X
X
Nuclear Regulatory Commission
-------
As shown in Table 12-4, a significant percentage of uranium
resources are on Federal and Indian lands, 23 percent and 7
percent respectively. Federal laws and regulations govern
uranium exploration and mining on these lands. The specific laws
and regulations applying to a particular operation depend on the
land category, but in all cases, some degree of review and
approval is required before any significant surface mining
operations can be undertaken. For permitting requirements,
operations on these lands fall into two broad categories: leased
lands and mining claim locations. Lands subject to leasing
include Indian lands (leased from the tribe with concurrence of
the Secretary of the Interior), acquired lands, and withdrawn
lands. The public domain lands, unless otherwise reserved, are
open to mining claim locations.
Table 12-4. Estimated additional uranium resources by land
status.(a)
Land Status Million Pounds U3O8 Percent
Federal lands
Public lands - BLM, FS 540 22.7
Other 120 5^2
Indian lands 170 7.1
State lands 80 3.2
Private fee lands 1,460 61!8
Totals 2,370 100.0
(a)Adapted from DOE86, based on $50/lb forward cost.
Federal regulation and supervision are particularly
significant in the western uranium-producing states, several of
which have large percentages of federally owned lands. These
include Arizona (43 percent), California (45 percent), Colorado
(36 percent), Idaho (64 percent), Montana (30 percent), Nevada
(87 percent), New Mexico (34 percent), Utah (66 percent),
Washington (29 percent), and Wyoming (48 percent). Most of these
states also have environmental requirements for mining
operations; an operator on Federal or Indian lands will normally
be subject to whichever requirements are the more stringent. In
addition, any Federal permits or approvals are subject to the
National Environmental Policy Act, which requires an
environmental review of the proposed operation prior to Federal
approval.
On lands subject to leasing, environmental reviews and
approvals are necessary at the prospecting, exploration, and
mining stages. This leasing function is carried out on most
12-6
-------
Federal lands by the Bureau of Land Management (BLM) in
consultation with the appropriate surface management agency.
On Indian lands, the leasing function is split; the mineral
lease is developed by the tribe with the Bureau of Indian Affairs
carrying out the responsibilities of the Secretary of the
Interior, while the BLM supervises operations under the lease.
Environmental requirements in both Federal and Indian leases
range from the mere statement in older leases that the lessee
shall comply with all appropriate Federal, state, and local
standards, to the current practice of including additional
specific standards and requiring monitoring and reporting to
document compliance. Likewise, reclamation requirements have
evolved from the requirement in older leases that the land be
reclaimed to the satisfaction of the Secretary of the Interior to
specific reclamation .plans being required as part of the approval
process. An outstanding example of what the Department of
Interior may require in the way of reclamation of a surface
uranium mine is the recently approved plan for the Jackpile-
Paguate mine on the Laguna reservation in New Mexico.
On lands subject to mining claim locations, environmental
review and approval of a plan of operations are required on land
managed by the BLM for any operations where the annual surface
disturbance will exceed 5 acres and for any surface operation in
environmentally sensitive areas.
Mining operations on public domain lands in the National
Forest System are managed by the Forest Service (FS), an agency
of the Department of Agriculture. FS approval is required for
activities that could result in significant surface disturbance.
12.1.2.2 State Regulations
Uranium mining on private and state-owned lands is subject
to regulation by the particular state and, in some instances, the
local governments. Most of the western states that have
significant uranium mining have enacted some degree of
environmental and surface protection legislation in recent years.
Laws, regulations, and guidelines applicable to uranium mining in
Arizona, Colorado, New Mexico, South Dakota, Texas, Utah, and
Wyoming are summarized below.
12.1.2.2.1 Colorado
Colorado is an NRC Agreement State and has been authorized
by the EPA to issue NPDES discharge permits. Both radiation and
water quality regulatory activities are under the jurisdiction of
the Colorado Department of Health. National ambient air quality
standards and various state emission control regulations apply to
uranium mining activities.
12-7
-------
Prospecting permits and mining leases for state-owned lands
are issued by the Board of Land Commissioners, affiliated with
the Colorado Department of Natural Resources. The Board has
policies and regulations concerning environmental impacts from
mining activities on state lands.
The Colorado Mined Land Reclamation Board, created in 1976
and administered by the Department of Natural Resources, issues
permits for all mining operations on all lands in the state, both
Federal and non-Federal, under the Colorado Mined Land
Reclamation Law.
12.1.2.2.2 New Mexico
In New Mexico, a mine plan must be filed with and approved
by the State Mining Inspector. However, the emphasis of the
review is on worker and mine safety rather than environmental
impacts. There are, at present, no state regulations governing
solid wastes and land reclamation for mining operations. The
plan and bonding requirements for the mining permit determine the
extent of any waste control and land reclamation.
Prospecting permits and mining leases for state-owned lands
are issued by the State Land Commissioners, who have policies and
regulations concerning environmental impacts from mining
activities on the state lands.
12.1.2.2.3 Texas
Uranium prospecting and mining activities in Texas are
regulated under the Texas Uranium Surface Mining and Reclamation
Act, administered by the Texas Railroad Commission on all lands
except those owned by the state. The regulations establish
environmental and reclamation standards, provide for review and
approval of mining plans, and require monitoring and bonding
sufficient to ensure compliance.
Prospecting permits and mining leases on state-owned lands
are issued by the General Land Office (GLO). Mining and
reclamation requirements are similar to those for the non-state
lands but are enforced by the GLO.
The Texas Guides and Regulations for Control of Radiation
apply to in-situ uranium mining (under NRC Agreement State
licensing) but not to surface uranium mining.
12.1.2.2.4 Utah
Uranium mining in Utah is regulated under the Utah Mined
Land Reclamation Act, by the Division of Oil, Gas, and Mining of
the Department of Natural Resources. A mining and reclamation
plan and bonding are required. Standards are promulgated for
environmental considerations as well as public health and safety
concerns. Reclamation requirements include regrading of slopes,
12-8
-------
burial of mineralized materials, and applying topsoil cover
sufficient to sustain adequate revegetation. Mining activities
on state-owned lands require a lease and approval of a plan of
operations from the Division of State Lands and Forestry of the
Department of Natural Resources.
12.1.2.2.5 Wyoming
Uranium mining in Wyoming is regulated under the Wyoming
Environmental Quality Act by the Land Quality Division of the
Wyoming Department of Environmental Quality. Regulations require
mining and reclamation plans, establish environmental standards,
and provide for monitoring and bonding to ensure compliance.
Mined land must be restored to a use at least equal to its
highest previous use. The state has established standards for
residual radioactivity on lands mined for uranium. Procedures
for proper handling of sub-ore and mineralized wastes are also
specified. An Air Quality Permit is required for construction of
a uranium mining and/or processing facility; compliance with
applicable ambient air quality standards and prevention of
significant deterioration provisions must be demonstrated.
12.1.2.2.6 Arizona
Arizona is an NRC Agreement State. There are no additional
state-imposed legislative or regulatory requirements concerning
exploration or prospecting permits, mining plans, or surface
reclamation.
12.1.2.2.7 South Dakota
South Dakota has established a Division of Land and Water
Quality within the Department of Water and Natural Resources.
Within this Division, the Exploration and Mining Program Office
is responsible for administering the Mined Land Reclamation Act.
The Act and implementing regulations require exploration permits,
prospecting permits, and mining plans. The mining plans must
include appropriate measures for reclamation.
12.1.2.3 State Reclamation Status
Reclamation status of mines within various mining districts
varies greatly based on the individual state permitting
regulations at the time the mine was operated. In most states
with stringent permitting and reclamation requirements, a
significant percentage of the mines have been reclaimed or are
undergoing reclamation.
Two primary reclamation techniques were noted during field
studies summarized in "Inactive Surface Uranium Mine Radon and
Particulate Emissions" (SCA89). The first method consists of
total backfill of the excavated pit, with waste material returned
in the sequence it was removed. The site is then regraded to
original contours and revegetated. The second, and most
12-9
-------
prevalent type of reclamation, consists of grading the waste
piles and pit wall to a 3:1 or 4:1 slope, with subsequent
topsoiling and revegetation. Table 12-5 summarizes the estimated
percentage of mines in each reclamation class for larger ore-
producing states (SCA89).
As shown in Table 12-5, the majority of the surface mines in
most states have had no reclamation or emissions controls
implemented. Leaseholders have typically left the mining areas
in a condition to comply with any regulatory requirements, which
in most cases, were quite limited. Therefore, many of the
original landowners had property returned in totally unreclaimed
condition with no financing available to repair the land. This
problem is prevalent in Arizona, South Dakota, and Nevada.
No existing controls for radon or particulate emissions from
inactive surface uranium mines have been specifically implemented
by any mine operator or regulatory agency for the sole reason of
lowering these emissions. However, reclamation of these mines
for other reasons, such as legal requirements, aesthetics, or
corporate policy leads to lower radiological emissions in most
cases.
12.1.2.3.1 Arizona
All mines are located on Navajo Indian land and are
unreclaimed and abandoned. As no reclamation requirements are or
were imposed on mining companies, the status of reclamation is
not expected to change.
12.1.2.3.2 Colorado
Some very minor reclamation in the form of sloping pit and
waste piles has been performed at the sites. However, the
reclamation did not include covering of waste piles or pit
surfaces. Thus, particulate and radon emissions have not been
reduced. Since no state reclamation requirements were imposed,
it is anticipated that the mines will remain unreclaimed.
12.1.2.3.3 South Dakota
No state reclamation requirements were in effect during the
time the mining activities were carried out. All mines are
unreclaimed and abandoned.
12.1.2.3.4 Texas
Approximately two-thirds of the surface uranium mines in
Texas have been or will be reclaimed by local mining companies
under regulations enforced by the Texas Railroad Commission.
Most of these mines required reclamation because they were
permitted by the State of Texas after the Surface Mining Act of
1975.
12-10
-------
Table 12-5. Estimated status
of surface uranium mine reclamation.
State
Total Ore Production
1,000 - 100,000 tons
Class I Class II Unreclaimed
(X) (X) <*>
Total Ore Production
> 100,000 tons
Class I Class II Unreclaimed
(X) (X) (X)
Arizona
Colorado
New Mexico
North Dakota
South Dakota
Texas
Utah
Washington
Wyoming
0
5
0
0
0
10
0
0
5
5
20
15
5
5
45
0
50
40
95
75
85
95
95
45
100
50
55
0
5
0
-
0
10
-
0
5
0
20
15
-
5
45
-
50
40
100
75
85
-
95
45
-
50
55
(a)Status defined as:
Class I - total backfill, recontourlng, and revegetatlon;
Class II - resloping of waste piles and pits, topsolllng, and revegetation;
Unreclaimed - property abandoned without restoration.
Mining companies in the region use two primary forms of
reclamation. One method entails a total backfill in which
material is returned to the pit in the sequence it was removed,
and land surfaces are brought back to as near original contours
as possible. The other method consists of sloping, topsoiling,
and revegetation of waste piles and pit walls, with subsequent
formation of a holding pond of acceptable water quality.
12.1.2.3.5 Utah
Surface mines in Utah are abandoned and unreclaimed.
status is not expected to change.
12.1.2.3.6 Wyoming
This
Mining areas in Wyoming include the Powder River Basin, the
Gas Hills, and the Shirley Basin. There are no active mines in
the Powder River Basin. Most mines in this area have been
reclaimed by sloping, topsoiling, and revegetation. In the Gas
Hills and Shirley Basin regions, the general mining practice was
12-11
-------
to place the wastes from active pits into inactive pits.
Reclamation in these areas is ongoing, with many mines reclaimed,
and others being reclaimed. The state is currently sponsoring
reclamation of some of the older Shirley Basin mines.
12.2 BASIS OF THE DOSE AND RISK ASSESSMENT
The assessment of the doses and risks posed by emissions of
radon-222 and radionuclides released in particulate form from
surface uranium mines is based upon site-specific evaluations of
the 2 active mines and 25 large inactive mines. The
characteristics of these mines are given in Table 12-6. Large
mines (total ore production > 1,000 tons) were selected for
evaluation since they account for more than 99 percent of the
total ore produced, and hence radionuclide emissions. The mines
selected are located in six different states: Arizona, New
Mexico, Colorado, South Dakota, Texas, and Wyoming. The results
obtained from this representative group of mines are extrapolated
to obtain an estimate of the doses and risks posed by all surface
uranium mines.
12.2.1 Radionuclide Source Terms
The source terms for surface uranium mines were developed
from site characterizations and radiological data collected
during site visits and field studies (Pi88, PNL82, SCA89).
Measured radon flux rates were developed for one mine within each
state (SCA89). For the other mines, the radon source terms are
estimated by correlating the appropriate flux data with measured
gamma exposure rates obtained by site surveys. The radon-222
emissions are given in Table 12-7. Particulate source terms are
estimated on the basis of measured radium-226 concentrations,
site-specific dusting factors, and the assumption that all
members of the uranium-238 decay series are in secular
equilibrium. The uranium source terms are shown in Table 12-8.
12.2.2 Other Parameters Used in the Assessment
Site-specific demographic data were developed for the 0-5 km
areas around each of the mines during site visits (SCA89). These
were used in conjunction with meteorological data obtained from
the nearest weather station. Details of the parameters supplied
as input to the assessment codes are presented in Appendix A.
12.3 RESULTS OF THE DOSE AND RISK ASSESSMENT
The outputs of the assessment codes used to evaluate the
doses and risks of fatal cancers caused by radon-222 and
radioactive particulate emissions from surface uranium mines
include the following:
1. working level exposure and the lifetime fatal cancer
risk to the most exposed individuals from radon-222 at
each surface mine;
12-12
-------
Table 12-6. Mines characterized in the field studies.
Geologic Region Mine
Size
(tons ore)
Reclamation
Status
Inactive Mines
Texas
Arizona-
New Mezico
Wyoming
South Dakota
Colorado
Kopplin
Manfca
Stoeltje
Wright-McCrady
Swientek
Ramco 20, 22
Jack Daniels #1
Jack Huskon #3
Evans Huskon #35
Ramco #21 East
Yazzie #2
>100,000
1,000 - 100,000
>100.000
>100,000
>100,000
>100,000
1,000 - 100,000
1,000 - 100,000
1,000 - 100,000
1,000 - 100,000
1,000 - 100.000
Morton Ranch #1704 >100,000
Lucky Me 70-1, 7E >100,000
Lucky Me 4X, 4P >100,000
Lucky Me V. Gas Hills >100,000
Texas
Wyoming
Darrow #1
Darrow #2,
Darrow #4
Darrow #5
Freezout
Gert #4-7
Johnson
Sage
Marge #1-3
Rob
Rhode Ranch
Shirley Basin
1.000 - 100,000
>100,000
1,000 - 100,000
>100,000
1,000 - 100,000
>100,000
1,000 - 100,000
1,000 - 100,000
1,000 - 100,000
>100,000
Active Mines
>100,000
>100,000
unreclaimed
unreclaimed
minor reclamation
unreclaimed
fully reclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
fully reclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
unreclaimed
continuous backfill
operating
12-13
-------
Table 12-7. Estimated radon-222 emissions from surface uranium
mines.
Geologic Region Mine
Radon-222
Gross Net*
(Ci/y)
Texas
Arizona-
New Mexico
Wyoming
South Dakota
Colorado
* Background
Kopplin
Manka
Stoeltje
Wright-McCrady
Swientek
Rhode Ranch
Ramco 20, 22
Jack Daniels #1
Jack Huskon #3
Evans Huskon #35
Ramco #21 East
Yazzie #2
Morton Ranch #1704
Lucky Me 70-1, 7E
Lucky Me 4X, 4P
Lucky Me W. Gas Hills
Shirley Basin
Darrow #1
Darrow #2, 3
Darrow #4
Darrow #5
Freezout
Gert #4-7
Johnson
Sage
Marge #1-3
Rob
12
19
10
80
11
—
47
16
17
<1
7.0
7.0
120
420
300
190
—
8.0
18
9.0
43
19
530
81
270
190
630
12
15
7.2
68
2.0
40
44
14
18
<1
5.7
6.5
110
370
270
150
920
5.4
12
5.9
32
17
480
52
240
170
600
radon considered as appropriate.
12-14
-------
Table 12-8. Estimated particulate emissions from surface uranium
mines.
