United States
nvironmental Protection
Agency
Office of
Reseach and
Development
Environmental Monitoring
and Support Laboratory
Las Vegas, Nevada 89114
EPA-600/7-77-082
August 1977
POTENTIAL RADIOACTIVE
POLLUTANTS RESULTING FROM
EXPANDED ENERGY PROGRAMS
Interagency
Energy-Environment
Research and Development
Program Report
-------
RESEARCH REPORTING SERIES
Research reports of the Office of Research and Development, U.S. Environ-
mental Protection Agency, have been grouped into seven series. These seven
broad categories were established to facilitate further development and
application of environmental technology. Elimination of traditional grouping
was consciously planned to foster technology transfer and a maximum
interface in related fields. The seven series are:
1. Environmental Health Effects Research
2. Environmental Protection Technology
3. Ecological Research
4. Environmental Monitoring
5. Socioeconomic Environmental Studies
6. Scientific and Technical Assessment Reports (STAR)
7. Interagency Energy-Environment Research and Development
This report has been assigned to the INTERAGENCY ENERGY-
ENVIRONMENT RESEARCH AND DEVELOPMENT series. Reports in this
series result from the effort funded under the 17-agency Federal
Energy/Environment Research and Development Program. These studies
relate to EPA's mission to protect the public health and welfare from adverse
effects of pollutants associated with energy systems. The goal of the Program
is to assure the rapid development of domestic energy supplies in an
environmentally compatible manner by providing the necessary environ-
mental data and control technology. Investigations include analysis of the
transport of energy-related pollutants and their health and ecological effects:
assessments of, and development of, control technologies for energy systems;
and integrated assessments of a wide range of energy-related environmental issues.
This document is available to the public through the National Technical Infor-
mation Service, Springfield, Virginia 22161.
-------
EPA-600/7-77-082
August 1977
POTENTIAL RADIOACTIVE POLLUTANTS
RESULTING FROM EXPANDED ENERGY PROGRAMS
by
*
Hong Lee, Thomas 0. Peyton,
Robert V. Steele, and Ronald K. White
Center for Resource and Environmental Systems Studies
Stanford Research Institute
Menlo Park, California 94025
*
Greenfield Attaway & Tyler, Inc.
San Rafael, California 94901
Contract No. 68-03-2375
Project Officer
Arthur N. Jarvis
Monitoring Systems Research and Development Division
Environmental Monitoring and Support Laboratory
Las Vegas, Nevada 89114
ENVIRONMENTAL MONITORING AND SUPPORT LABORATORY
OFFICE OF RESEARCH AND DEVELOPMENT
U.S. ENVIRONMENTAL PROTECTION AGENCY
LAS VEGAS, NEVADA 89114
-------
DISCLAIMER
This report has been reviewed by the Environmental Monitoring and Support
Laboratory—Las Vegas, U.S. Environmental Protection Agency, and approved for
publication. Approval does not signify that the contents necessarily reflect
the views and policies of the U.S. Environmental Protection Agency, nor does
mention of trade names or commercial products constitute endorsement or
recommendation for use.
ii
-------
FOREWORD
Protection of the environment requires effective regulatory actions which
are based on sound technical and scientific information. This information must
include the quantitative description and linking of pollutant sources, trans-
port mechanisms, interactions, and resulting effects on man and his environment.
Because of the complexities involved, assessment of specific pollutants in the
environment requires a total systems approach which transcends the media of air,
water, and land. The Environmental Monitoring and Support Laboratory-Las Vegas
contributes to the formation and enhancement of a sound integrated monitoring
data base through multidisciplinary, multimedia programs designed to:
develop and optimize systems and strategies for monitoring
pollutants and their impact on the environment
demonstrate new monitoring systems and technologies by applying
them to fulfill special monitoring needs of the Agency's operating
programs.
This report, "Potential Radioactive Pollutants Resulting from Expanded
Energy Programs," will assist the EPA's quality assurance program in directing
its efforts to meet the needs of laboratories engaged in monitoring radioactive
pollutants or potential pollutants resulting from the expanded energy programs.
The data contained in this report should be of value to those agencies involved
in standard setting for energy development activities and for those monitoring
the pollutants associated with these developments. For further information
on this subject, contact the Quality Assurance Branch, Environmental Monitoring
and Support Laboratory-Las Vegas, Nevada.
fge/B. Morj
Director
Environmental Monitoring and Support Laboratory
Las Vegas
iii
-------
CONTENTS
Foreword
Figures v*
Tables vii
Summary ^-x
I INTRODUCTION 1
II ALTERNATIVE EXPANDED ENERGY PROGRAMS 3
A. Background 3
B. U.S. Energy Supply Scenarios 4
C. Summary of Scenarios 7
III RADIATION EXPOSURES 11
A. Radionuclides and Sources 11
B. Exposure Pathways and Health Effects 13
1. Exposure Pathways 13
2. Liquid Effluent Releases and Dose
Computation Methodology 15
3. Atmospheric Emissions and Computation
Methodology 18
4. Local Dose 22
C. Human Exposure Calculation 28
1. Water Dose 31
2. Atmospheric Dose 31
3. Air-Food Dose 34
IV ENERGY SYSTEMS 38
A. Coal for Direct Combustion 38
1. Coal Characteristics and Locations 38
2. Mining Methods 41
3. Coal Preparation 43
4. Coal Transportation 43
5. Electrical Power Generation 44
6. People 45
7. Radiological Aspects 45
B. Coal Gasification and Liquefaction 55
1. Coal Gasification Technologies 56
2. Coal-Liquefaction Technologies 56
3. People 58
4. Radiological Aspects 58
C. Oil Shale 65
1. Mining 65
2. Conversion 65
3. People 68
4. Radiological Aspects 68
iv
-------
IV ENERGY SYSTEMS (continued)
D. Geothermal Energy Systems 71
1. Introduction 71
2. Exploration and Development 72
3. Production 73
4. Radiological Aspects 74
E. Nuclear Systems 75
1. Uranium Mining 76
2. Uranium Milling 77
3. Conversion to Uranium Hexafluoride 80
4. Uranium Enrichment 80
5. Fuel Fabrication 81
6. Reactor Operations 81
7. Spent Fuel Reprocessing 85
V RADIOLOGICAL PROJECTIONS 89
A. Energy Scenario II Emissions and Effects .... 89
1. Coal 89
2. Oil Shale 92
3. Geothermal 94
4. Nuclear Systems 94
B. Energy Scenario III Emissions and Effects .... 100
1. Coal 100
2. Oil Shale 101
3. Geothermal 101
4. Nuclear Systems 101
C. Energy Scenario IV Emissions and Effects .... 105
1. Coal 105
2. Oil Shale 108
3. Geothermal 109
4. Nuclear Systems 109
D. Energy Scenario V Emissions and Effects 113
1. Coal 113
2. Oil Shale 116
3. Geothermal 116
4. Nuclear Systems 117
VI DISCUSSION OF RESULTS 121
References 126
-------
FIGURES
Number Page
1 ERDA-48 Scenario II, Synthetics From Coal and Shale 5
2 ERDA-48 Scenario III, Intensive Electrication 6
3 ERDA-48 Scenario IV, Limited Nuclear Power 8
4 ERDA-48 Scenario V, Combination of All Technologies 9
vi
-------
TABLES
Number Page
1 Levels of Energy Resource Use in ERDA-48 Scenarios 10
2 Natural Light and Heavy Source Radionuclides that
Occur Significantly in Nature 12
3 Reactor Fission and Activation Products 14
4 Modes of Human Exposure 15
5 Principal Exposure Pathways for Radiation
Exposure from Nuclear Reactor Effluents 16
6 Summary of Dose Equivalent Rates (MRem/Year) from Various
Radionuclides Composing the Natural Background Radioactivity
in the United States for External (E), Airborne (A), and
Internal (I) Exposures 21
7 Macroregional Exposure Computation Variables 23
8 Proposed Rules (FR23421, 1975) 28
9 Maximum Permissible Concentration Formulas 29
10 Maximum Permissible Concentration 30
11 Water Transport Dose Variables and Q-Dose Algorithms 32
12 Macroregional Dose Algorithms 33
13 Local Atmospheric Exposure Algorithms 35
14 Fallout Dose Algorithms 37
15 Uranium and Thorium Content of Various Coals 40
16 Potassium-40 Content of Various Coals 41
17 Radon-222 and Radon-220 Concentrations in Various Coals 46
18 Rate of Release of Radon from Coal Mining in Support
of a 1000-MWe Power Plant 47
vii
-------
Number
Page
19 Particulate Emissions From a 1000-MWe Coal-Fired Power Plant ... 50
20 Release of Radon Isotopes From a 1000-MWe Power Plant 50
21 Emission of Radionuclides in Particulate Matter From a
1000-MWe Power Plant: No Elemental Concentration in
Fly Ash Assumed 52
22 Emission of Radionuclides in Particulate Matter From a
1000-MWe Power Plant: Concentration of Uranium, Lead,
and Polonium in Fly Ash Assumed 54
23 Maximum Radon-222 Release From 30-Year Ash Storage
Pile from 1000-MWe Power Plant 55
24 Radionuclide Emissions to the Air From a 275 Million
Standard Cubic Feet per Day Coal Gasification
Facility Using Navajo Coal 61
25 Radionuclide Emissions to the Air From a 100,000 Bbl/Day
Coal-Liquefaction Facility Using Powder River Coal 63
26 Radionuclide Emissions to the Air From a 100,000 Bbl/Day
Coal-Liquefaction Facility Using Illinois Coal 64
27 Radionuclide Emissions to the Air From a 100,000 Bbl/Day
Oil Shale Mining, Retorting, and Upgrading Operation 70
28 Representative Quantities of Radioactive Materials
in Spent Fuel per 1000 MWe-Year Uranium Reprocessing
Requirement 87
viii
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SUMMARY
The radionuclide releases and the resulting population exposure doses from
several energy systems for four projected energy utilization scenarios were cal-
culated and compared. The energy system components examined were: coal mining,
processing, combustion, and ash disposal; coal gasification and liquefaction;
oil shale mining, processing, residue disposal and product utilization; geo-
thermal development and operations; uranium mining, milling, conversion, enrich-
ment and fabrication; nuclear reactor operations; and fuel reprocessing and
waste disposal. The energy utilization scenarios included one that projected a
high level of synthetic fuel production, one that projected high electrical
power utilization—mostly derived from nuclear power developments beyond 1985,
and one that assumes moderate developments in all energy systems along with
energy conservation measures.