Geologic Region Mine
Uranium-238(a)
(Ci/y)
Texas
Arizona-
New Mexico
Wyoming
South Dakota
Colorado
Kopplin
Manka
Stoeltje
Wright-McCrady
Swientek
Rhode Ranch
Ramco 20, 22
Jack Daniels #1
Jack Huskon #3
Evans Huskon #35
Ramco #21 East
Yazzie #2
Morton Ranch #1704
Lucky Me 70-1, 7E
Lucky Me 4X, 4P
Lucky Me W. Gas Hills
Shirley Basin
Darrow #1
Darrow #2, 3
Darrow #4
Darrow #5
Freezout
Gert #4-7
Johnson
Sage
Marge #1-3
Rob
6.7E-4
6.6E-4
3.2E-4
4.0E-3
6.5E-4
7.4E-4
7.8E-4
3.5E-6
1.4E-4
2.1E-4
2.8E-2
1.6E-1
1.2E-1
6.4E-2
1.9E-3
4.8E-3
2.5E-3
1.1E-2
5.6E-3
4.7E-3
6.7E-4
2.9E-3
1.7E-3
7.3E-3
(a) Uranium-238 assumed to be in secular equilibrium with its
decay products.
12-15
-------
2. the number of fatal cancers committed per year in the
regional (0-80 km) populations around each surface mine
from radon-222 emissions;
3. dose equivalent rates and the lifetime fatal cancer risk
to the most exposed individuals from radioactive
particulate emissions;
4. the collective dose equivalent rates and fatal cancers
committed per year in the regional populations from
radioactive particulates; and
5. the estimated collective risk (deaths/year) and the
distribution of the fatal cancer risk among all persons
living within 80 km of surface uranium mines.
12.3.1 Radon Releases
The estimated radon exposures and the lifetime fatal cancer
risks to nearby individuals from radon-222 releases from the
study mines are summarized in Table 12-9. The estimated risks
(deaths/year) to the regional populations around these mines are
shown in Table 12-10. Estimated exposures range from 2E-7 to 4E-5
working levels for nearby individuals. The highest individual
lifetime fatal cancer risk is estimated to be 5E-5, while the
highest population risk is 1E-3 deaths/year.
Table 12-11 presents the frequency distribution of the fatal
cancer risk estimated for all surface uranium mines. This
distribution is developed by summing the individual distributions
obtained for each mine within a given region and adjusting each
regional distribution by the estimated percentage of the total
mines within the region represented by the study mines. The
regional distributions are then summed to obtain the overall
distribution presented in Table 12-11. The total number of fatal
cancers per year due to radon releases from surface uranium mines
in the regions studied is estimated to be 3E-2.
12.3.2 Particulate Emissions
The uranium-238 source terms presented in Table 12-8 were
used to evaluate the impacts of particulate releases from
inactive surface uranium mines. For each region, only the mine
sites with the largest estimated particulate releases were
evaluated.
The results of the analysis show that: organ dose rate
equivalents are below 15 mrem/y for the nearby individuals at
all sites; for the collective populations, organ dose equivalents
are below 50 person-rem/y for all sites; inhalation is the
dominant exposure pathway in all cases; thorium-230, uranium-238,
and uranium-234 are the predominant radionuclides contributing to
the doses and risks; and the organs receiving the highest dose
equivalents are the lungs, endosteum, and the red marrow (SCA89).
12-16
-------
Table 12-9. Estimated exposures and risks to individuals living near
surface uranium mines.
Region/Mine
Texas
Kopplin
Manka
Stoeltje
Wright -McCrady
Svientek
Rhode Ranch
Arizona-NM
Ramco 20, 22
Jack Daniels //I
Jack Huskon #3
Evans Huskon //35
Ramco #21 East
Yazzie #2
Wyoming
Morton Ranch #1704
Lucky Me 70-1, 7E
Lucky Me 4X, 4P
Radon
Concentration
(pCi/1)
1.2E-2
2.1E-3
1.3E-3
3.8E-3
1.7E-3
8.7E-4
2.6E-4
1.1E-3
4.1E-3
1.2E-6
3.4E-5
3.8E-5
1.2E-4
3.9E-4
2.9E-4
Lucky Me W. Gas Hills 2.0E-4
Shirley Basin
South Dakota
Darrov //I
Darrow //2, 3
Darrow #4
Darrow #5
Freezout
Colorado
Gert #4-7
Johnson
Sage
Marge //I -3
Rob
(a) Distance to the
6.9E-3
4.0E-5
8.5E-5
4.1E-5
4.4E-4
1.2E-4
3.2E-3
6.4E-4
2.9E-3
2.0E-3
1.7E-3
maximum exposed
Maximum Maximum Lifetime
Exposure Fatal Cancer Risk Distance (a'
(WL) to Individual (meters)
3.4E-5
6.3E-6
4.0E-6
1.3E-5
4.7E-6
2.9E-6
1.7E-6
3.2E-6
1.2E-5
7.7E-9
2.2E-7
2.5E-7
7.8E-7
2.5E-6
1.9E-6
1.3E-6
3.5E-5
1.7E-7
3.5E-7
1.7E-7
1.6E-6
4.8E-7
2.2E-5
4.1E-6
1.9E-5
1.3E-5
1.1E-5
individual .
5E-5
9E-6
6E-6
2E-5
7E-6
4E-6
2E-6
4E-6
2E-5
1E-8
3E-7
3E-7
1E-6
3E-6
3E-6
2E-6
5E-5
2E-7
5E-7
2E-7
2E-6
7E-7
3E-5
6E-6
3E-5
2E-5
2E-5
250
750
750
1,500
250
1,500
15,000
750
750
15,000
15,000
15,000
15,000
15,000
15,000
15,000
7,500
4,000
4,000
4,000
2,500
4,000
25,000
15,000
15,000
15,000
15,000
12-17
-------
Table 12-10.
Estimated fatal cancers per year in the regional
(0-80 km) populations due to radon-222 emissions
from surface uranium mines.
Geologic Region Mine
Fatal Cancers per Year
Texas
Arizona-
New Mexico
Wyoming
South Dakota
Colorado
Kopplin
Manka
Stoeltje
Wright-McCrady
Swientek
Rhode Ranch
Ramco 20, 22
Jack Daniels #1
Jack Huskon #3
Evans Huskon #35
Ramco #21 East
Yazzie #2
Morton Ranch #1704
Lucky Me 70-1, 7E
Lucky Me 4X, 4P
Lucky Me W. Gas Hills
Shirley Basin
Darrow #1
Darrow #2, 3
Darrow #4
Darrow #5
Freezout
Gert #4-7
Johnson
Sage
Marge #1-3
Rob
5E-4
4E-4
2E-4
1E-3
3E-5
1E-4
9E-5
3E-5
4E-5
4E-7
1E-5
1E-5
5E-5
2E-4
1E-4
7E-5
8E-5
4E-6
9E-6
4E-6
2E-5
1E-5
5E-4
6E-5
3E-4
2E-4
5E-4
12-18
-------
Table 12-11. Estimated distribution of the fatal cancer risk
caused by radon-222 emissions from all surface
uranium mines.
Risk Interval Number of Persons Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
0
4,000
200,000
30,000,000
0
0
0
0
1E-3
5E-3
2E-2
Totals 30,000,000 3E-2
Table 12-12 summarizes the lifetime fatal cancer risks to
nearby individuals and the committed fatal cancers (deaths/year)
in the regional populations from radioactive particulate
emissions for each site. No individual is estimated to have a
lifetime fatal cancer risk greater than 2E-5. The total fatal
cancers per year for all regions due to particulate emissions are
estimated to be 9E-3.
Table 12-12. Estimated lifetime fatal cancer risks from
particulate emissions.
Nearby Individuals Regional (0-80 km)
Lifetime Fatal Population
Region Cancer Risk Deaths/y
Texas 9E-8 2E-3
Arizona-New Mexico 1E-7 9E-4
Wyoming 2E-5 5E-3
South Dakota 2E-6 4E-4
Colorado 6E-6 9E-4
12.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS
Radon and particulate emissions can be controlled by
covering various areas in and around the mines with an earthen
cover. Table 12-13 shows the estimated depths of cover needed to
reduce radon emissions to 20, 6, and 2 pCi/m2/sec for various
initial flux values. The initial flux values in Table 12-12 are
based on flux levels measured over low-grade mineralized
material, overburden, and or pit surfaces of selected mines
12-19
-------
(SCA89). The estimated cover thicknesses are based on earthen
cover designs developed for uranium mill tailings piles (see
SCA89). * v
The cost to place an earthen cover over a mine to reduce
radon emissions to 20, 6, and 2 pCi/m2/sec for various initial
flux values is shown in Table 12-14. The information is based on
estimated costs to cover low grade material, overburden, and/or
pit surfaces at selected mines (SCA89).
Table 12-13. Estimated depths of cover to reduce radon-222
emissions at surface uranium mines.
Initial Flux
(pCi/m2/sec)
20 pci/nr/sec
J)epth of Cover (meters) Needed for,
i /0A^
-------
12.5 REFERENCES
DOE86 U.S. Department of Energy, Energy Information
Administration, "Statistical Data of the Uranium Industry,"
DOE/EIA-0478, 1986.
EPA83 U.S. Environmental Protection Agency, "Potential Health
and Environmental Hazards of Uranium Mine Wastes," EPA
520/1-83-007, Office of Radiation Programs, Washington,
DC, June 1983.
Pi88 Pierce, P.E., "Report of Site Visits to Operating Surface
Uranium Mines," prepared by SC&A, Inc., for the U.S.
Environmental Protection Agency, Office of Radiation
Programs, Washington, DC, August 1988.
PNL82 Pacific Northwest Laboratory, "Radon and Aerosol Release
from Open Pit Uranium Mining," PNL-4071, prepared for the
U.S. Nuclear Regulatory Commission, Office of Nuclear
Regulatory Research, NUREG/CR-2407, Washington, DC, August
1982.
SCA89 SC&A, Inc., "Radiological Monitoring at Inactive
Surface Uranium Mines," prepared for the U.S.
Environmental Protection Agency, Office of Radiation
Programs, Washington, DC, February 1989.
12-21
-------
-------
13. PHOSPHOGYPSUM STACKS
13.1 SOURCE CATEGORY DESCRIPTION
13.1.1 General Description
Phosphogypsum is the principal byproduct generated from the
wet process of producing phosphoric acid (H3PO4) from phosphate
rock. This process, conducted at about 23 facilities in the
United States, utilizes about 80 percent of the phosphate rock
produced. The states most involved in phosphate rock production
and the percentage produced in each are Florida (80 percent) ,
Idaho (7 percent), North Carolina (6 percent), Tennessee
(3 percent), Utah (2 percent), and Alabama and Wyoming (minor
amounts) . Most of the phosphoric acid resulting from this
process is used in the production of agricultural fertilizers.
In 1985, 51 million metric tons (MT) of marketable phosphate
rock were produced, of which about 41 million MT (80 percent)
were used to produce phosphoric acid by the wet process (BOMS 5) .
Since about 3.6 MT of marketable rock are required to produce
one MT of P205 (Gu75), approximately 12 million MT of P2O5 (16 mil-
lion MT of H3PO4) were produced from this rock in 1985. This
generated an estimated 52 million MT of phosphogypsum based on
4.5 MT of phosphogypsum per MT P2Os (Gu75) . (a>
The wet process for manufacturing phosphoric acid involves
four primary operations: raw material feed preparation, phosphate
rock digestion, filtration, and concentration. The phosphate
rock is generally dried in direct-fired rotary kilns, ground to a
fineness of less than 150 urn for improved reactivity, and di-
gested in a reaction vessel with sulfuric acid to produce the
product, phosphoric acid, and , the byproduct, phosphogypsum.
The phosphogypsum (gypsum) is transferred as a slurry to
onsite disposal areas referred to as phosphogypsum stacks. These
stacks are generally constructed directly on virgin or mined-out
land with little or no prior preparation of the land surface.
The gypsum slurry is pumped to the top of the stack where it
forms a small impoundment, commonly referred to as a gypsum pond.
Gypsum is dredged from the pond on top of the stack and used to
increase the height of the dike surrounding the pond. The phos-
phogypsum stacks become an integral part of the overall wet
process. Because the process requires large quantities of water,
the water impounded on the stack is used as a reservoir that
supplies and balances the water needs of the process. Thus, the
stack is not only important as a byproduct storage site, but also
contributes to the production process.
Although 75 phosphogypsum stacks were reported to exist in
the United States during 1985 (PEI85) , only 66 can be identified
today. The three inactive phosphogypsum stacks reported earlier
(a) Estimated values rounded to two significant figures.
13-1
-------
to be located in Nacogdoches County, Texas, were later identified
as scrubber water ponds at a superphosphate plant and not phos-
phogypsum stacks (Si85). Likewise, a large (203 hectare(aT)
stack in Donaldsonville, Louisiana, was incorrectly reported in
1985 as seven small stacks (Wa88a). Texasgulf Company's stacks
in Aurora, North Carolina, reported earlier as three operating
stacks, were recently identified as four idle stacks and one
operating stack (Wi88). Phosphogypsum from three stacks in
California and one stack in Oklahoma is either being sold or has
been sold for agricultural purposes, leaving little or no
phosphogypsum at the stack sites. No stack ever existed at Long
Beach, California, although it was reported earlier as an unknown
(PEI85). A stack did exist near Southgate, California, but it
has been completely removed and utilized (St88a), most likely for
agricultural purposes.
Of the 66 identifiable phosphogypsum stacks, 63 are ad-
dressed in this assessment. One stack in Alabama has an area of
only about 15 m2; thus it was not considered. The three stacks
in Idaho, identified in 1985 only as inactive and abandoned, are
actually five inactive stacks located between the towns of Kel-
logg and Smelterville in northern Idaho (Ap88). Only three of
the five stacks are of sufficient size to be considered and
included in the total of 63. Omission of the three stacks, one
in Alabama and two in Idaho, does not significantly influence the
results of the assessment.
The 63 stacks considered in this assessment are identified
in Table 13-1. The location, size, and status are given for each
stack. Phosphogypsum stacks are present in 12 different states,
with two-thirds located in just four states, Florida, Texas,
Illinios and Louisiana. Of the stacks studied, 27 are operating,
22 are inactive, and 14 are considered idle. An operating or
active stack is one that is currently receiving gypsum, and an
inactive stack is one that is permanently closed. A stack is
classified as idle if there are definite plans to reactivate it
and it has the characteristics of an active stack, e.g., water
may be maintained on the stack top surface and utilized in the
water balance for the facility. The phosphogypsum stacks range
in area from 2 to almost 300 hectares (ha), and heights of the
stacks range from 3 to about 60 meters.
A summary of phosphogypsum stacks in each state is given in
Table 13-2. The information in this table relates the phospho-
gypsum stacks to individual states and gives the distribution of
stack and stack areas within each category (operating, idle, and
inactive). The phosphate industry predominates in Florida. Over
half of the operating stacks exist in Florida, which accounts for
56 percent of the total base area of all operating stacks. The
total base area of all phosphogypsum stacks in the United States
(a) 1 hectare (ha) = 10,000 m2.
13-2
-------
United States.
Facility Name
Districhem Inc.TaJ
Agrico Chemical Oo.
Royster Phosphate, Inc. (a)
Brewster Phosphates
CF Industries, Inc.
CF Industries, Inc.
Conserv, Inc. l(b)
2
Estech, Inc.
Farmland Industries, Inc.
Gardinier, Inc.
Seminole Fertilizer 1
Corp. 2
IMC Corp.
Occidental Chemical Oo. 1
(Suwannee River) 2
Occidental Chemical Co.
(Swift Creek)
Royster Co. 1
2
USS Agri-Chemicals, Inc.
USS Agri-Chemicals, Inc.
Nu-West Industries, Inc. (a)
J.R. Simplot Co. 1
2
Bunker Hill Co. 1
2
3
Allied Chemical Co.
Beker Industries Corp.
Mobil Chemical Co.
Northern Petrochemical Co.
Olin Corp. 1
2
SECO, Inc.
U.S. Industrial
Chemicals Co.
Agrico Chemical Co. 1
2
3
Agrico Chemical Co.