For nonnuclear energy systems, the dominant radionuclide contributing to
population exposure doses is radon-222. The major sources of radon-222 releases
are the coal ash piles created from the combustion of coal to produce electric-
ity. On a per unit energy basis, the geothermal energy system is a comparable
contributor to radon-222 exposure doses. Exposure doses derived from other
radionuclides and from other nonnuclear energy systems are comparatively in-
significant.
For nuclear energy systems, the radionuclides contributing to population
exposure doses are ranked as follows: 1) tritium, 2) uranium-238, 3) radon-222,
4) krypton-85, 5) ruthenium-106, 6) xenon-133, 7) iodine-131 and iodine-129,
8) actinides, and 9) other fission and activation products. The sources of
tritium population exposure doses are tritium releases to the environment from
spent fuel reprocessing operations and from nuclear reactor operations. The
uranium-238 releases, mostly during milling operations, also contribute to rela-
tively significant population exposure doses. The population exposure doses
derived from radon-222 releases from uranium mine and mill sites, however, are
about the same order of magnitude as that derived from the release of radon-222
from coal ash piles.
There were wide differences in the data, both estimated and measured, on
radionuclide release rates from nuclear energy system component operations and
facilities, and the data on radioactive material releases from nonnuclear energy
systems operations and facilities are extremely sparse and fragmented. Without
more and better data, only order of magnitude estimates could be made on the
effects of radionuclide releases from the various energy systems. Many assump-
tions, especially concerning the exposed population size, were necessary for the
comparisons. These assumptions should be carefully considered in assessing the
results.
ix
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I INTRODUCTION
The Quality Assurance Branch of the U.S. Environmental Protection Agency's
Environmental Monitoring and Support Laboratory-Las Vegas is responsible for
providing an effective radiological quality assurance program for the Agency's
energy program. A prerequisite for such a program is the identification and
documentation of the important potential radioactive pollutants that could re-
sult from a program of energy expansion. Although proposed programs to develop
domestic energy sources include the development of a variety of sources, near-
term interest has focused on the development of geological resources such as
the mining and use of western coal, mining and processing of oil shale, expan-
sion of nuclear energy programs, and development of geothermal energy. The
development of these geological resources—coal, oil shale, uranium, and geo-
thermal energy—will increase the release of radioactivity to the environment.
For example, open pit mining will continuously expose new layers of radio-
activity-bearing materials to the elements, processing and use of the minerals
will inject the radioactive contents into the environment, and fissioning of
nuclear fuels will generate massive quantities of radioactive fission products.
In the past, the lack of understanding of the hazards of radioactive materials
has resulted in tragic and costly errors in the handling, use, and disposal of
radioactive materials. Examples are the inhalation of radon by uranium miners,
the ingestion of radium by watch dial painters, and the creation of hazardous
uranium mill tailings piles.
It is the recognition of these and other past errors that prompted the cur-
rent work to identify, analyze, and document the important potential radioactive
pollutants resulting from a program of energy expansion so that measures can be
taken to forestall possible unsafe radiological practices and unacceptable radio-
logical contamination of the environment that might otherwise evolve. The re-
sults of the study must be considered preliminary, however. In the absence of
detailed development plans, many assumptions concerning exposure pathways and
exposed populations were necessary.
The radiological aspects of several energy systems are reported here.
They include: coal mining, processing, combustion, and ash disposal; coal gas-
ification and liquefaction; oil shale mining, processing, residue disposal, and
product utilization; geothermal development and operations; and nuclear systems.
The nuclear system components and operations covered include uranium mining and
milling; uranium conversion, enrichment, and fuel fabrication; light water, high-
temperature gas, and breeder reactors; and fuel reprocessing and waste disposal.
Because high-temperature gas reactors and breeder reactors are still in the de-
velopment stage and high level nuclear waste management remains unresolved,
coverage of these subject areas is cursory.