Arcadian Corp. 1
2
3
4
Agrico Chemical Co. (a)
Agrico Chemical Oo. (a)
Nu-South Industries, Inc. (a)
i ^**m*u ^m^r^nm*^ ^^«»g *• ^ ^•••^ v
Location
Helena, AR
Bartow, FL
Palmetto, FL
Bradley, FL
Plant City, FL
Bartow, FL
Nichols, FL
Bartow, FL
Bartow, FL
Tampa, FL
Bartow, FL
Mulberry, FL
White Springs,FL
White Springs, FL
Mulberry, FL
Bartow, FL
Ft. Meade, FL
Conda, ID
Pocatello, ID
Kellogg, ID
E. St. Louis, IL
Marseilles, IL
Depue, IL
Morris, IL
Joliet, IL
Streator, IL
Tuscola, IL
Ft. Madison, IA
Donaldsonville,
IA
Geismar, LA
Hahnville, LA
Uncle Sam, LA
Pascagoula, MS
-— ir • • • " r • ' jj ir ™ •-•• — ----- —
Stack Height
Status of Stack (m)
Inactive
Operating
Operating
Inactive
Operating
Idle(a)
Operating
Operating
Inactive
Operating
Operating
Operating
Operating
Operating
Operating
Operating
Operating
Operating
Operating
Inactive
Operating
Operating
Idle
Operating
Inactive
Inactive
Inactive
Inactive
Inactive
Operating
Inactive
Idle(a)
Inactive
Inactive
Idle
Inactive
Inactive
Inactive
Operating
Idle
Idle
Idle
Operating
Operating
Operating
Operating
23 W
21
21
9
28
40
10
27
9
20
54
6
27
24 (c)
22
20
18
18
24
18
23
24
12 (d)
20 (d)
8(e)
s(®)
8(e)
9
9
13
4
27
5
18
16
30
9
5
12
20 (a)
12 (a)
12 (a)
6(a)
4
20
20
Base
Area (ha)
9
140 (a)
121
50
162
146
32
31
11 (a)
92
138
64
227
157 (c)
40
40
53
30
18
20
61
36
17 (d)
81
2(e)
5(®)
20 («)
7
18
40
28
85 (a)
8(a)
10
32
20
20
24
203 (f)
38 (a)
14 (a)
11 (a)
9(a)
9
284 (9)
101
13-3
-------
Table 13-1. Hie location and characteristics of phosphogypsum stacks in the
United States (continued).
Stack Height Base
Facility Name location Status of stack (m) Area (ha)
Fanners Chemical Co.
W.R. Grace and Co.
Texasgulf Chemicals
Amoco Oil Co.
Kerley Agricultural
Chemicals of Texas
Mobil Mining and
Minerals Div.
Phillips Chemical Co
Chevron Chemical Co.
Chevron Chemical Co.
1
2
Co. 1
2
3
4
5
1
2
Inc.
1
2
3
•
Joplin, MO
Joplin, MO
Aurora, NC
Texas City, TX
Pasadena, TX
Pasadena, TX
Pasadena, TX
Magna, UT
Rock Springs, WY
Inactive
Inactive
Inactive
Idle (a)
Idle (a)
Idle (a)
Idle (a)
Operating (a)
Idle
Idle
Inactive
Inactive
Inactive (a)
Operating
Idle
Inactive (n)
Operating
15
10 (a)
10 (a)
26 (a)
18 (a)
38 (a)
19 (a)
20 (a)
11
3
11
27
27
30
27
5
10 (i)
28
10
10
16 (a)
30 (a)
5l(a)
51 (a)
51 (a)
14
2
11
24
36
61
14
121
182
(a) Jo88c.
(b) Numbers 1,2,3, etc. refer to different stacks at a facility.
(c) Ba88; (d)Si88; (e)Ap88; (f)Wa88b; (g)Wa88a; (h)Co88; (i)Default value.
Note: Information in this table is from PEE85, except for that identified by
footnotes (a), and (c) to (i).
Table 13-2. Summary of the phosphogypsum stacks in each state.
Total Base Areas, hectares (a)
State
Number of
Stacks
Operating
Idle
Inactive
Arkansas
Florida
Idaho
Illinois
Iowa
Louisiana
Mississippi
Missouri
North Carolina
Texas
Utah
Wyoming
Total
1
20
6
8
3
7
1
3
5
7
1
1
63
1343 (16)
117 (2)
40 (1)
0
505 (4)
101 (1)
0
51 (1)
61 (1)
0
182 (1)
2400 (27)
0
146 (1)
17 (1)
117 (2)
0
63 (3)
0
0
148 (4)
30 (3)
0
0
521 (14)
9 (1)
81 (3)
27 (3)
71 (5)
64 (3)
0
0
48 (3)
0
71 (3)
121 (1)
0
492 (22)
(a) Number of stacks is shown in parentheses.
13-4
-------
is 3,413 ha, of which 71 percent is associated with operating
stacks, 15 percent with idle stacks, and 14 percent with inactive
stacks.
13.1.2 Composition of Phosphooypsum
Phosphogypsum is primarily calcium sulfate, CaS04'2H20,
which is only slightly soluble in water, about 2 g/1. The phos-
phogypsum contains appreciable quantities of uranium and its
decay products. This is due to the high uranium concentration in
phosphate rock which ranges from 20 to 200 ppm uranium-238 (6.7
to 67 pCi/g)(a). This is 10 to 100 times higher than the uranium
concentration in typical rocks (1 to 2 ppm). The radionuclides
of significance are: uranium-238, uranium-234, thorium-230,
radium-226, radon-222, lead-210, and polonium-210. When the
phosphate rock is processed through the wet process, there is a
selective separation and concentration of radionuclides. Most of
the radium-226, about 80 percent, follows the phosphogypsum,
while about 86 percent of the uranium and 70 percent of the
thorium are found in the phosphoric acid (Gu75).
Table 13-3 shows the average radionuclide concentrations
measured in 50 phosphogypsum samples collected in 1985 by EPA
from five stacks in central Florida (Ho88a). For comparison, the
radionuclide concentrations normally observed in uncontaminated
rock and soil are also presented. The concentrations measured in
the phosphogypsum samples are similar to those previously re-
ported (Gu75) and exceed those in background soil by 10 (uranium)
to 60 (radium-226) times. These radionuclides and radon-222 are
possible sources of airborne radioactivity. Radon-222, the decay
product of radium-226, is a gaseous element which may diffuse
into the air. Also, these radionuclides in particulate form may
be resuspended into the air by wind and vehicular traffic. These
are the two principal mechanisms for airborne releases of radio-
activity from phosphogypsum stacks that will be addressed in this
assessment.
Table 13-3. Average radionuclide concentrations in
phosphogypsum, pCi/g dry weight.
Material Ra-226 U-234 U-238 Th-230 Po-210 Pb-210
Gypsum
Background
Soil
31
0.5
3.3
0.3
3.2
0.3
5.1
0.3
27
0.5
36
0.7
(a) 1 ppm U-238 = 0.333 pCi/g or 0.67 pCi/g total uranium,
U-238 + U-234.
13-5
-------
13.1.3
Existing Control Technology
The phosphate industry does not actively pursue the control
of radon emissions from phosphogypsum stacks (Jo88a, Be88a).
However, the crust that forms naturally on inactive stacks or
over inactive areas of operating stacks significantly reduces the
radon emissions. Water maintained on active portions of ©Derat-
ing stacks also deters radon emissions.
There is no uniform or widespread effort or policy within
the phosphate industry to control particulate emissions. Dust
control measures, consisting of either spraying dusty areas with
water or establishing vegetation on areas subject to wind or
water erosion, have been applied at some stacks.(a) Both Con-
sery, Inc. (Nichols, Florida) and Mobil Chemical Company (Depue,
Illinois) have used water at times to control dusty areas. The
following companies have either planted vegetation or allowed the
natural development of indigenous vegetation in areas subject to
wind or water erosion: Northern Petrochemical Company (Morris
Illinois), Agrico Chemical Company (Ft. Madison, Iowa, and Don-
aldsonville, Louisiana), and Mobil Mining and Minerals Division
(Pasadena, Texas).
Apparently, special effort has been made at the Gardinier
stack to stabilize the sloping sides. The sides of the stack
were covered with 8 to 15 cm of topsoil and then seeded into a
hardy grass. This control measure not only eliminated erosion in
the area seeded, but the added top soil attenuated the radon-222
flux by an average of about 23 percent (Ha85).
Thus, some effort has been made at phosphogypsum stacks to
control erosion, which has often led to a reduction in airborne
emissions. However, in general, particulate emissions have not
been considered sufficient to warrant controls, primarily because
these emissions are naturally deterred as a result of the crust
that exists on inactive surfaces of a stack and the water cover
or high moisture content of gypsum on active portions of operat-
ing stacks.
Exclusion fences and/or company patrols prevent access by
the public to most stacks, which averts unauthorized entry onto
the stacks as well as the removal of any phosphogypsum.
13.1.4 Byproduct Uses of Phosphogypsum
Byproduct uses of phosphogypsum fall into three categories:
(1) chemical raw material, (2) agricultural applications, and (3)
construction material (L185).
The first category involves the recovery of sulfur from the
phosphogypsum, which is only at the experimental stage in the
United States. A pilot plant was scheduled to begin operation in
(a) Information obtained in a 1985 survey of individual companies
(PEI85).
13-6
-------
the fall of 1988 at the Agrico Chemical plant at Uncle Sam,
Louisiana (Kr88). The sulfur recovered from the phosphogypsum is
used in the manufacture of sulfuric acid, which is necessary to
produce phosphoric acid by the wet process. An aggregate or lime
may be a byproduct of the sulfur recovery process which could
improve the economic feasibility of the process (Ni88).
Phosphogypsum has many agricultural applications. As phos-
phogypsum hastens the leaching of salts from soil, it is espe-
cially useful as an amendment to salty soils (L185). About
180,000 MT/y are shipped from the Chevron plant in Utah to Cali-
fornia for use as a soil conditioner for sodic soils (Kr88).
Phosphogypsum from California stacks was sold for the same pur-
pose at a rate of 270,000 MT/y until all stacks were exhausted.
As a fertilizer, it is an excellent source of sulfur and calcium.
For example, phosphogypsum has been used on peanut crops in North
Carolina and Georgia for many years (L185). Other peanut produc-
ing states, e.g., Alabama, South Carolina, Texas, and Virginia,
also use phosphogypsum on their crops (Kr88).
Typical agricultural application rates are 2 MT per hectare
when used as a fertilizer; as a soil amendment, an initial appli-
cation of 20 MT per hectare is followed by biannual applications
of 10 MT per hectare (Li80). According to calculations by
Roessler (Ro86), the application every four years of 2 MT per
hectare over a 50-year period with no radium removal would add
0.38 pCi/g radium-226 to the soil, assuming a phosphogypsum
specific activity of 30 pCi/g, a soil till depth of 15 cm, and a
soil density of 1.5 g/cc. As a soil amendment, an additional
4.1 pCi/g radium-226 is incorporated into the soil based on the
assumptions outlined above. Background soil in central Florida
contains about 0.5 pCi/g radium-226 (Table 13-3).
As a construction material, phosphogypsum has a variety of
applications, especially in other countries. No phosphogypsum is
currently used in the United States for the manufacture of gypsum
wallboard. However, radon measurements conducted in a room
constructed of Japanese phosphogypsum wallboard at EPA's Eastern
Environmental Radiation Facility could not detect any increase in
the indoor radon concentration (Se88). The emanation fraction
was believed to be less than 2 percent. In this country,
phosphogypsum's primary use is in road construction. Combining
fly ash or cement with phosphogypsum produces a mixture suitable
for road bases. This has been demonstrated in the Houston,
Texas, area (L185, Kr88). A demonstration road and parking area
is planned in Bartow, Florida, that will use phosphogypsum in
both the road bed and surface materials (F188). All previous
uses of phosphogypsum for road construction in Florida have been
limited to use as a road base.
Less than one million MT/y of phosphogypsum are being used
in the United States at present. This represents about 1 or 2
percent of the U.S. annual production. The bulk of the usage is
for agricultural applications in California and the peanut
producing states in the southeast (about 450,000 MT/y). The
13-7
-------
remaining quantities are sold for road bed construction in Texas
and Florida (about 140,000 MT/y) (Kr88).
Historic usage since 1984 shows a general decline, primarily
due to the closing of the California facilities, as seen in the
following data (Jo88b):
1984 660,000 MT
1985 460,000 MT
1986 540,000 MT
1987 360,000 MT
These totals are based on the results of a survey of 22 of
the 42 facilities listed in Table 13-1. Since some companies
representing one or more facilities did not respond to the survey
(Jo88b), some disagreement between the mail survey (Jo88b) and
the telephone survey by Kramer (Kr88) mentioned previously is
expected in the totals. While neither survey represents a total
response for the industry, each survey gives an approximate total
usage rate.
13.2 RADIONUCLIDE EMISSIONS
This section presents estimates of the quantity of radon-222
and radioactive particulates emitted to the air from phosphogyp-
sum stacks. The quantity of radon-222 emitted annually from each
stack is estimated for realistic conditions regarding the radon
fluxes, stack areas, and particularly the hydrology of the stack
surface.
Only the radioactive particulate emissions associated with
vehicular traffic on or near a model stack are considered. Wind
suspended particulate emissions are not a significant source of
radioactivity because of the moisture content of the gypsum in
operating stacks and the crust that forms on inactive stacks.
13.2.1 Radon-222 Emissions
The amount of radon-222 emitted from phosphogypsum stacks
depends on highly variable factors, such as the uranium (and
radium-226) concentration in phosphate rock, emanation fraction,
vegetation cover, porosity, moisture, temperature, and barometric
pressure. These factors, in turn, vary between sites, between
locations on the same site, and with time (Ha85). These
variations make it difficult to assess the radon-222 emission
rate unless many flux measurements are made (Ho88a).
The amount of radon-222 released annually from phosphogypsum
stacks was estimated by dividing the stack into separate regions
with significantly different radon fluxes and measuring the flux
from the surface area of each region. The radon-222 flux is the
amount of radon-222 (picocuries) that escapes from a given area
of stack surface (square meters) during a given time (seconds).
The regions considered on active stacks were those covered by
water (ponds and ditches) or saturated by water (beaches), sur-
face areas consisting of dry, loose material, the roadway along
13-8
-------
the stack top, and the thinly crusted stack sides. Only two
regions were considered on inactive stacks, the hard, thick-
crusted top and the dry, thinly crusted sides. The radon fluxes
for each of these regions were determined by measurements (Ho88a,
B188). A summary of the results is given in Table 13-4. Except
for the beaches, which are saturated land masses that protrude
into the ponds, sufficient measurements were obtained in each
region to result in an average value acceptable for this assess-
ment. (a) Because the beaches are totally saturated with water,
small flux values were expected, and additional measurements were
considered unnecessary for this assessment.
A generic stack, based on the IMC Corp. stack near Mulberry,
Florida, which consists of the regions defined above and is
representative of Florida phosphogypsum stacks, was used to
estimate the radon-222 source terms. The base area and height of
each stack are known (see Table 13-1). The areas of the top and
sides were estimated using these dimensions and assuming the
stacks to be rectangular (length twice the width) with a 1:3
(0.333) slope to the sides, except for those stacks noted in
Appendix 13-A. The areas, so computed, are listed in Appendix
13-B. The fluxes associated with the various regions of the
stack and the percent of the regional areas to the total top area
are listed in Table 13-5.
For active Florida stacks, 60 percent of the top was consid-
ered to be covered by water resulting in no radon release. The
fluxes for the other stack regions are the average values from
Table 13-4. Roadways on active stacks were considered to consist
of 50 percent loose material (20 pCi/m2/s) and 50 percent dry,
hard-packed material (6.8 pCi/m2/s), or 13 pCi/m2/s radium-222.
The average radon fluxes for the thick, hard-crusted top surface
and dry, thin-crusted sides of inactive stacks are the averages
of measured values listed in Table 13-4. The characteristics of
idle stacks appear intermediate between those of active and
inactive stacks. The top surface is either mostly covered by
water (if the stack is a part of the plant water balance) or dry
with a thick, hard-crusted surface, similar to an inactive stack.
Thus, a conservative radon flux of 4 pCi/m2/s was applied to
the total top area of idle stacks. The radon flux applied to the
sides of idle stacks in Florida was 12 pCi/m2/s, the midpoint
between the average flux measured on the sides of active and
inactive stacks.
Since all phosphogypsum stacks, except for those located in
northern Florida, North Carolina, Idaho, Utah, and Wyoming,
resulted from processing central Florida phosphate rock, their
fluxes were considered the same as on stacks in central Florida
(see Table 13-5, Column 2). However, stacks located in Louisiana
were considered an exception to this because of significant
climatic differences between the two regions that result in a
greater rainfall vs. evaporative rate in Louisiana. Fluxes on
(a) Average values are the arithmetic means.
13-9
-------
Table
of
Florida (Ho88a, B188).
Stack Region/Facility
merits on phosphogypsum stacks in
Flux
Number of
Measurements
Range
Average W
Tee
Loose-Dry Material
Conserv (Mulberry, FL)
Gardinier (East Tampa, FL)
Grace (Bartow, FL) (b)
Royster (Mulberry, FL)
Beaches(°)
IMC Corp. (Mulberry, FL)
Roadway (dry-hard pack)
ACTIVE STACKS
128
336
519
126
Grace (Bartow, FL)
Sides
Royster (Mulberry, FL)
Grace (Bartow, FL) (b)
Top
Estech (Bartow, FL)
Sides
Royster (Mulberry, FL)
23
98
75
INACTIVE STACKS
130
99
2 -340 25
0.2 - 99 20
0.2 - 65 16
0.6 - 81 21
0.35- 0.71 0.5
1.2 - 16
1.3 - 23 7
1.7 - 40 11
0.6 - 14
4-44 15
(a) Average values are the arithmetic means.
(b) Now the Seminole Fertilizer Corporation.
(c) Measurements made by IMC Corp. personnel (Ba88).
13-10
-------
Table 13-5. Radon-222 flux values applied to various regions of phosphogypsum stacks,
Region of
Stack
Central
Florida (a)
Flux (Percent of Top Area) pCi/n^/s
North
Florida (b)
North
Carolina
Louisiana(°) Idaho(d)
Wyoming
and Utah (e>
CO
I
Top
Pond/Ditches
Beaches
Dry material
Roads
Top
ACTIVE STACKS
0.0
0.5
20
13
(60%)
(15%)
(20%)
(5%)
0.0
0.2
8
5
(60%)
(15%)
(20%)
(5%)
0.0
0.1
4
3
(60%)
(15%)
(20%)
(5%)
0.0
0.3
13
9
(60%)
(15%)
(20%)
(5%)
0.0
0.5
4.5
10
(25%)
(5%)
(65%)
(5%)
0.0
0.1
1.2
3
(25%)
(5%)
(65%)
(5%)
(f)
INACTIVE STACKS
(f)
14
Sides
Sides
15
4
12
(f)
(f)
(f)
(f) (f)
IDLE STACKS
1 2.6
2 8
9.5
7
9.5
2.5
(f)
(f)
(a) Values applied to stacks in all states except North Carolina, Louisiana, Idaho, Utah, Wyoming,
and in the White Springs region of Florida.
(b) Values apply to the three Occidental Chemical Co. stacks near White Springs, FL
(Jo88d, Ma82, Ro79).
(c) St88b.
(d) Si88.
(e) Co88.
(f) Stacks of this category do not exist in this state or region.
-------
the Louisiana stacks were based on the results of measurements
made on the sides and beach areas of two Louisiana stacks
(St88b). The fluxes that relate to the top-dry material and
roads were determined by assuming the same ratios with the sides
as on stacks in central Florida, 20/9 and 13/9, respectively, and
multiplying these ratios by the flux measured on the sides of the
Louisiana stacks, 6 pCi/m2/s (see Table 13-5, column 5). Except
for the stacks noted in Appendix A, areas of stack top and sides
were determined assuming a slope to the stack sides of 0.333.
Fluxes associated with stacks located in northern Florida
and North Carolina were determined by using measured radium-226
concentrations of 12, 6, and 31 pCi/g for north Florida (Ro79,
Ma82, Jo88d), North Carolina(a), and central Florida phosphogyp-
sum, respectively, and scaling to the flux values used for the
central Florida stacks. For example, the flux that relates to
dry-loose areas of northern Florida stacks is computed to be
8 pCi/m2/s of radon-222 (12/31 x 20).
The phosphogypsum facilities in Idaho process rock obtained
nearby, whereas facilities in both Wyoming and Utah process, or
had processed, phosphate rock that is mined near Vernal, Utah.
The flux values for stacks located in Idaho, Utah, and Wyoming
were determined using a model based on the characteristics and
operation of the two J.R. Simplot stacks near Pocatello, Idaho,
and differences in the radium-226 concentrations of the phospho-
gypsum produced. The arid conditions in this region (low rain-
fall, high rate of evaporation) and the low water content of the
phosphogypsum slurry result in stack conditions considerably
different from those observed at Florida stacks, especially on
the top surface. A much smaller pond (relative to the total top
surface area) exists on the top of the active stack, while no
water was present on the idle stack, and a thick, hard crust
covered a large fraction of the active top surface, which forms
rapidly in the dry climate.
These different conditions are reflected in the radon flux
values measured on the two J.R. Simplot Co. stacks in Idaho (see
Table 13-6). Due to the thick, hard crust, the top surface is
similar to the inactive stacks in Florida, while the flux from
the sides is not significantly different from that measured in
Florida. The radon flux values assumed for each region of the
Idaho stacks are listed in Table 13-5. Fluxes on the roadways
were not measured but were estimated to be 10 pCi/m2/s based on
the Florida fluxes of 20 pCi/m2/s for loose material (25 percent
of roadway) and 7 pCi/m2/s for dry, hard-packed material
(75 percent of roadway). The percentages of the top areas cov-
ered by water and saturated as beach area are much lower than on
a Florida stack, while a much higher percentage is considered as
dry material. Also, the idle stacks (J.R. Simplot Co.) and
inactive stacks (Bunker Hill Co.) in Idaho are considered similar
and were assigned identical fluxes for their top and side areas.
(a) Based on unpublished results of analyses conducted at the
EPA's Eastern Environmental Radiation Facility.
13-12
-------
Table 13-6. Results of radon-222 flux measurements on
phosphogypsum stacks in Idaho (Ly88, Ho88b).
Number of Flux fpCi/m2/s)
Measurements Range Average
Active Stack
Top 41 0-20 4.5
Sides 10 4-31 14
Idle Stack
Top 16 1-30 7.3
Sides 5 2-18 9.5
The stacks in Utah and Wyoming are treated identically to
the Idaho stacks because of similar climate and presumed similar
facility operation. The radium-226 content of phosphogypsum
resulting from phosphate rock mined near Vernal, Utah, is known
to be low, about 5 to 8 pCi/g (Co88). Similar to the calculation
used above for the northern Florida and North Carolina stack
fluxes, concentrations of 6.5, 6.5, and 25 pCi/g radium-226 were
used for the Utah, Wyoming, and Idaho phosphogypsum, respectively
(Co88, Ho88b). By scaling to the Idaho regional stack values
(see Table 13-5, Column 6), fluxes for the Utah and Wyoming
stacks were determined. For example, the flux that relates to
the sides of the active stack located in Wyoming is 4 pCi/m2/s
radon-222 (6.5/25 x 14).
Estimates of the annual radon-222 emissions from individual
phosphogypsum stacks are presented in Table 13-7. These emis-
sions were calculated using the information given in Table 13-5
and the stack top and side areas listed in Appendix 13-B. The
resulting emission rates are expressed in Ci/y. For example, the
radon-222 emission rate for the IMC Corporation's Mulberry,
Florida, stack is determined by the following expression:
Emission Rate = { 1.211xl06m2 [0.5 pCi/m2/s (0.15)
+ 20 pCi/m2/s (0.20) + 13 pCi/m2/s (0.05)]
+ 3.821xl05m2 (9 pCi/m2/s) }
10~12 Ci/pCi x 3.16 x 107 s/y
= 289 Ci/y.
Note: This emission rate has been rounded to 290 Ci/y in Table
13-7.
13-13
-------
The total emissions from all stacks are approximately
5,700 Ci/y (i.e., the sum of all individual source terms in
Table 13-7). About half of the total emissions are from Florida
(approximately 2,900 Ci/y).
The last column of Table 13-7 gives the average radon-222
flux values computed for the overall (top and sides) stack sur-
face. The average flux for the 63 stacks ranges from about
12 pCi/m2/s to 1 pCi/m2/s. Excluding the 10 stacks that
consist of low-radium content phosphogypsum, North Carolina (5),
northern Florida (3), Utah (1) , and Wyoming (1), the average
fluxes for active and inactive stacks are 5.9 and 8.2 pCi/m2/s,
respectively, and for all 53 stacks the average flux is
7.0 pCi/m2/s. The average flux for the ten stacks that consist
of low-radium content phosphogypsum is 1.8 pCi/m2/s. The
average flux values fpr active stacks will increase as the stacks
become taller.
13.2.2 Radioactive Particulate Emissions
The emission rates due to vehicular traffic were estimated
by using an EPA fugitive dust emissions model (EPA77). This
model is applicable to dust particles with effective diameters of
30 urn or less and considers the silt content of the soil, mois-
ture, and the average vehicle speed. The emission factors gener-
ated by this model yield the quantity of fugitive dust emissions
from unpaved roads per vehicle-mile of travel. Wind erosion was
not considered because quantities of dust produced from this
source are insignificant due to the moisture content of active
stacks and the crust that exists on inactive stacks.
The quantity of fugitive dust emissions from vehicular
traffic on unpaved roads at a phosphogypsum facility was esti-
mated using the following empirical equation (EPA77):
E = (0.81s) X _S X 365-w x f (13-1)
30 365
Where: E = Emission Factor, pounds per vehicle-mile
s = Silt content of road surface, percent (100 percent)
S = Average vehicle speed, miles per hour (30 miles/hr)
w = Mean annual number of days with 0.01 inches or more
of rainfall (120 days)
f = Average fraction of emitted particles in the <30 urn
diameter suspended particle size range; particles
having diameters greater than 30 urn will settle
rapidly near the roadway (0.32).
Equation 13-1 is reported to be valid within ± 20 percent
for vehicle speeds in the range of 30 to 50 miles per hour
(EPA77).
13-14
-------
The values shown in the parentheses following the definition
of each parameter for Equation 13-1 are believed appropriate to a
phosphogypsum facility. Applying these values to Equation 13-1
resulted in an emission factor of 17.4 Ibs per vehicle-mile. The
total annual emissions from the model stack, 1.97E+7 g/y, was
based on an estimated 2,500 miles of traffic per year (10 miles/
day X 5 days/week X 50 weeks/year). This distance relates to a
31-ha stack which represents a conservative estimate of the
traffic observed at the 30- to 32-ha stacks at Royster and
Conserv during the long-term EPA study (Ho88a). The annual
radionuclide emissions associated with fugitive dust, listed in
Table 13-8, were determined by multiplying the total annual
emissions, 1.97E+7 g/y, by the average concentrations of radionu-
clides in phosphogypsum that are listed in Table 13-3.
To help assess the significance of particulate emissions and
the applicability of the above model, airborne particulate sam-
ples were collected upwind and downwind of a gypsum stack. High-
volume airborne particulate samplers were operated continuously
for a four-month period at upwind (460 m southeast) and downwind
(115 m northwest) locations of the W.R. Grace(a) stack No. 2 in
Bartow, Florida. Background airborne particulate samples were
collected concurrently in a region of Polk County, Florida, that
is unaffected by the phosphate industry. Filters, replaced
weekly, were combined into monthly samples and analyzed for their
radionuclide content. Concentrations of radionuclides determined
were adjusted for the background contribution and for the small
amounts of radionuclides present in unexposed filters (Ho88a).
The average net concentrations of radionuclides determined for
the upwind and downwind locations are presented in Table 13-9.
The activity ratios of the radionuclides measured in the particu-
late samples do not reflect those in phosphogypsum (see
Table 13-3), which strongly indicates that the source of the
material collected by the high-volume samplers was not the phos-
phogypsum stack. Also, the very low radionuclide concentrations
measured in the airborne samples, less than a femtocurie per
cubic meter, demonstrate the insignificance of this exposure
pathway at phosphogypsum stacks.
(a) Now the Seminole Fertilizer Corporation.
13-15
-------
Table 13-7.
Estimates of annual radon-222 emissions from phosphogypsum
stacks.
Facility Name
Districhem, Inc.
Agrico Chemical Co.
Royster Phosphate, Inc.
Brewster Phosphates
CF Industries, Inc.
CF Industries, Inc.
Conserv, Inc.
Estech, Inc.
Farmland Industries, Inc.
Gardinier, Inc.
Seminole Fert. Corp.
IMC Corp.
Occidental Chemical Co.
(Suwannee River)
Occidental Chemical Co.
(Swift Creek)
Royster Co.
USS Agri-Chemicals, Inc.
USS Agri-Chemicals, Inc.
Nu-West Industries, Inc.
J.R. Simplot Co.
Bunker Hill Co.
Allied Chemical Co.
Beker Industries, Corp.
Mobil Chemical Co.
Northern Petrochemical Co.
Olin Corp.
SECO, Inc.
U.S. Industrial
Chemicals Co.
Agrico Chemical Co.
l(b)
2
1
2
1
2
1
2
1
2
1
2
3
1
2
1
2
3
location
Helena, AR
Bartow, FL
Palmetto, FL
Bradley, FL
Plant City, FL
Bartow, FL
Nichols, FL
Bartow, FL
Bartow, FL
Tampa, FL
Bartow, FL
Mulberry, FL
White Springs, FL
White Springs, FL
Mulberry, FL
Bartow, FL
Ft. Meade, FL
Conda, ID
Pocatello, ID
Kellogg, ID
E. St. Louis, IL
Marseilles, IL
Depue, IL
Morris, IL
Joliet, IL
Streator, IL
Tuscola, IL
Ft. Madison, IA
(a) The Rn-222 flux averaged over all regions of the
(b) Numbers 1, 2, 3, etc.,
refer
to different stacks
Average
Rn-222 Rn-222
Emissions Flux (a)
(CVy) (pCi/m2/s)
32
250
220
92
310
340
58
71
27
170
310
100
400
290
36
35
43
62
43
59
120
97
43
170
6
13
50
19
40
75
45
190
16
35
69
77
43
42
stack.
at a facility.
10.4
5.7
5.7
5.8
6.0
7.2
5.7
7.0
7.7
5.8
6.9
4.9
5.6
5.7
2.8
2.7
2.5
6.4
7.3
9.1
6.1
8.3
7.9
6.4
9.4
8.2
7.8
8.4
6.9
5.9
5.1
6.8
6.5
10.7
6.7
11.7
6.7
5.5
13-16
-------
Table 13-7.
Estimates of annual radon-222 emissions from phosphogypsum stacks
(continued).
Average
Rn-222 Rn-222
Emissions Flux (a)
Facility Name
Agrico Chemical Co.
Arcadian Corp.
Agrico Chemical Oo.
Agrico Chemical Oo.
Nu-South Industries, Inc.
Farmers Chemical Oo.
W.R. Grace and Oo.
Texasgulf Chemicals Oo.
Amoco Oil Co.
Kerley Agricultural
Chemicals of Texas, Inc.
Mobil Mining and
Minerals Div.
Phillips Chemical Co.
Chevron Chemical Oo.
Chevron Chemical Oo.
1
2
3
4
1
2
1
2
3
4
5
1
2
1
2
3
location
Donaldsonville, IA
Geismar, IA
Hahnville, IA
Uncle Sam, IA
Fascagoula, MS
Joplin, MO
Joplin, MO
Aurora, NC
Texas City, TX
Pasadena, TX
Pasadena, TX
Pasadena, TX
Magna, UT
'Rock Springs, WY
(a) The Rn-222 flux averaged over all regions of the
(b) Numbers 1, 2, 3, etc.,
refer
to different stacks
(Ci/y)
230
57
26
21
12
11
380
250
70
26
26
8
13
24
21
20
31
4
30
83
110
130
46
78
71
stack.
(pCi/m2/s)
4.0
4.7
5.8
6.1
4.2
3.9
4.2
7.8
7.8
8.1
8.1
1.5
1.3
1.4
1.3
1.2
6.9
6.2
8.5
10.6
9.7
6.6
10.0
2.0
1.2
at a facility.
Table 13-8.
Annual radionuclide emissions in fugitive dust from
a model 31-ha phosphogypsum stack.
Radionuclide
Emission Rate
(Ci/y)
Radionuclide
Emission Rate
(Ci/y)
Uranium-238
Uranium-234
Thorium-230
Radium-226
Radon-222 (a)
(ft) Assumed to
6.3E-5
6.5E-5
l.OE-4
6.1E-4
6.1E-4
be in radioactive
Lead-214(a)
Bismuth-214 (a)
Lead-210
Folonium-210
equilibrium with radium-226.
6.1E-4
6.1E-4
7.1E-4
5.3E-4
This results
in a maxtrrum value.