-------
In this report, proposed alternatives to expanded energy programs to meet
projected U.S. energy demands through the end of the century are discussed,
and four of six energy scenarios were selected for assessing the radiological
effects of expanded energy development. This is followed by discussions on
radionuclides of possible concern, their sources, means of release to the en-
vironment, pathways to human exposures, and the calculative methods used to
convert projected release rates to population exposure doses for the purpose
of ranking the relative importance of the radionuclide releases.
Next, the energy systems, resources, and system components are discussed,
along with the radiological aspects associated with the energy systems, re-
sources, and system components. Basic radionuclide content and release data
are also presented. Using these basic release data and the projected energy
utilizations by type, the radionuclide releases for 1985 and 2000 are then cal-
culated and presented. The presentation of these releases in terms of popula-
tion exposure doses follows. The final section is a discussion of the results.
-------
II ALTERNATIVE EXPANDED ENERGY PROGRAMS
A. Background
Subsequent to the Arab oil embargo of 1973-74, the United States began to
reevaluate its energy policies with the goals of reducing the impact of pos-
sible future embargoes in the near term and achieving a state of energy self-
sufficiency in the long term. The main thrust of the programs formulated to
achieve the goal of energy self-sufficiency was a significant expansion in the
rate of exploitation of domestic energy resources to meet an energy demand
that was expected to double by the year 2000. In addition, it was realized
that the rate of growth in energy demand could be moderated if waste in energy
consumption were reduced and some energy conservation efforts were begun.
To meet projected U.S. energy demands through the end of the century using
domestic energy resources, it will be necessary to rely heavily on those re-
sources with large reserves that have been relatively underused in the past.
Although reliance on traditional sources of natural gas and crude oil, which
together contribute over 75% of current energy supplies, will continue to be
heavy, the production of these resources peaked in the early 1970s and will
probably continue to decline to the end of the century. Some temporary in-
creases in production will occur as Alaskan oil and other newly discovered re-
serves are brought into production, but these will serve mainly to offset the
declining production from existing fields. Thus, it will be necessary to turn
to the more abundant reserves of coal, oil shale, and uranium to meet growing
energy demand. These sources will be supplemented by geothermal, solar, and
other renewable energy sources.
The greatly expanded production of these energy sources will have wide-
spread economic, social, and environmental impacts. The assessment of these
impacts, including the effects of increased emissions of radionuclides, requires
some quantitative estimate of the rate at which each energy resource will be ex-
ploited. Although no one can predict the future with certainty, there have been
a number of projections of energy supply levels out to the year 2000. These
projections are based on past trends, economics of supply and demand, likely
technological developments, and quantities of energy resources in place. These
projections can be used as a basis to quantify the possible impacts of future
energy development.
-------
B. U.S. Energy Supply Scenarios
A number of different groups have made energy supply and demand projec-
tions for the years 1985 to 2000. Each of these projections is based on an
explicit set of assumptions about energy economics and technology, and each
involves many uncertainties about the future energy picture, both domestic
and international. For the purposes of the analysis to be done here, it is
desirable to have a set of scenarios that bracket the nation's possible energy
futures. Such a set should consider the range of technological options that
may become available, and the effect of the various energy policy options that
the nation may decide to carry out. The set of scenarios proposed by the U.S.
Energy Research and Development Administration (ERDA) in its document ERDA-48,
A National Plan for Energy Research, Development and Demonstration: Creating
Energy Choices for the Future, adequately fulfills these criteria. [Ref. 1]
The six energy scenarios developed in ERDA-48 provide a wide range of pos-
sible energy futures, emphasizing different technology options. Two of the
scenarios—Scenario 0, No New Initiatives, and Scenario I, Improved Efficiencies
in End Use—were judged to be of limited interest in terms of assessing the
radiological effects of expanded energy development. The other four scenarios
are discussed briefly below.
Scenario II—Synthetics from Coal and Shale
The dominant feature of this scenario is a high level of production of
synthetic fuels from coal and oil shale, and its goal is a reduction in the
high levels of imported petroleum that would otherwise be required to sustain
a growing demand for liquid fuels. Additional energy supplies are provided
through the enhanced recovery of oil and gas, and through nuclear power, con-
tinuing to grow at something like its historical rate. In this scenario, the
contribution of geothermal energy remains relatively small. The projected
growth of various energy supplies as a function of time is shown in Figure 1.
Scenario III—Intensive Electrification
In this scenario the emphasis is on the maximum utilization of electric
power. The widespread introduction of electric cars is assumed along with
increased efficiency in other uses of electricity. On the production side, the
use of coal for electric power production is favored over conversion to syn-
thetics. Greater use of nuclear power, including breeders, is projected, com-
bined with faster development of geothermal, solar, and other alternative
sources. This scenario is depicted in Figure 2.
Scenario IV—Limited Nuclear Power
In this scenario, it is assumed that for one reason or another (such as a
nationwide nuclear moratorium) the level of nuclear power development is limited
to those plants already built or on order. The difference between this low
-------
1975
SOLAR AND
OTHER
GEOTHERMAL
>NUCLEAR
HYDRO
> COAL
NATURAL
GAS
SHALE OIL
CRUDE OIL
FIGURE I. ERDA-48 SCENARIO II, SYNTHETICS
FROM COAL AND SHALE
-------
SOLAR AND
OTHER
GEOTHERMAL
NUCLEAR
HYDRO
COAL
%*% NATURAL
GAS
1980
1965
1990
1995
2000
YEAR
FIGURE 2. ERDA-48 SCENARIO HI, INTENSIVE
ELECTRIFICATION
-------
level of nuclear power and the higher level projected in other scenarios is
largely made up by accelerating geothermal and solar development at a rate even
greater than in Scenario III. The total level of coal production is about the
same as for Scenario III, including the same levels of synthetic fuel produc-
tion. This scenario is depicted in Figure 3.
Scenario V—Combination of All Technologies
This scenario assumes a moderately high level of development of all the
technologies considered in Scenarios II through IV, combined with an end-use
conservation strategy that reduces energy demand by about 8% in 1985 and 13% in
2000. The result of this scenario is an elimination of petroleum imports by the
end of the century. The main effect of end-use conservation is a reduction in
demand for electricity that results in the use of coal and nuclear fuel for elec-
tricity production being reduced significantly below the levels of Scenarios II
and III. This scenario is depicted in Figure 4.
C. Summary of Scenarios
The ERDA-48 Scenarios II through V are summarized in Table 1 and compared
with 1975 data for the categories of energy supply considered in this study.
Resource energy in the table is given in units of quads—1015 British thermal
units (Btu).* As can be seen from the table, the four scenarios delineate a
wide range of future resource consumption levels.
In the year 2000, the use of coal ranges over a factor of 1.6, geothermal
over a factor of 10, and nuclear over a factor of 4. The use of coal and oil
shale resources for the production of synthetic fuels ranges from 0 to 29 quads
in the year 2000. In the nearer term (1985) the levels of resource use are not
nearly as sensitive to the assumptions of the various scenarios, except in the
case of geothermal which varies by a factor of 5 between Scenario II and Sce-
nario IV.
All the scenarios show significant future increases in the use of coal,
nuclear power, and geothermal energy compared with present levels, and signifi-
cant levels of production of synthetic fuels (except Scenario III) compared to
total gaseous and liquid-fuel use. These scenarios, thus, provide good frames
of reference within which to assess the impact of radiological pollutants from
an expanded energy program.
*0ne Btu equals 1055 joules, so a quad is approximately 1018 joules,
-------
CD
160
ISO
140
130
120
110
100
90
I
CO
o
80
0 70
60
50
40
30
20
10
1975
GEOTHERMAL
NUCLEAR
HYDRO
COAL
NATURAL
GAS
CRUDE OIL
I960
1985 1990
YEAR
1995
2000
FIGURE 3. ERDA-48 SCENARIO IV, LIMITED NUCLEAR POWER
8
-------
INDUSTRIAL
""GASIFICATION
GEOTHERMAL
>NUCLEAR
HYDRO
COAL
NATURAL
GAS
2000
FIGURE 4. ERDA-48 SCENARIO V, COMBINATION
OF ALL TECHNOLOGIES
-------
TABLE 1. LEVELS OF ENERGY RESOURCE USE IN ERDA-48 SCENARIOS
(1015 Btu/year)
Scenarios
II
Resource
Coal
Strip
Underground
Cleaned
Liquefied
Gasified
Electricity
Coke
Heat
Chemical
Oil shale
Geo thermal
Electricity
Heat
Nuclear
LWR
HTGR
LMFBR
1985
23.28
11.64
11.64
19.64
0.73
1.42
12.51
2.81
3.49
0.24
1.00
0.69
0.69
—
10.85
10.61
0.24
2000
49.77
20.90
28.87
27.06
15.91
5.30
15.21
3.54
4.90
2.60
8.00
1.40
1.40
—
40.49
36.59
3.90
III
1985
20.10
10.05
10.05
18.60
-
-
12.51
2.81
2.48
0.24
_
1.60
1.40
0.20
13.21
12.97
0.24
2000
30.51
12.81
17.70
29.00
-
-
18.38
3.54
3.60
2.60
_
6.60
5.60
1.00
44.39
36.61
3.90
3.88
IV
1985
19.98
9.99
9.99
16.33
0.73
1.42
11.43
2.58
1.58
0.25
1.00
3.20
3.00
0.20
10.85
10.60
0.25
2000
45.87
19.27
26.60
23.16
15.91
5.30
13.71
2.93
3.83
2.00
8.00
14.93
13.93
1.00
11.37
10.97
0.40
V
1985
18.13
9.06
9.07
14.48
0.73
1.42
9.51
2.59
1.71
0.24
1.00
1.60
1.40
0.20
13.21
12.97
0.25
2000
39.11
16.43
22.68
16.40
15.91
5.30
8.43
2.34
3.14
2.00
8.00
6.60
5.60
1.00
24.30
16.50
3.90
3.90
1975*
15.5
8.40
7.10
7.60
8.82
2.40
2.04
0.11
0.06
0.06
1.65
1.65
*Data for 1975 are from References 2, 3, 4, 5.