13-17
-------
Table 13-9. Average net airborne radionuclide concentrations
measured at the W.R. Grace stack.(a)
Average Net Concentrations, pCi/m3(b)
Location U-238 U-234 Th-230 Ra-226
Upwind
Downwind
l.OE-4
1.5E-4
1.1E-4
1.6E-4
1.1E-4
1.5E-4
1.1E-4
1.9E-4
(a) Now the Seminole Fertilizer Corporationc
(b) Concentrations after background values have been subtracted.
13.2.3 Methodology
The location of the maximum exposed individual was deter-
mined at each stack by using official county highway and U.S.
Geological Survey maps to locate the residence nearest to the
stack in each of 16 annular sectors. In some cases, individual
companies supplied updated locations (Jo88c, TFI89). The
AIRDOS-EPA (Mo79) and DARTAB (Be81) codes were then used to esti-
mate the maximum exposure to radon-222 and the highest increased
chance of lung cancer for an individual in one of these actual
residences. The radon-222 decay product equilibrium fraction at
the residence was determined as a function of the distance from
the stack. The dose equivalents resulting from radioactive
particulate emissions were estimated by using airborne pathway
models for inhalation, ingestion, ground contamination, and
immersion, followed by application of the above computer codes.
Collective risks and dose equivalents for the regional
population due to radon-222 and radioactive particulates were
calculated from the annual collective exposure (person WLM) and
collective dose equivalents (person rem), respectively, using
AIRDOS-EPA and DARTAB codes. Exposure pathways were identical to
those applied to the maximum exposed individual. The population
distribution within 80 km of each stack was determined using the
computer program SECPOP (At74), which utilizes 1980 census data
to compute the population in each annular sector. Collective
exposures to radon-222, expressed in person WLM, were estimated
for each stack by multiplying the estimated radon-222 progeny
concentration (WL) in each annular sector by the population in
that sector and by the conversion factor 51.56 WLM/y per WL. The
parameters used in the AIRDOS-EPA code for each stack are shown
in Appendix A and in Tables 13-1 and 13-7. Meteorological param-
eters from selected nearby weather stations were used for each
stack. The cumulative WL exposure of each population segment was
adjusted using a radon decay product equilibrium fraction that is
related to the distance from the center of the stack to that
population segment.
An emission height of 1 m was assumed for all stacks. This
is a conservative assumption which may overestimate the maximum
13-18
-------
individual risk but not significantly in most cases. Figure 13-1
iSows the effect of release height on the fatal cancer risk from
a model stack with a base area of 121 ha. Beyond a distance of
800 m from the center of the model stack, the individual risk is
the same for 1 m and 12.5 m release heights. A more realistic
release heiqht may be a value of one-half the physical stack
helghS? Forexample, a release height of 12.5 m would be assumed
for a stack that is 25 m tall. This is reasonable considering
that a significant fraction of the radon emissions occurs from
the sides of the stack and that radon from both the sides and top
of the stack is carried toward the ground near the base of the
stack as a result of downwash.
Only 3 of the 63 stacks have physical heights that exceed
30 m. At only 2 of the 60 stacks with physical heights of 30 m
or less does an individual reside less than 800 m from the center
of the stack. By using the 1-m release height assumption, only
two individuals have lifetime fatal cancer risks that may be
overestimated by a factor of two or three. The 1-m stack height
assumption has essentially no effect on the population risk
assessments, because nearly all of the exposed J^1*1*"*' i" *£
regions reside many kilometers from the stack where stack height
has no significant influence on the risk calculation.
The maximum annual dose equivalents and the increased risk
of fatal cancer to nearby individuals from fugitive dust emis-
sions were estimated by determining the total annual emissions
from model stacks using the EPA fugitive dust model (Section
132.2, Table 13-8) and applying the AIRDOS-EPA (Mo79) and DARTAB
(Be81) codes. The model stacks were assumed to be in Polk
County, Florida. An average model stack was derived that had the
average base area, 90 ha, of the 27 presently operating phospho-
qypsum stacks. Also, minimum and maximum generic stacks were
considered and assigned base areas of 9 ha and 284 ha, respec-
tively, which reflect the smallest and largest existing active
stacks (see Table 13-1). Vehicular traffic on a stack, and thus
emissions, is assumed proportional to the stack area. Inactive
and idle stacks are assumed to have no vehicular traffic. The
maximum exposed individual was assumed to live about 1,750 m from
the center of the stacks.
The ICRP lung model was used in this assessment (ICRP66).
To apply this model, a 1.0 urn (AMAD) particle size was assumed
with the following lung clearance classes:
Y class - U-238, U-234, Th-230
W class - Ra-226, Bi-214, Po-210, Pb-210
D class - Pb-214
Collective (population) risks for the region due to fugitive
dust emissions from vehicular traffic are based on the assess-
ments of small, average, and large phosphogypsum stacks located
in Polk County, Florida, which have areas of 9 ha, 90 ha, and
284 ha respectively (see Table 13-1). Emissions from the small,
average, and large generic stacks were estimated by multiplying
13-19
-------
u>
I
NJ
O
H>
(D
rt
H-
9
0
0*
rt
01
O
o*
3
O
(D
H-
0)
x-
X
o
I
Ul
2000
4000
1.0 meter
12.5 meter
25 meter
37. 5 meter
- 50 meter
6000
8000
Distance from Center of Model Stack (meters)
Figure 13-1. Effect of release height on individual risk for a model stack.
-------
the annual radionuclide emissions from a 31-ha stack (Table 13-8)
by the ratio of their areas, 0.29, 2.9, and 9.2, respectively, as
vehicular traffic on a stack is assumed proportional to the stack
area. These annual emissions were then applied to the AIRDOS-EPA
and DARTAB codes to complete the assessment. The population and
its distribution within 80 km of the model stacks were taken from
an earlier EPA generic study performed in Polk County, Florida
(EPA84). The meteorological parameters used were taken from the
Orlando Jet Port Station.
13.3 RESULTS OF THE HEALTH IMPACT ASSESSMENT
This section contains an assessment of the dose equivalents
caused by fugitive dust emissions and the risk of cancer caused
by radon-222 and fugitive dust emissions from phosphogypsum
stacks. The health impact assessment addresses the following
specific topics:
(1) working level exposure and the lifetime fatal cancer
risk to the maximum exposed individual from radon-222 at
each phosphogypsum stack,
(2) dose equivalent rates and annual fatal cancer risks to
the maximum exposed individual from radioactive
particulate emissions at three generic stacks with
maximum, minimum, and average areas,
(3) the number of fatal cancers committed per year in the
regional population(a> at each phosphogypsum stack due
to radon-222, and
(4) the collective dose equivalent rates and fatal cancers
committed per year in the regional population from
radioactive particulate emissions at three generic
stacks with maximum, minimum, and average areas.
13.3.1 The Maximum Exposed Individual
13.3.1.1 Risks from Radon-222
In Appendix 13-C, Table 13-C-l lists the highest individual
risks for each of the 63 stacks considered in this assessment.
Included for each stack are the location of the individual with
respect to distance from the stack and the radon-222 concentra-
tion and working-level exposure at that location. The highest
lifetime individual risks are on the order of <1 fatal lung
cancers in 10,000.
The stacks that result in the 10 highest lifetime individual
risks are listed in Table 13-10 in order of descending risk.
Also listed are the location of the individual's residence in
(a) The regional population is the total number of people who
reside within 80 km of a stack.
13-21
-------
""" 13~10' 2^±f^iVidUal llfetlne risks esttotel *> «at to. .^MB -^a. ta
(O
CO
f-» ^ — «•• 3j 4'™^*™*** «* *****^-w^a •
Da/^s-vn
Facility/Location Concentration
(pa/1)
Mobil Mining & Minerals Div. ,
Pasadena, TX #3
Olin Corporation, Joliet, IL #1
Mobil Mining & Minerals Div.,
Pasadena, TX #2
Royster Phosphate, Inc. , Palmetto, FL
Agrico Chemical Co. , Uncle Sam, IA
Seminole Fertilizer Corp., Bartow, FL #2
Mobil Mining & Minerals Div. ,
Pasadena, TX #1
C.F. Industries, Plant City, FL
Kerley Agricultural Chem. of Texas, Inc. ,
Pasadena, TX
NU-West Industries, Inc., Oonda, ID
(a) Distance from the center of the stack to
2.1E-2
2.0E-2
1.9E-2
1.8E-2
1.5E-2
1.6E-2
1.4E-2
1.5E-2
1.3E-2
1.3E-2
the mav-jiffim ev
Maximum Maximum Lif
Exposure Fatal Canoe
(WL) to Indivi
6.5E-5
6.2E-5
6.1E-5
5.7E-5
5.1E-5
5.1E-5
4.9E-5
4.7E-5
4.0E-5
3.9E-5
Dosed individual
9E-5
9E-5
8E-5
8E-5
7E-5
7E-5
7E-5
6E-5
6E-5
5E-5
Distance (a)
(meters)
1,000
900
1,300
1,200
2,100
1,200
2,300
1,200
1,300
900
-------
relation to the center of the stack and the increased concentra-
tion and exposure at that location to radon-222 from the stack.
The magnitude of the individual risk is a function of the
radon-222 source term and the distance and direction of the
individual's residence from the stack. Of the 10 stacks causing
the greatest individual risks, 4 are in Texas, 3 in Florida, and
1 each in Louisiana, Idaho, and Illinois.
13.3.1.2 Dose Equivalents and Risks from Particulates
The dose equivalent rates to the maximum exposed individual
due to fugitive dust emissions from three model phosphogypsum
stacks are listed in Table 13-11. The areas of the three model
stacks relate to the areas of the smallest (minimum), average,
and largest (maximum) currently active stacks. It was assumed
that the maximum exposed individual resided 1,750 m from the
center of each stack. Only those organ dose equivalents that
contribute 10 percent or more to the risk are included in
Table 13-11. Just the lung and endosteal bone meet this
criterion. The dose equivalent rates to the endosteal bone of
the maximum exposed individuals range from a minimum of
0.04 mrem/y to a maximum of about 1.0 mrem/y. The dose
equivalent rate to the lung was about 45 percent less.
The last column of Table 13-11 lists the lifetime fatal
cancer risks to individuals living 1,750 m from the center of
each model stack. These estimated risks are conservative (i.e.,
overestimated) because the model treats all particles less than
30 urn as having an AMAD of 1 urn. Even so, these risks due to
fugitive dust emissions are one or two orders of magnitude
smaller than the risks related to radon-222 emissions (see
Table 13-10).
13.3.2 The Regional Population
13.3.2.1 Risks from Radon-222
The 10 regional populations at highest risk of fatal lung
cancer due to the radon-222 emissions from phosphogypsum stacks
are listed in Table 13-12 (see Appendix C, Table 13-C-2, for the
collective risk to the 80-km regional population around each
stack). The populations within the 80-km regions are also
listed.
13-23
-------
Table 13-11. Estimated increased risk of fatal cancer and the dose equ
rates from maximim exposure to fugitive dusts for an individual
living near phosphogypsum stacks.
. . /a) Dose'Equivalent Maximum Lifetime
Facility^ ' organ Rate, mrem/y Risk of Fatal Canner
Minimum Model Stack
Average Model Stack
Maximum Model Stack
lung
Endosteal
lung
Endosteal
lung
Endosteal
0.023
0.040
0.20
0.37
0.57
1.0
8E-8
7E-7
2E-6
(a) Ihe distance to the maximm exposed individual, as selected by the
computer code DARTAB, was 1,750 m at all three stacks.
Table 13-12. The 10 regional populations estimated to receive the highest
collective risks from radon-222 emissions from phosphoqvDsum
stacks.
1980 Population Committed Fatal Cancers
Facility/Location within 80 km per Year (0-80 km)
Mobil Mining and Minerals Div. ,
Pasadena, TX #3
Mobil Mining and Minerals Div. ,
Pasadena, TX #2
Olin Corp. , Joliet, IL #1
Mobil Mining and Minerals Div. ,
Pasadena, TX #1
Gardinier, Inc., Tampa, FL
Phillips Chemical Co. , Pasadena, TX
C.F. Industries, Inc., Plant City, FL
Kerley Agricultural Chemicals,
Pasadena, TX
Seminole Fertilizer Corp. , Bartow,
FL #2
Agrico Chemical Co., Uncle Sam, IA
3,000,000
3,000,000
7,400,000
3,000,000
2,200,000
3,000,000
2,200,000
3,000,000
1,400,000
1,900,000
1E-1
1E-1
1E-1
9E-2
5E-2
5E-2
3E-2
3E-2
3E-2
3E-2
13-24
-------
The higher collective risks result from stacks located
within or close to large metropolitan areas. The highest collec-
tive risk occurs in the densely populated Houston-Galveston,
Texas, area, where it is estimated that a fatal cancer due to
radon-222 emissions from the Mobil Mining and Minerals Division,
stack #3, will occur about every 10 years. In fact, three of the
four stacks causing the highest collective risks (and five of the
top ten) are located in Pasadena, Texas, a suburb bordering
Houston on the southeast. The Joliet, Illinois, population is
also at risk to the extent of about 1 fatal cancer every 10 years
due to the Olin Corp. stack, while 1 fatal cancer in 20 years is
estimated to occur in the regional population of the Gardinier,
Inc. stack which includes the greater Tampa, Florida, area.
An additional output of the DARTAB computer code provides
the frequency distribution of lifetime fatal cancer risks for
each phosphogypsum stack. It gives the number of people in each
of a series of lifetime risk intervals and the number of cancer
deaths that occur annually within each risk interval. This
information is summarized in Table 13-13 for all of the 63 stacks
assessed in the United States. These data reflect the number of
deaths expected to occur annually within the 0-80 km population
listed. For example, 95 million people are at risk in the 63
regional populations due to their exposure to radon-222 from all
phosphogypsum stacks, and, within that population, less than one
fatal lung cancer is expected to occur per year.
Table 13-13. Estimated distribution of the fatal cancer risk
caused by radon-222 emissions from phosphogypsum
stacks.
Risk Interval Number of Persons(a* Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
400,
17,000,
77,000,
0
0
0
0
000
000
000
0
0
0
0
9E-2
5E-1
3E-1
Totals 95,000,000 9E-1
(a)Populations are overestimated because they have not been
corrected for overlap.
13-25
-------
Similar distributions are presented in Appendix 13-D for
five regions that contain groupings of 3 to 10 phosphogypsum
stacks sited within a relatively small area. These distributions
have been summarized in Table 13-14. This summary presents the
regional populations and the number of estimated committed fatal
cancers per year resulting from the exposure to radon-222 from
all regional phosphogypsum stacks. The region of greatest risk
is Houston-Galveston, with about one fatal cancer committed every
three years due to the seven phosphogypsum stacks located in
Pasadena and Texas City, Texas. In the Bartow, Florida, regional
population, the committed fatal cancer rate drops to about 1
fatal cancer in 10 years.
Table 13-14. A summary of the committed fatal cancers due to
radon-222 emissions from phosphogypsum stacks
located in five regions in the United States.
Total
_ . ,-x Cancer Deaths
Region No. of Stacks\ai PopulationW per year
Houston/Galveston, TX
Bartow, FL
Northeast, IL
Baton Rouge/New Orleans, LA
Pocatello, ID
7
10
6
7
3
20,000,000
18,000,000
24,000,000
8,900,000
420,000
0.4
0.1
0.1
0.05
0.004
(a) See Appendix 13-D to identify stacks included in each region.
(b) Most regional populations are significantly overestimated
because they have not been corrected for overlap.
A large portion of the populations within these five regions
is exposed to emissions from more than one stack. This results
in an overestimate of the population at risk while underestimat-
ing the risk to some individuals. These distributions do not
account for overlap (exposure from multiple stacks) in the
exposed populations. An assessment for some of the stacks in
Florida suggests that the number of persons exposed in each
geographic area is overestimated by the number of stacks in the
area, while the risk is generally understated by the ratio of the
total emissions in that area to the stack with the highest
emissions in the area. For one section of Florida, it is
estimated that the number of persons exposed is overestimated by
a factor of seven, while the risks are understated by a factor of
three.
13-26
-------
13.3.2.2 Risks from Radioactive Particles
The collective risks to the regional (0-80 km) population
from the radioactivity associated with fugitive dust emissions
from the three model phosphogypsum stacks are tabulated in Table
13-15. The population, listed in the second column, is assumed
to be the same within the three regions. The risk for the aver-
age of the 27 active stacks is 2E-4, or about two fatal lung
cancers in 10,000 years. In the last line of Table 13-15 is an
estimate of the collective risk, due to fugitive dust emissions,
to all 27 regional populations within 80 km of an active phospho-
gypsum stack. This risk, 5E-3 (2E-4 x 27 stacks), is a maximum
risk because all particles less than 30 um are assumed in the
assessment to have an AMAD of 1 um and to be respirable.
Table 13-15. Estimated number of fatal cancers from fugitive
dust emissions for the population living within 80
km of the model phosphogypsum stacks.
Collective Risk
Population within (Committed Fatal
Facility an 80-km Radius Cancers per Year)
Minimum Model Stack 1,500,000 2E-5
Average Model Stack 1,500,000 2E-4
Maximum Model Stack 1,500,000 7E-4
Total United States 41,000,000 5E-3
(a) Collective risk to all individuals living within 80 km of the
27 active phosphogypsum stacks, assuming the same generic
population for each stack (i.e., 27 x 1,500,000 = 40,500,000).
13.4 SUPPLEMENTARY CONTROL OPTIONS AND COSTS
13.4.1 Introduction and Summary
This section deals with the cost and effectiveness of
mitigating radon emissions from gypsum stacks.
A preliminary examination of various means of mitigating
emissions from a representative Florida gypsum stack led to the
conclusion that the only practical mitigation technique is cover-
ing the stacks with a layer of earth sufficient to reduce the
emissions to the desired level (Be88b). A new method of gypsum
disposal is being used in North Carolina, where the gypsum is
disposed of in mined-out areas. However, this disposal technique
is still in development and may not be practical in Florida which
13-27
-------
has a high water table. None of the other states, except Idaho
Utah, and Montana, mine phosphate ore and therefore have no
mined-out areas to use. For this reason, disposal in mined-out
areas was not considered here.
Three different approaches for covering a representative
Florida gypsum stack with earth were examined (Be88b). The first
is covering the stack after it has reached the end of its useful
life. The second is covering the sides of the stack as its
height increases, but leaving the top uncovered until the stack
is closed. (The top cannot be covered during operation because
it contains the settling pond.) The third is phased disposal of
gypsum by means of a staged gypsum stack.
The only option considered here is the first, that of cover-
ing the stack after it has reached the end of its useful life.
The second option, covering the sides as the stack grows
reduces the emissions only during the life of the stack, so that
over a 50- to 100-year period the reduction is only about 10
percent more than that of the first option. The cost of this
second option is not much greater than that of covering the
entire stack when it is closed ($53 versus $56 million for the
representative Florida stack).
In the third option, a large stack is built in stages, with
two sides of each stage being covered with earth as the stage
grows, and the top of each stage being covered soon after the
next stage is put into operation. This approach (compared to the
option of covering the sides as they grow) reduces emissions only
while the stack is operating. It is estimated that in a 50-year
period, the radon emissions are about 36 percent less than in the
case where the sides of a single-stage stack are covered as it
grows and the top is covered at the end (Be88b). Over a 100-year
period, the reduction would be about 28 percent, and for longer
periods of time, the percentage reduction would be less. At very
long times, the staged stack gives off somewhat more radon than
the normal stack, because its surface area is somewhat larger.
The estimated cost of covering the representative Florida
multi-stage stack is not greatly different from that of covering
a single stack of equal volume. For the 50-meter high model
stack with a base area of 121 hectares, the costs are about
10 percent higher for the staged stack, $60 million versus
$55 million (Be88b). However, there are significant uncertain-
ties regarding its practicability. First, while the multi-stage
stack reduces emissions in the short term, eventually it emits
more radon than the large, single stack because it has more
surface area. Second, if the number and size of the stages are
not carefully selected, the multi-stage stack might always emit
more radon than a single stack. Third, while the multi-stage
approach appears to be a feasible method, there may be practical
problems not apparent at this time.
13-28
-------
Because the mitigation cost and effectiveness are not much
different, and because of the questionable practicability, the
staged-stack approach is not considered here.
The method used to estimate emissions, cost, and mitigation
effectiveness was first to group the existing gypsum stacks into
geographic categories based upon the stack characteristics
(method of construction and radon flux) and the effectiveness of
earth cover (which depends on the rainfall and evaporation rates
and the soil characteristics). Within each group, operating and
inactive stacks were considered separately, because inactive
stacks can be covered immediately, while active stacks cannot.
Idle stacks were considered to be inactive.
Because active stacks cannot be covered completely until the
end of their useful life, their size was estimated at the end of
their life using the method described in Appendix 13-E. This
leads to much larger sizes for the active stacks than the present
size used in the first part of this chapter, and hence a somewhat
larger value for the radon emissions and risk. This is a more
realistic approach than that of estimating the cost and effec-
tiveness of covering stacks that, in reality, will continue to
operate.
The risks associated with the various radon emission rates
were estimated by the method described in Appendix 13-E.
Reducing the flux to 6 pCi/m2/s reduces the annual nation-
wide deaths per year from 1 to 0.9 at a cost of about $0.5 bil-
lion dollars. Reducing the flux to 2 pCi/mVs reduces the
annual nationwide deaths per year to 0.3 at a cost of about
$1 billion.
13.4.2 Determination of Emissions. Costs, and Effectiveness
The existing stacks were separated into four groups based on
their differing cover effectiveness. The effectiveness of earth
cover is indicated by the value of b in Table 13-16. The larger
the value of b, the less radon will escape through a given thick-
ness of cover. Appendix 13-E describes the method used to esti-
mate b. The groups based on effectiveness are as follows:
1. Florida, Arkansas, North Carolina, and Texas, where
b lies between 1.6 and 1.8 m~ ,•
2. Iowa, Illinois, and Missouri, where b lies between
1.3 and 1.4 m"1;
3. Louisiana and Mississippi, where b lies between 2.2
and 2.4 m"1; and
4. Idaho, Utah, and Wyoming, where b lies between 0.77
and 0.91 m"1.
13-29
-------
Table 13-16. Characteristics of gypsum stacks.
State/Company location Area Height RN EVP b Status Capacity
Districhem, Inc.
Helena
9 23 52 43 1.8
U
Florida
Royster Mulberry
Royster Mulberry
USX (Ft. Meade Chemical) Bartow
Conserv Nichols
Conserv Nichols
Occidental white Springs
Occidental white Springs
Occidental white Springs
Estech Agricola
Brewster Bradley
USX (Ft. Meade Chemical) Fort Meade
Seminole Fert. Corp. Bartow
Seminole Fert. Corp. Bartow
Farmland Industries Pierce
Royster Phosphate, Inc. Palmetto
Gardinier Tampa
CF Industries Bartow
Agrico Bartow
CF Industries Plant city
IMC Mulberry
Idaho
J.R. Simplot
J.R. Simplot
Nu-West Industries
Bunker Hill Co.
Bunker Hill Co.
Bunker Hill Co.
Pocatello
Pocatello
Conda
Kellogg
Kellogg
Kellogg
Area = Base area, hectares.
Height = Average height, meters.
18
30
20
31
32
40
40
53
40
50
61
64
227
92
121
138
146
140
162
157
24
18
18
27
10
20
22
18
9
9
23
6
27
20
21
54
40
21
28
24
51
51
51
54
54
55
55
55
54
51
51
54
54
51
47
47
51
54
53
51
48
48
48
48
48
48
48
48
48
48
48
48
48
48
48
48
48
48
48
48
1.7
1.7
1.7
1.8
1.8
1.8
1.8
1.8
1.8
1.7
1.7
1.7
1.8
1.7
1.6
1.6
1.7
1.8
1.7
1.7
0
0
C
O
0
0
O
O
C
C
0
0
O
O
0
O
I
0
0
O
230
230
430
180
180
1020
1020
1020
U
U
430
280
280
520
170
650
630
380
760
1550
17 12
81 20
36 24
2 8
5 8
20 8
11 40
11 40
14 35
17 25
17 25
17 25
0.83 I
0.83 O
0.84 O
0.97 C
0.97 C
0.97 C
RN
Evap.
b
U
Status
Cap.
320
320
280
U
U
U
= Rainfall rate, in/yr.
= Lake evaporation rate, in/yr.
= Coefficient in R = exp(-bx), where R is ratio of the covered
to the uncovered radon flux and x is cover thickness, meters
= Unknown
= 0 is operating (active), I is idle (considered to be inactive
for the purpose of this analysis), and C is inactive (closed).
= Plant P205 production, thousands of metric tons/yr
13-30
-------
Table 13-16. Characteristics of gypsum stacks (continued).
State/Corpany location Area Height RN EVP b Status Capacity
Illinois
Allied Chemical
Olin
Olin
SECO, Inc.
Beker
E. St. Louis
Joliet
Joliet
Streator
Marseilles
Northern Petrochemical Morris
U.S. Industial Chemical Tuscola
Mobil Depue
7 9
10 5
35 27
10 18
18 9
28 4
32 16
40 13
36 39
36 35
36 35
35 36
34 37
34 36
38 36
35 38
1.3
1.4
1.4
1.3
1.3
1.3
1.4
1.3
C
C
I
C
C
C
C
o
u
110
110
u
u
u
u
110
Agrico
Agrico
Agrico
Louisiana
Agrico
Arcadian Corp.
Arcadian Corp.
Arcadian Corp.
Arcadian Corp.
Ft. Madison
Ft. Madison
Ft. Madison
Hahnville
Geismar
Geismar
Geismar
Geismar
20 9
20 30
24 5
36 39
36 39
36 39
9 4$aj 62 43
9 6jaj 63 43
11 12$aj 63 43
14 12$aj 63 43
38 20(a' 63 43
1.3 C
1.3 C
1.3 C
2.3
2.3
2.3
2.3
2.3
0
O
I
I
I
U
U
U
420
160
160
160
160
Agrico
Agrico
Mississippi
Nu-South Industries
Missouri
W.R. Grace
W.R. Grace
Farmers Chemical Co.
Donaldsonville 203 12Ja} 60 43 2.2 0
Uncle Sam 284 20 64 42 2.4 O
Joplin
Joplin
Joplin
10
10
28
U
U
15
40
40
40
44
44
44
1.4
1.4
1.4
C
C
C
420
800
220
U
U
U
Area = Base area, hectares.
Height = Average height, meters.
RN = Rainfall rate, in/yr.
Evap. = Lake evaporation rate, in/yr.
b = Coefficient in R = exp(-bx), where R is ratio of the covered
to the uncovered radon flux and x is cover thickness, meters.
U = Unknown . .
Status = 0 is operating (active), I is idle (considered to be inactive
for the purpose of this analysis), and C is inactive (closed).
Cap. = Plant P2O5 production, thousands of metric tons/yr.
(a) Three have a 1:5 slope; one has a 1:3 slope; and two have a 1:8 slope.
The slope of one stack is unknown.
(b) This stack has a 1:10 slope.
13-31
-------
Table 13-16. Characteristics of gypsum stacks (continued).
State/Company location Area Height RN EVP b status Capacity
Texasgulf
Texasgulf
Texasgulf
Texasgulf
Texasgulf
Amoco
Amoco
Kerley Agri Chem
Phillips Chemical
Mobil
Mobil
Mcbil
Utah
Chevron Chemical
Wyoming
Chevron Chemical
Aurora
Aurora
Aurora
Aurora
Aurora
Texas City
Texas City
Pasadena
Pasadena
Pasadena
Pasadena
Pasadena
Magna
16
30
51
51
51
26
18
38
19
20
52
52
52
52
52
43
43
43
43
43
1.8
1.8
1.8
1.8
1.8
I
I
I
I
O
1150
1150
1150
1150
1150
2
14
11
14
24
36
61
3
11
11
27
27
27
30
52
52
48
48
48
48
48
46
46
46
46
46
46
46
1.8
1.8
1.7
1.7
1.7
1.7
1.7
I
I
C
I
C
C
0
u
u
u
u
220
220
220
121
Rock Springs 182 10
16 55 0.91 C 90
8 45 0.77 O 180
Area
Height
RN
Evap.
b
= Base area, hectares.
= Average height, meters.
= Rainfall rate, in/yr.
= Lake evaporation rate, in/yr.
= Coefficient in R = exp(-bx), where R is ratio of the covered
to the uncovered radon flux and x is cover thickness, meters.
= Unknown
Status = O is operating (active), I is idle (considered to be inactive
for the purpose of this analysis), and C is inactive (closed).
Cap. = Plant P2O5 production, thousands of metric tons/yr.
U
13-32
-------
Differing stack construction techniques divide the stacks
into three groups:
1. Louisiana and Mississippi, where the slopes of the sides
are more gentle (about 1:5 to 1:10) than in the rest of
the country;
2. North Carolina, where the slopes of the sides are
about 1:1.8; and
3. the rest of the country where the slopes are
generally about 1:3.
The radon flux and fraction of the top surface of the stack
that is covered by water separate the stacks into the following
three groups:
1. Idaho, Utah, and Wyoming, where the fraction of top
surface covered by water is about half that of other
regions and the radium content of the gypsum is somewhat
lower;
2. North Carolina and northern Florida where the radium
content of the gypsum is lower; and
3. all others.
By examining stacks in the following five groups, these
differences can be accounted for:
1. Florida (except northern Florida), Texas, and Arkansas;
2. Illinois, Iowa, and Missouri;
3. Louisiana and Mississippi;
4. North Carolina and northern Florida; and
5. Idaho, Utah, and Wyoming.
Table 13-17 contains the mean characteristics of the stacks
in each group.
The average radon flux, based on the entire stack surface
area, is given in Table 13-18 along with the total radon emis-
sions and the thicknesses of earth cover required to reduce the
average flux to 6 and 2 PCi/m2/s. See Appendix 13-E for the
method used to calculate the effect of earth cover.
The cost of reducing the average radon flux to 6 and
2 pCi/mVs is given in Table 13-19. There are no gypsum stacks
with an average flux greater than 20 pCi/mVs, so no mitigation
cost would be incurred to reach this level. The costs were
estimated using the unit costs and methods described in
13-33
-------
?7 Estimates include the cost of a drain system and
/iX?er 5°r the t0p- Note that in some cas^s less than
one foot (0.3 meters) of earth is required to reduce the average
™ ?™iUV° 6^r-2 Pci/*Vs. For these cases, a minimum of
The estimated costs of covering both active and inactive
stacks are based upon the mean surface area in each group calcu-
lated, using the method described in Appendix 13-E. Table 13-19
gives the cost by group of reducing the radon flux to 6 and
2 pci/m /s. The estimated cost for the entire population of
stacks is $0.5 billion if the radon flux were to be limited to
6 pCi/mVs, and $1 billion if it were to be limited to
2 pci/m /s. Continuing costs for cover and drain system main-
tenance are estimated to be about $10 million per year.
f Tfble 13-2° gives the estimated cancer deaths for each group
of states for the three cases of no action, reduction of the
radon flux to 6 pCi/m^/s, and reduction of the radon flux to
2 pCi/m /s. The estimated risk for these cases was obtained by
scaling the risk from Table 13-C-2 by the ratio of the total
emissions for each group in Table 13-18 to the total emissions
for each group from Table 13-7 (see Appendix 13-E) . The total
risk with no action is estimated to be one cancer death per year;
with radon controlled to 6 PCi/m2/s, it is 0.8 cancer deaths
per year; and with radon controlled to 2 pCi/m2/s it is 0 3
cancer deaths per year. •«»«..»
13-34
-------
Table 13-17. Mean characteristics of the stacks in each group.
Mean Base Average Average Estimated Total
Area Height Side Area Top Area Life Number
(ha) (m) (ha) (ha) (y)
Active Stacks
Florida, Texas,
& Arkansas 91
Illinois, Iowa,
& Missouri 63
Louisiana &
Mississippi 149
North Carolina
& N. Florida 100
Idaho, Utah,
& Wyoming 100
83 (22)
-------
Table 13-18. Radon emissions from grouped gypsum stacks.
Average Cover thickness (m) Total Radon
T . . Radon P-UX to Reduce Flux to Emissions Ci
location (pCi/mVs) 6 2 USr 6 2
(pci/m2/s) cover
Inactive Stacks
Florida, Texas,
& Arkansas
Illinois, Iowa,
& Missouri
Louisiana &
Mississippi
North Carolina
Idaho, Utah,
& Wyoming
Total Radon
Emissions, Ci/y
Active Stacks
Florida, Texas,
& Arkansas 8.4
Illinois, Iowa,
& Missouri 8.8
Louisiana &
Mississippi
North Carolina
& N. Florida
Idaho, Utah,
& Wyoming
8.6
2.6
13
8.3
7.5
7.8
1.3
3,500 2,500 840
360 250 82
0.30(0.20) 0.84
0.30(0.29) 1.1
0.30(0.16) 0.64 1,600 1,100 380
0 0.30(0.21) 190 190 120
°-94 2.2 1,300 590 200
0.30(0.19) 0.84
0.30(0.18) 1.0
0.30(0.11) 0.59
0 0
7.3 0.30(0.22) 1.5
1,200 660 220
520 420 140
160 120 40
65 65 65
380 310 100
9,300 6,200 2,100
* Values shown in parentheses are actual thicknesses needed to achieve
indicated reduction. For cost purposes, a minimum value of one foot
(0.3 meters) was used.