-------
Ill RADIATION EXPOSURES
A. Radionuclides and Sources
The energy-related radionuclides considered in this report can be divided
into three general categories:
(1) The natural light and heavy parent radionuclides (thorium and uranium),
which occur significantly in nature
(2) The radioactive daughter products of the above heavy radionuclides
(3) Fission and transmutation products.
Table 2 presents the radionuclides of category (1), the natural light and heavy
parent radionuclides. Natural tritium and carbon-14 are continuously produced
from cosmic ray interactions in the upper atmosphere; the remaining natural
radionuclides in Table 2 are primarily of terrestrial origin. Eisenbud [Ref. 6]
lists several more singly occurring natural radionuclides; however, the above
are estimated to be the most significant for energy emission considerations.
Thorium and uranium are the naturally occurring heavy radionuclides that
decay into various daughters that are, in turn, ubiquitous in nature. The
daughters can be assumed to be in secular equilibrium with the parent, which
occurs when the half-life of the parent is much longer than that of the daugh-
ters. In secular equilibrium the activity of the daughters is roughly equal to
that of the parent. Thus, in any natural deposits of thorium-232, uranium-235,
and uranium-238 one would assume that the daughters' radioactivity is the same
as the major parent. These three series of radioactive heavy elements are all
found in the earth's crust or atmosphere and account for much of the radiation
to which man is exposed.
The third category, fission and activation products, results primarily
from neutron reactions. Products can be fission fragments of fissile isotopes
and activation products of structural or coolant material and fuel materials.
State-of-the-art nuclear systems primarily revolve around uranium-235,
which is a readily fissile isotope. Fission generally occurs when elements of
heavy mass possess characteristic cross sections for thermal or fast neutron
absorption, and liberate the absorbed excess energy by splitting into two
smaller elements and neutrons. The average number of neutron products, as well
as the fission yields of fragments, vary with energy of the absorbed neutron.
11
-------
TABLE 2. NATURAL LIGHT AND HEAVY SOURCE RADIONUCLIDES THAT
OCCUR SIGNIFICANTLY IN NATURE*
Half-Life
Radionuclide
Tritium
Carbon-14
Potassium-40
Vanadium-50
Rubidium-87
Lanthanum-138
Neodymium-144
Samarium-147
Lutetium-176
Thorium-232
Uranium-238
Uranium-235
(years)
12.3
5730
1
.26xl09
6x10 15
4
1
2
1
2
1
4
7
.8xl010
. 12x10 n
.4xl015
. 05x10 ll
.2xl010
. 41x10 10
.SlxlO9
.IxlO8
Principal
Energy (MeV)
0
0
1
1
0
1
0
0
0
1
1
2
0
3
4
4
4
4
4
4
0
0
0
Radiations
Type
Elemental
Abundance
(%)
.0186 6~
.156
.33
.46
.78
.55
.28
.21
.81
.43
.83
.23
.43
.95
.01
.15
.20
.37
.40
.58
.143
.185
.204
(89%)
(11%)
(30%)
(70%)
(24%)
(76%)
(25%)
(75%)
(18%)
(57%)
( 8%)
(11%)
(54%)
( 5%)
B"
B~
Y (EC)
0"
Y (EC)
B"
B~ (80%)
Y (EC) (70%)
Y (EC)
a
a
B~
a
a
a
a
a
a
a
Y
Y
Y
0
0
27
0
23
15
2
100
99
0
.012
.25
.9
.09
.0
.1
.6
.28
.72
*From Ref. 6.
12
-------
Activation products resulting from neutron absorption (activation) in re-
actor material and actinides resulting from neutron capture by the heavy fuels
material are also found in the spent fuel elements of nuclear reactors. For
example, equations (1) and (2) show the production of plutonium-239 from the
conversion of uranium-238.
ln H. 239u* _> 239Np
239Np ^ 239pu
This is the basis of the breeder reactor operation whereby converted
uranium-238 (as plutonium-239) is recycled as a mixed oxide with uranium. Ta-
ble 3 lists some of the more important radioactive fission and activation prod-
ucts. These products were selected from various publications [Ref. 7. 9, 49],
and all have environmentally significant half-lives.
B. Exposure Pathways and Health Effects
Exposure to radioactivity can result from both external and internal proc-
esses. Based on animal and epidemiological studies, the chronic health effects
include cancer induction and mutagenic effects (somatic and genetic) .