13-36
-------
Table 13-19,
Location
Cost of mitigation.
Cost. Millions of 1987 Dollars
Radon Flux of
6 pCi/m2/s
Radon Flux of
2 pCi/m2/s
Florida, Texas,
& Arkansas 180
Illinois, Iowa,
& Missouri 48
Louisiana and
Mississippi 64
North Carolina 0^a'
Idaho, Utah,
& Wyoming 160
Total U.S. 450
390
110
120
24
300
940
(a) Flux from North Carolina stacks is less than 6 without
mitigation.
Note: There are no stacks with an average radon flux greater
than 20 pCi/m2/s.
13-37
-------
Table 13-20. Risk of cancer death.
Location NO Action
Florida, Texas,
& Arkansas O.s(a)
Illinois, Iowa,
& Missouri 0.2
Louisiana &
Mississippi o.07
North Carolina o.Ol
Radon Flux
6 pCi/m2/s
0.5
O.l
0.05
0.01
Reduced to
2 pCi/mVs
0.2
0.5
0.02
0.01
Idaho, Utah,
& Wyoming
0.02
0.01
0.004
Total U.S.
0.8
(a) Estimate rounded to one significant figure
0.3
13-38
-------
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Wa88b Walker, R., Freeport Chemical Company, Uncle Sam, LA, oral
communication, July 1988.
Wi88 Winn, E.B., Jr., Texasgulf, inc., written communication to
S.K. Beal, SC&A, Inc., August 1988.
13-42
-------
APPENDIX 13-A
Assumed Slopes of Phosphogypsum Stack Sides
Used to Compute Surface Areas
13-A-l
-------
The only stack dimensions available for this assessment were
the heights and base areas. To compute the top and side areas
necessary for determining the radon-222 source terms, the slope
of the sides must be known. Selection of a slope for the stack
sides was based on observations, personal communications, the
literature(a), and particularly a consideration of a height-slope
combination that would result in a reasonable top surface area.
A value of 1:3 (0.333) was used for the slope of all stack
sides except for the following stacks:
Facility
Stack Assumed Slope
Districhem Inc., Helena, AR
Seminole Fertilizer Corp., Bartow, FL
Agrico Chemical Company,
Donaldsonville, LA
Arcadian Corp., Geismar, LA
Agrico Chemical Co., Hahnville, LA
Agrico Chemical Co., Uncle Sam, LA
Nu-South Industries , Inc . ,
Pascagoula, MS
Texasgulf Chemicals Co., Aurora, NC
2
3
4
1
1
1
1
2
3
4
5
1:2
1:2.5
1:5
1:5
1:5
1:5
1:5
1:8
1:10
1:1.7
1:2.2
1:2.1
1:2.2
1:1.6
(0.500)
(0.400)
(0.200)
(0.200)
(0.200)
(0.200)
(0.200)
(0.125)
(0.100)
(0.558)
(0.450)
(0.476)
(0.450)
(0.625)
(a) Beal, S.K. and Thompson, S., "Preliminary Assessment of
Cost and Effectiveness of Mitigating Radon Emissions From
Phosphogypsum Stacks," Prepared by S. Cohen and Associates,
Inc. for the U.S. EPA, Contract No. 68-02-4375, Work
Assignment No. 1-20, McLean, VA, June 1988.
13-A-2
-------
APPENDIX 13-B
Dimensions of Phosphogypsum Stacks Used to
Calculate Radon-222 Source Terms
13-B-l
-------
Facility/Location
Districhem Inc. , Helena, AR
Agrico Chemical Co. , Bartow FL
Royster Phosphate, Inc. , Palmetto FL
Brewster Phosphates, Bradley, FL
C.F. Industries, Inc., Plant City, FL
C.F. Industries, Inc., Bartow, FL
Conserv, Inc., Nichols, FL #i(°)
#2
Estech, Inc., Bartow, FL
Farmland Industries, Inc. , Bartow, FL
Gardinier, Inc., Tampa, FL
Seminole Pert. Corp. , Bartow, FL #1
#2
IMC OoLpoi.atd.oti, Mulberry, FL
Occidental 1ip*'ii'^1 On #1
Vff^l^A Oyw*i tviij L'l JL*\
flllACJC CrtJI. JJKJo r 1 • 92
(Suwannee River)
Occidental Chemical Co.,
White Springs, FL (Swift Creek)
Rcyster Co. , Mulberry, FL #1
#2
USS Agri-Chemicals, Inc. Bartow PL
J ^™ 1 !• • IH^M-^M^f +r*^+% 1 AJb^LU^^W* f ^J
USS Agri-Chemicals, Inc. ,
Ft. Meade, FL
Nu-West Industries, Inc. , Conda, ID
J.R. Slmplot Co., |i
Pocatello, ID |2
Bunker Hill Co. , Kellogg, ID #1
12
13
u CLLecua ui pnr
Length, m(a)
424.3
1,673.3
1,555.6
1,000.0
1,800.0
1,708.8
800.0
787.4
469.0
1,356.5
1,661.3
1,131.4
2,130.7
QQA A
894.4
894.4
1,029.6
774.6
600.0
632.5
1,104.5
848.5
583.1
1,272.8
200.0
316.2
632.5
xspnogypsum sta
Width, m(a)
212.1
836.7
\j*j\f • g
777.8
500.0
900.0
854.4
400.0
393.7
234.5
678.2
830.7
565.7
1,065.4
886.0
447.2
447 2
~™ ' • mt
514.8
387.3
300 0
••* w • v
316.2
552.3
424.3
291.5
636.4
100.0
158.1
316.2
icks.
Heigh-
23
PI
28
40
10
07
& /
20
54
6
27
24
22
Oft
18
18
18
23
24
-------
u>
I
u>
laDJLe AJ— D-J.. CB\,uan\x*i UJJUBIB
Facility/location
3.L.U1K9 01
mUL CU.V2U0 V*- f~*
Length, m(a)
Allied Chemical Co., E. St. Louis, IL 374.2
Beker Industries Corp., Marseilles, IL 600.0
Mobil Chemical Oo., Depue, IL 894.4
Northern Etetrocnemical Co., Morris, IL 748.3
OlinCorp., Joliet, IL #1 1,303.8
12 400.0
SBOO, Inc., Streator, IL
U.S. Industrial Chemicals Co.,
Tuscola, IL
Agrico Chemical Company,
Ft. Madison, IA
Agrico Chemical Oo. ,
Donaldsonville, IA
Arcadian Corp. , Geismar, IA
#1
#2
#3
W **
#1
#2
If™*
i4
H »
Agrico Chemical Co., Hahnville, IA
Agrico Chemical Oo. , Uncle Sam, LA
Nu-South Industries, Inc.,
Bascagoula, MS
Fanners Chemical Oo. , Joplin,
W.R. Grace & Oo. , Joplin, MD
Texasgulf Chemicals Co. ,
Aurora, NC
Amoco Oil Oo. ,
Texas City, TX
MD
#1
#2
Tf™*
#1
#2
44
45
If**
#1
42
447.2
800.0
632.5
632.5
692.8
2,016.0
871.8
529.2
469.0
424.3
424.3
2,383.3
1,421.3
748.3
447.2
447.2
565.7
774.6
1,010.0
1,010.0
1,010.0
529.2
200.0
Width, m(a>
187.1
300.0
447.2
374.2
651.9
200.0
223.6
400.0
316.2
316.2
346.4
1,008.0
435.9
264.6
234.5
212.1
212.1
1,191.6
710.6
374.2
223.6
223.6
282.8
387.3
505.0
505.0
505.0
264.6
100.0
Height, m
9
9
13
4
27
5
18
16
30
9
5
12
20
12
12
6
4
20
20
15
10 W
10 (**)
26
18
38
19
20
11
3
Top Area, ha(b)
4.262
13.432
30.141
25.365
55.937
6.290
3.921
21.402
6.163
15.168
20.971
168.365
23.748
5.917
3.996
5.541
6.614
179.837
31.722
18.709
6.335
6.335
9.278
21.345
29.373
38.925
41.719
9.199
1.492
Side Area, ha
2.888
4.816
10.389
2.779
30.630
1.803
6.407
11.172
14.585
5.093
3.191
35.538
15.025
8.246
7.141
3.527
2.433
104.967
69.622
9.795
3.863
3.863
7.796
9.491
23.959
13.247
10.951
5.063
0.535
-------
Table 13-fr-l. Estimated dimensions and areas of phc^ptogypsum stacks (continued, .
Polity/location length, m (a) widt*, m(a) H^t, m ^ ^ ta(b)
Kerley Agricultural Chemicals 469.0 234 5 11 ,«,
of Texas, inc. , Pasadena, IX U 6*791 4.435
MobU Mining & Minerals Div., #l(°) 692.8 345 4 57
: 2
'
i
Chevron Chemical co. , Reck Springs, W i^OT.'g gssis io(d) Ss.'SJ Jj^S
l^??L£a^^
36 ^-'2' 3' etc" refer to Afferent stacks at a facility.
value.
%*r
I
<>.
-------
APPENDIX 13-C
Lifetime Fatal Cancer Risks and Committed Fatal Cancers
DueT to Radon-222 Emissions from Phosphogypsum Stacks
13-C-l
-------
Ji canc?r risks and working-level exposures due
M o £°r the maximum exposed individual are given in
Table 13-01 for each phosphogypsum stack. These results were
used to generate Table 13-10 in Section 13.3.1. Table 13-C-2
shows the estimated committed fatal cancers per year within 80 km
^ ^ S*Cti°» ^'^
13-C-2
-------
13-0-1.
lifetime fatal cancer risks to nearby individuals caused by radon-222 emissions
from phosphogypsum stacks.
OJ
1
o
1
Facility/Location
Districhem Inc., Helena, AR
Agrico Chemical Co. , Bartow FL
Itoyster Phosphate, Inc., Palmetto FL
Brewster Phosphates, Bradley, FL
C.F. Industries, Inc. , Plant City, FL
C.F. Industries, Inc., Bartow, FL
Conserv, Inc., Nichols, FL #llb>
#2
Estech, Inc., Bartow, FL
Farmland Industries, Inc., Bartow, FL
Gardinier, Inc., Tampa, FL
Seminole Pert. Corp., Bartow, FL #1
#2
IMC Corporation, Mulberry, FL
Occidental Chemical Co. , #1
White Springs, FL #2
(Suwannee River)
Occidental Chemical Co. ,
White Springs, FL (Swift Creek)
Itoyster Co. , Mulberry, FL #1
Radon
Concentration
(PCV1)
3.6E-3
1.6E-3
1.8E-2
6.8E-3
1.5E-2
4.2E-3
2.3E-3
2.9E-3
5.8E-4
8.4E-3
7.7E-3
8.2E-3
1.6E-2
4.1E-3
8.7E-4
7.4E-4
9.2E-4
3.7E-3
** er*s_o
Exposure
(WL)
1.2E-5
7.2E-6
5.7E-5
2.2E-5
4.7E-5
1.5E-5
7.4E-6
9.1E-6
2.2E-6
2.8E-5
2.6E-5
2.6E-5
5.1E-5
1.7E-5
3.3E-6
2.8E-6
3.6E-6
1.2E-5
1 *"* £
MaYiTium Lifetime
Fatal Cancer Risk
to Individual
2E-5
1E-5
8E-5
3E-5
6E-5
2E-5
1E-5
1E-5
3E-6
4E-5
4E-5
4E-5
7E-5
2E-5
5E-6
4E-6
5E-6
2E-5
1T7-S
Distance (a)
(meters)
1,400
4,800
1,200
1,200
1*>/w\
,200
2,600
1,100
1,100
3,000
1cf\f\
,500
1^f\f\
,600
1,200
1*\/\f\
.200
4,000
2,800
3,000
3,200
1,000
1.000
*2
USS Agri-Chemicals, Inc., Bartow, FL
USS Agri-Chemicals, Inc., Ft. Meade, FL
Nu-West Industries, Inc., Conda, ID
J.R. Simplot Co.. #1
Focatello, ID
#2
2.9E-3
1.2E-2
1.3E-2
2.0E-3
6.1E-3
9.0E-6
3.6E-5
3.9E-5
6.5E-6
2.1E-5
1E-5
5E-5
5E-5
9E-6
3E-5
(a) Distance from the center of the stack.
(b) Numbers 1, 2, 3, etc., refer to different stacks at a facility.
900
1,000
900
1,200
2,000
-------
Table 13-O1.
H
UJ
O
I
from phosphogypsum stacks (continued) .
Radon
Facility/Location Concentration
(PCV1)
Bunker Hill Co., Kellogg, ID #l
-------
u>
O
Ul
_lCUJ-LTr A_T^_» _!_• rirTVi IMITIV*r*irl 1 I |««.-«"^ MMI_«» m^*=*tm^=^ -i_-L£W£=» •_•_* m.m^
from phosphogypsum stacks (continued) .
Radon
Facility/Location Ooncentration
(pCi/D
W.R. Grace & Co., Joplin, MO #lT ^«^^
-------
Table 13-C-2. Summary of committed fatal cancers per year within 80 km of
phosphogypsum stacks.
Districhem Inc. , Helena, AR
Agrico Chemical Co. , Bartow FL
Royster Phosphate, Inc. , Palmetto FL
Brewster Phosphates, Bradley, FL
C.F. Industries, Inc., Plant City, FL
C.F. Industries, Inc., Bartow, FL
Conserv, Inc., Nichols, FL #1
#2
Estech, Inc., Bartow, FL
Farmland Industries, Inc., Bartow, FL
Gardinier, Inc., Tampa, FL
Seminole Pert. Corp. Bartow, FL #1
#2
IMC Corporation, Mulberry, FL
Occidental Chemical Co., #1
White Springs, FL #2
349,261
1,717,059
2,059,168
1,809,809
2,153,710
1,698,291
2,162,868
2,183,813
1,585,674
1,582,493
2,189,940
1,548,237
1,448,342
2,147,892
214,674
217,985
8F-4
Ofi *m
2E— 2
fctl.J 4b
2E— 2
wJLf ft
6E-3
3E-2
3E-2
6E-3
7E-3
2E-3
ff^t *J
IE-2
5E-2
•XLJ m»
3E-5
•JU ft
3E-2
8E-4
•I'M "•
8E-4
Occidental Chemical Co., 228,859 iE-3
White Springs, FL (Swift Creek)
Royster Co., Mulberry, FL #1 1,734,734 6E-3
#2 1,780,345 4E-3
USS Agri-Chemicals, Inc., Bartow, FL 1,405,177 5E-3
USS Agri-Chemicals, Inc., Ft. Meade, FL 1,416,722 TE-S
Nu-West Industries, Inc., Conda, ID 97,600 3E-4
J.R. Simplot Co., #1 162,576 9E-4
Pocatello, ID #2 162,576 3E-3
Bunker Hill Co., Kellogg, ID #1 131,813 7E-5
#2 131,813 8E-5
#3 132,473 3E-4
Allied Chemical Co., E. St. Louis, IL 2,454,271 9E-3
Beker Industries Corp., Marseilles, IL 1,665,266 4E-3
Mobil Chemical Co., Depue, IL 675,690 3E-3
Northern Petrochemical Co., Morris, IL 6,100,385 iE-2
Olin Corp., Joliet, IL #1 7,448,591 1E-1
#2 7,458,031 9E-3
SECO, Inc., Streator, IL 801,552 2E-3
U.S. Industrial Chemicals Co., 640,239 2E-T
Tusoola, IL
Agrico Chemical Company, #1 335,158 9E-4
Ft. Madison, IA #2 335,623 5E-4
#3 335,334 5E-4
Agrico Chemical Co., Donaldsonville, IA 1,290,433 i£-2
Arcadian Corp., Geismar, IA #1 1,022,410 4E-3
#2 1,034,122 2E-3
#3 1,019,591 1E-3
#4 1,021,499 8E-4
13-C-6
-------
Table 13-0-2. Summary of ccramitted fatal cancers per year within 80 km of
phosphogypsum stacks (continued).