1. Exposure Pathways
HERMES is a comprehensive radiological exposure pathway model that is
used to calculate regional dose commitments from nuclear reactors and reprocess-
ing facilities. [Ref. 7] This Dose Calculation Model computes the exposure
doses for air and water sources that flow through various ecological pathways
to expose man either externally or internally. Table 4 presents the dose modes
for internal and external exposures.
Experience has indicated that certain radionuclides and pathways are
much more critical than others. Generally, for a specific radionuclide a few
pathways will dominate. Likewise, for a specific component of a fuel's cycle,
a singular mode of exposure will dominate. What is important in assessing the
critical pathways is a knowledge of the recipient-to-donor concentration ratio
throughout the model.
Dose assessment is a complex task and involves demography and habits,
intermedia transport models, food chain dose factoring, source-emission fac-
toring, internal metabolism, and route of entry of specific radionuclides.
These can be divided into specific pathways for specific nuclides or fuel cycle
components. Table 5 presents an example of critical pathways for specific nu-
clear reactor effluents. To present and utilize complex computational models
for ranking is beyond the scope of this report. Therefore, simplified methods
are presented.
13
-------
TABLE 3. REACTOR FISSION AND ACTIVATION PRODUCTS
Fission Products
Tritium
Krypton-85
Krypton-85m
Krypton-88
Strontium-89
Strontium-90
Yttrium-91
Zirconium-93
Zirconium-95
Niobium-95
Molybdenum-95
Molybdenum-99
Tellurium-99
Ruthenium-103
Ruthenium-106
Antimony-125
Tellurium-127m
Tellurium-129m
Tellurium-132
Iodine-129
Iodine-131
Iodine-132
Iodine-133
Iodine-135
Xenon-131m
Xenon-133m
Xenon-133
Xenon-135m
Xenon-135
Xenon-138
Cesium-134
Cesium-135
Cesium-137
Barium-140
Lanthanum-140
Cerium-141
Cerium-144
Promethium-147
Europium-155
Activation Products
Tritium
Nitrogen-13
Carbon-14
Sodium-22
Sodium-24
Argon-39
Argon-40
Calcium-41
Manganese-54
Iron-55
Iron-59
Cobalt-58
Cobalt-60
Nickel-61
Copper-64
Zinc-65
Zirconium-95
Niobium-95
Molybdenum-99
Uranium-235
Uranium-236
Uranium-237
Uranium-238
Uranium-239
Neptunium-237
Plutonium-238
Plutonium-239
Plutonium-240
Plutonium-241
Plutonium-242
Plutonium-243
Americium-241
Americium-242m
Americium-242
Americium-243
Americium-244
Curium-242
Curium-243
Curium-244
14
-------
TABLE 4. MODES OF HUMAN EXPOSURE
External Dose
Internal Dose
Air submersion
Transpiration
Soil exposure
River bank exposure
Water immersion
Water surface exposure
Irrigated fields
Air Pathway
Transpiration (tritium)
Inhalation
Water Pathway
Water consumption
Fish consumption
Water fowl consumption
Food Pathway
Vegetables (crops)
Meat
Milk
Dairy foods
Eggs
2. Liquid Effluent Releases and Dose Computation Methodology
Once an emission factor has been determined for a specific radionu-
clide, the dose received by a human or a population from water emissions can be
computed by considering stream dilution, aquatic biota concentration, radioac-
tive decay, and human water and aquatic biota consumption habits. Because we
are only concerned about ranking various radionuclides, and not the actual com-
putation of dose, the following static model given as equation 3 [Ref. 9] can be
used to compare aquatic effluents:
D - g - e"Ui
W M
n
P DF
where D «= water dose for radionuclide, yCi/yr
w
Q • release, yCi/yr
M - mass flow of receiving waters, cm3/yr
15
-------
TABLE 5. PRINCIPAL EXPOSURE PATHWAYS FOR RADIATION EXPOSURE
FROM NUCLEAR REACTOR EFFLUENTS*
Radionuclide
Radioiodine
Tritium
Noble gases
Cesium
Transition
metals (iron,
cobalt,
nickel, zinc,
manganese)
Direct radiation
Discharge
Mode Principal Exposure Pathways
Airborne Ground deposition-external irradiation
Air inhalation
Grass-cow-milk
Leafy vegetables
Water Drinking water
Fish consumption
Shellfish
Airborne Air inhalation and transpiration
Submersion
Water Drinking water
Food consumption
Airborne External irradiation
Airborne Ground deposition-external irradiation
Grass-cow-milk
Grass-meat
Inhalation
Water Sediments-external irradiation
Drinking water
Fish consumption
Water Drinking water
Shellfish consumption
Fish consumption
External irradiation
Critical Organ
Whole body
Thyroid gland
Thyroid gland
Thyroid gland
Thyroid gland
Thyroid gland
Thyroid gland
Whole body
Skin
Whole body
Whole body
Whole body and skin
Whole
Whole
Whole
Whole
Whole
Whole
Whole
body
body
body
body
body
body
body
Gastrointestinal tract
Gastrointestinal tract
Gastrointestinal tract
Whole body
^Reference 8.