1980 Population Committed Fatal Cancers
Facility/Location within 80 km per Year (0-80 km)
Agrico Chemical Co. , Hahnville, LA
Agrioo Chemical Co., Uncle Sam, IA
Nu-South Industries, Inc.,
Pascagoula, MS
Fanners Chemical Co., Joplin, MO
W.R. Grace & Co., Joplin, MD #1
#2
Texasgulf Chemicals Co., #1
Aurora, NC #2
#3
#4
#5
Amoco Oil Co., #1
Texas City, TX #2
Rerley Agricultural Chemicals,
Pasadena, TX
Mobil Mining & Minerals Div. , #1
Pasadena, TX #2
#3
Phillips Chemical Co., Pasadena, TX
Chevron Chemical Co. , Magna, UT
Chevron Chemical Co. , Rock Springs, WY
1,783,951
1,909,222
671,827
361,128
358,231
356,982
374,248
372,327
382,084
382,559
380,246
2,621,365
2,620,216
2,986,765
2,992,382
2,999,952
3,002,031
2,985,632
1,147,033
41,108
9E-4
3E-2
7E-3
1E-3
4E-4
4E-4
2E-4
4E-4
7E-4
6E-4
6E-4
2E-2
2E-3
3E-2
9E-2
1E-1
1E-1
5E-2
4E-3
4E-4
13-C-7
-------
-------
APPENDIX 13-D
Frequency Distributions of Lifetime Fatal Cancers
Caused by Radon-222 Emissions from
Phosphogypsum Stacks in Selected Regions
13-D-l
-------
Table 13-D-l.
Risk Interval
Estimated distribution of lifetime fatal cancer
risk caused by radon-222 emissions from seven
phosphogypsum stacks in Texas.la'
Number of Persons^)
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
Totals(°)
0
0
0
0
350,000
8,400,000
12,000,000
20,000,000
Deaths/y
0
0
0
0
8E-2
3E-1
6E-2
4E-1
(a) Phosphogypsum stacks included in this summary are: Amoco Oil
Co., Texas City, TX (2); Kerley Agricultural Chemicals of
Texas, Inc., Pasadena, TX (l); Mobil Mining and Minerals Div
,™ «aSa?e?a' TX (3); PhilliPs Chemical Co., Pasadena, TX (1).
(b) Populations are overestimated because they have not been
corrected for overlap.
(c) Totals may not add due to independent rounding.
Table 13-D-2.
Risk Interval
Estimated distribution of lifetime fatal cancer
risk caused by radon-222 emissions from 10
phosphogypsum stacks in the Bartow, FL, region.
Number of Persons
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
0
0
0
0
7,400
2,600,000
15,000,000
1E-3
6E-2
7E-2
Totals
18,000,000
1E-1
(a) Phosphogypsum stacks included in this summary are: USS Agri-
Chemicals, Inc., Bartow, FL (l); Seminole Fertilizer Corp
Bartow, FL (2); Royster Co., Mulberry, FL (2); C.F.
Industries, Inc., Bartow, FL (1); Farmland Industries, Inc.
Bartow, FL (1); IMC Corp., Mulberry, FL (1); (1); Conserv
Inc., Nichols, FL (2).
(b) Populations are overestimated because they have not been
corrected for overlap.
(c) Totals may not add due to independent rounding.
13-D-2
-------
Table 13-D-3. Estimated distribution of lifetime fatal cancer
risk caused by radon-222 emissionsfrom six
phosphogypsum stacks in Illinois.
Risk Interval
Number of Persons( '
Deaths/y
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
IE-6 to 1E-5
< IE-6
Totals
(c)
0
0
0
0
20,000
2,200,000
22,000,000
24,000,000
0
0
0
0
5E-3
6E-2
8E-2
1E-1
Phosphogypsum stacks included in this summary are: .Beker
industries Corp., Marseilles, IL (1) ; Mobil Chemical Co
Depue, IL (1); Northern Petrochemical Co., Morris, IL(1) ,
Olin Corp. Joliet, IL (2); SECO, Inc., Streator, IL (1) .
Populations are overestimated because they have not been
corrected for overlap.
(c) Totals may not add due to independent rounding.
(b)
Table 13-D-4.
Risk Interval
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
IE-6 to 1E-5
< IE-6
Estimated distribution of lifetime fatal cancer
risk caused by radon-222 emissions from seven
phosphogypsum stacks in Louisiana.^
Totals
(c)
Number of Persons
0
0
0
0
10,000
740,000
8,200,000
9,000,000
(b)
Deaths/y
o
0
0
o
3E-3
2E-2
3E-2
5E-2
(a) Phosphogypsum stacks included in this summary are: Arcadian
(} Cor?? leismar, LA (4); Agrico Chemical GO.; Donaldsonvllle,
LA (1); Agrico Chemical Co., Hahnville, LA (1); Agrico
Chemical Co., Uncle Sam, LA (1).
(b) Populations are overestimated because they have not been
corrected for overlap.
(c) Totals may not add due to independent rounding.
13-D-3
-------
Table 13-D-5.
Risk Interval
1E-1 to 1E+0
1E-2 to 1E-1
1E-3 to 1E-2
1E-4 to 1E-3
1E-5 to 1E-4
1E-6 to 1E-5
< 1E-6
Totals
Estimated distribution of lifetime fatal cancer
risk caused by radon-222 emissions from three
phosphogypsum stacks in Idaho.Taj
Number of Persons
0
0
0
0
2,200
69,000
350,000
420,000
Deaths/y
0
0
0
0
4E-4
2E-3
2E-3
4E-3
(a)
sss sr:
=orecteaoov
(o) Totals may not add due to independent rounding.
13-D-4
-------
APPENDIX 13-E
Calculational Methods for Estimating
Costs of Reducing Radon from
Phosphogypsum Stacks
13-E-l
-------
"-E.I Stack Characteri
13-E.2 Maximum stack
minimum area of the top of a gypsum stack? The firs? ih
amount of gypsum that must beaccommodaSd during one stack
the second is the minimum
per year, gypsum is produced at a ate of 5 M pe yea or
Raising the entire stack 1.2 meters thus leads to a top
area
A = 0.274P(1/1.2)(1/0.72) = 0.32P m2
where 0.72 is the density of freshly added gypsum (45 Ib/ft3) .
13-E-2
-------
The other factor affecting the minimum size is the adequacy
of the pond to collect a substantial fraction of the gypsum.
perfectly mixed,
= (q + vA)C <2)
and A the settling area of the pond.
Defining R
A = q(l-R)/(vR) <3>
The incoming concentration is about 0.2 kg of gypsum per kg
h*.-, fe s asf s??u.^
water is
q = 5xlQ3p/3. 15x10? kg gyp/s x l/ .25 kg H2o/kg gyp x
1/1000 m3/kg H20
= 6.35xlO~7P m3/s
settling velocity of the gypsum particles is not known,
b
u yyyy iux. ci J^ACI*IW ^^v%^»*^*«-^ —- — , /o\
MT) of phosphoric acid per year. VB^_°*ffii™ ™*
R m o.OOOl, A = 12X104 m2 , and q = 6.35x10 XLOXJ.U
m3/s, v can be solved to obtain v = 0.079 m/s.
Thus,
A = 6.35X10-7P m3/S (1-0.0001)/(0.079 m/S X 0.0001)
= 0.08P m2 * '
Because there must always be at least two settling ponds, the
area must be twice this, or,
A = 0.16P m2
with P in MT/year.
Because the coefficient for P in equation (6) is less than
that in equation (1), equation (1) should be used.
13-E-3
-------
13-E.3 Estimate of
With the various "don emission rates
» C W6re estimated *>y scaling the total risk of all
stacks in each group from Table 13-C-2 by the ratio of the total
emissions for each group in Table 13-18 to the total of the
emissions from Table 13-7 for all the stacks within a group For
example from Table 13-E-l, 0.60 x (4700/3500) = o?80, w£i?h is
" and
Table 13-E-l. Values used to scale risk.
Group
Emissions From
Table 13-7
(Ci/y)
Emissions From
Table 13-18 fCi/y)
Avg. Flux Flux=6 Flux=2
FL,
IL,
LA,
NC
ID,
TX,
IA,
MS
UT,
AK
MO
WY
3500
750
1400
90
680
4700
880
1800
260
1700
3200
670
1200
260
900
1100
220
420
190
300
Risk from
Table 13-C-2
0.60
0.16
0.056
0.005
0.009
13~E'4 Calculation of Costs of Covering a Gvpsum st-a^v
The complete basis for the costs is given in Appendix B
"Generic Unit Costs for Earth-Cover-Based Radon-222 Control
Tenhrn rmo«s it t.rVi-i ^-.v, f~-~ 4-v._ j. _^—4. , . „ >-W*IWAWJ.
part uses data from Me87a and
The cost of earth was taken to be $2.62 per cubic meter.
Hauling (10-mile round trip) costs $11.64 per cubic mete? Plac-
ing grading, and compacting cost $5.52 per cubic meter. This
cost for placing, grading, and compacting tends to be on the
higher end of the Me87a data to account for working on a 1-3
06 n0t ad3usted here for the relatively few stacks
er (N°rth Carolina) or Antler (Louisiana and
slopes. Seeding costs were taken to be $0.54 per
Where seeding may not be practical (Idaho, Utah
and Wyoming) , it was assumed that the top was covered with
0.5 meters of gravel ($9.88 per cubic meter) and the sides with
?;Lm±£S °J ri?£aP- ($3°-5° per cubic mete^> • Mobilization costs
were n^lncL^™* ^^ th& ™* f°rCS and equipment)
square meter.
i-h^i- +** C°St °f theustack drai" system was estimated by assuming
that there are peripheral drains every 10 meters of height that
run completely around the stack. The bottom peripheral drain is
not counted as this is a normal part of every stack, whether it
13-E-4
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is covered or not. Downspouts were assumed to be located every
by $24, the cost of 1 meter of 10-inch pipe.
A drain system may not be needed if «»
onver- is sufficiently permeable to act as a drain
Became increased permeability acts to reduce the radon mitiga-
tion effectiveness, more earth would be required if an inner
liver of more permeable earth were used in conjunction with an
outSr Jayer of ?ess permeable earth. Thus, if this technique
were used the cost of earth would be higher, but there would be
no coslrorthe drain system. It is assumed that the increase
cancels out the reduction.
Also included is a synthetic cover for the top at an
iof of^sfacf *SiX Tot ™
S^nan
f S S?,rST2*.£S
S^
or Se costs of covering each of the active and inactive stacks
within each geographical group are given in Table 13-E-2.
13-E.5 Effectiveness of Farth Cover
The ratio of the radon flux (pCi/m2/s) from a covered
surface to that from an uncovered surface is given by:
R = exp(-bx) (?)
where R is the ratio, b is a coefficient, and x is the cover
thickness (NRC84).
The coefficient b is a function of the moisture Content of
the soil. This, in turn is estimated empirically from the bulk
density of the soil, the true density of the soil particles, the
rainfall rate, the lake evaporation rate, and the fraction of the
Soil that consists of particles that will pass through a 200-mesh
screen.
13-E-5
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Table 13-E-2. cost breakdown.
S"?o«?1i1J?M Cost' Minions
C=»P=nent gf^7 "gl£2* %£? °g^_™
Florida. Texasr and Arkansas
Active Stacks, N=14 Inactive Stacks, N=ll
Cost of Earth o 9 24
Hauling 38 ii'J °'3 °'8
Place, Grade, Compact i.'s 5.*1 J'7 ?*?
Seeding 06 n f. 1*7
Drain System J'Q ?'o n'? °'2
Synthetic Cover 1.5 £.5 °'J °'J
•!•••> O. . 5
Total 9 6 51 fi
y.b 21.6 4.3 7>9
Added costs of cover and drain maintenance - 0.3/year (active)
Illinois. TO^. andMisso,^ 0.06/year (inactive)
Active Stacks, N=2 Inactive Stacks, N=12
Cost of Earth 0.6 22 no
Hauling 26 96 n'? °'5
Place, Grade, Compact 1.3 4.*6 cC\ ?'J
Seedin^ 0.4 o.4 01 i"^
Drain System 13 ± J ?'J °-1
Synthetic Cover 0.4 o.4 J.'J J.'g
T°tal 6-6 18.3 2.9 ~
Added costs of cover and drain maintenance - 0.2/year (active)
Louisiana anH in..^..^ 0.05/year (inactive)
Active Stacks, N=4 Inactive Stacks, N=3
Cost of Earth 1.4 3.0
Haulin9 6.1 13 o S*8 ?*7
Dl ar^A /^v.->^« « j_ __ u-0 1.7
Place, Grade, Compact 2."9 6.'2 0.4 0.8
.0
- •* — w • w V • 8
Synthetic Cover i.6 i.6 1>4 £;~
r — ^ •—- •—»-™- ^ w^^Mifcsw*^^ ^ Amy
Seeding in in ~ ,
Drain lyste,. Jis ois Sil S.'J
Total 13 a oc fi „ «
J-J.O 25.6 3.0 4.5
Added costs of cover and drain maintenance - 0.5/year (active)
0.06/year (inactive)
13-E-6
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Table 13-E-2. Cost breakdown (continued)
Component
Cost, Millions
of 1987 Dollars
Flux=6 Flux=2
Cost, Millions
of 1987 Dollars
FlUX=6 FlUX=2
Carolina
Active Stacks, N=4
Inactive Stacks, N=4
Cost of Earth
Hauling
Place, Grade, Compact
Seeding
Drain System
Synthetic Cover
0
0
0
0
0
0
0.4
1.9
0.9
0.3
0.6
1.9
0
0
0
0
0
0
0
0
0
0
0
0
Total ° 6-°
Added costs of cover and drain maintenance
0.2/year (active^
Active Stacks, N=3 Inactive Stacks, N=5
Cost of Earth
Hauling
Place, Grade, Compact
Seeding
Drain System
Synthetic Cover
Total
Added costs of cover
3.0
13.0
6.2
15.0
2.3
1.0
40.5
and drain
6.9
31.0
15.0
15.0
2.3
1.0
71.2
maintenance - 0.3
O.C
0.3
1.3
0.6
2.1
0.1
3.6
8.0
/year (
19/year
1.5
6.7
3.2
2.1
0.1
3.6
17.2
active)
(inactive)
13-E-7
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The coefficient, b (cm/s) , is given by:
b = (L/D)V2
D = 0.07 exp[-4( m - mp2 + m
» = O.OIM / ( I/R -
and M by
M = 3.1PV2 - Q.03E + 3.9fcm - 1.0
13-E-8
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13-E.6 References
of
Be88b Beal, S.K. and Thompson, S., 'P^1^111^/3
Cost and Effectiveness of Mitigating Radon
From Phosphogypsum Stacks." Prepared by S. Cohen and
Associates, Inc. for the U.S. EPA, Contract No.
68-02-4375, Work Assignment No. 1-20, McLean, VA, June
1988.
Ca88 Cameron, J.E., "Land Planning for Phosphogypsum Stacks in
Central Florida," draft of a paper intended for publica-
tion in 1988.
Me87a Means, R.S., Inc., Boston, MA, Assemblies Cost Data, 12th
Ed., 1987.
Me87b Means, R.S., Inc., Boston, MA, Facilities Cost Data, 12th
Ed., 1987.
NOA82a National Oceanic and Atmospheric Administration "Evapora-
tion Atlas for the Contiguous 48 United States," NOAA
Technical Report NWS 33, U.S. ^P^erit of Commerce,
National Oceanic and Atmospheric Administration, National
Weather Service, Washington, DC, June 1982.
NOA82b National Oceanic and Atmospheric Administration, "Monthly
Normals of Temperature, Precipitation, and Heating and
Cooling Degree Days 1951- 1980," Climatography of the
United States No. 81 (by state), U.S. Department of Com-
merce, National Oceanic and Atmospheric Administration,
Environmental Data and Information Service, National
Climatic Center, Asheville, NC, September 1982. Note:
there is one publication of this number for each state.
NRC84 Rogers, V.C., Nielson, K.K., and Kalkwarf, D.R., "Radon
Attenuation Handbook for Uranium Mill Tailings Cover
Design," NUREG/CR-3533, prepared for the U.S. Nuclear
Regulatory Commission, Washington, DC, April 1984.
Sp88 Spivey, L., U.S. Department of Agriculture, personal
communication with SC&A personnel, August 1988.
M.S. Government Printing Office: 1990-718-808 13-E~9
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