-------
t. = time between consumption and emission, day
n B number of pathways
P = pathway transfer factor, cm3/yr
DF = stream dilution factor upon entrance to pathway
A = radionuclide decay constant, day'1
There obviously are many pathways by which humans can become exposed;
however, the two most important are the consumption of aquatic biota and water.
Thus, the parameter P (pathway transfer factor) requires further definition.
For water: P = I F
w w
For food: P = R[CF I + CF I ]
I S S
where I « intake of shellfish, cm3/yr
s
intake of fish, cm3/yr
I = drinking water consumption rate, cm3/yr
F = water treatment loss factor
R = food preparation loss factor
CF = fish concentration factor
CF = shellfish concentration factor
s
Thus, the overall water exposure dose equation is:
Dw " M [IwF (DF) e± + R(DF) (e±) (I CF + V'V1
The simple evaluation for ranking relative exposures between aquatic radionu-
clides would be the computation presented in equation 4, and would be a func-
tion of each of the radionuclide variables Q, F, DF, X, R, CF , and CF.
s
Site-specific differences for various components of the fuel cycles
would be M (receiving water stream flows), t± = (time between emission and ex
posure), Q for each radionuclide, and other factors. Assuming unit streams,
17
-------
consumption habits, and elapsed times, equal comparisons can be made. The in-
put values for a unit site might be: [Ref. 9]
M = 2000 ft3/s = 1.78 x 1015 cm3/yr
I = 8 x 105 cm3/yr
I = 1.2 x 103 cm3/yr
I = 5.1 x 103 cm3/yr
t^ = 30 days
3. Atmospheric Emissions and Computation Methodology
As in the case of water emissions, we are concerned with computing
exposures for each specific radionuclide following atmospheric emissions.
Such a computation would involve a dispersion factor, radionuclide decay,
transpiration, deposition and fallout, terrestrial-biotic interactions re-
sulting in a food-chain dose, and direct inhalational and external exposures.
Thus, in computing exposures from air sources we need to take into account
both atmospheric and ingestion processes.
Equation 5 is used to provide a Gaussian concentration profile in the
vertical direction and a uniform concentration in horizontal directions due to
fractional wind frequencies and stability classes in various sectors. [Ref. 9]
* -*-/2\1/2 FWF e-("2/2a|) e-*t
Q'~Q~I^/ —r~:—; (5)
V / ua 2irr/n
where x = ground level airborne concentration, Ci/m3
V = time integrated ground level concentration-exposure, Ci-s/m3
Q' - source release rate, Ci/s
Q = time integrated release (i.e., total release/yr), Cl
FWF = fractional wind frequency in a sector
r = distance from the stack, meters (m)
n • number of sectors
2irr/n » sector width at distance r, m
h = effective stack height, m
18
-------
a = standard deviation of the vertical distribution of an assumed
Gaussian cloud, m
u = average wind speed in the sector, m/s
X = decay constant of radionuclide , s
and t = transit time to distance r from the stack (t = r/u) , s.
This model is the basis for HERMES where direct inhalation or trans-
piration doses can be computed by equation 6: [Ref . 9]
D = r Q (DCF) (6)
where D = dose rate, mrem/yr
X/Q'= atmospheric dispersion factor as computed above, s/m3
Q = annual release rate, Ci/yr
DCF = an appropriate dose conversion factor for the radionuclide and
exposure mode of interest, mrem-m3/Ci-s
Further computations give fallout, bioconcentration factors, and so
on, to compute the dose for each specific radionuclide. The above models are
too complex for our purposes.
To rank atmospheric emissions we would need unit model sites and con-
ditions for various components of the cycles. A macrocondition pertains to a
condition of the earth as a whole (for long-lived and widely distributed radio-
nuclides) , whereas microconditions pertain to local conditions around the
source (for short-lived and locally distributed emissions) . Macromodels have
been derived from worldwide fallout models to assess the impact of atmospheric
nuclear testing. In our case, however, we are limited to more defined source
emission patterns for various components of fuel cycles. Thus, at this stage,
prior to any common algorithmic expression, it is necessary to define which
radionuclides would be of importance in worldwide, quasi-equilibrium exposure
situations. The long-lived fission and activation products that would be ex-
pected to have macrodistributions would be the gases tritium and krypton-85,
and the volatile iodine-129. Carbon-14 would also be expected to have a sig-
nificant macrodistribution. [Ref. 10]
Often overlooked in calculating exposures is the macroregional dis-
tribution of radon-222 progeny (-»• lead-210 -»• polonium-210) . Radon is the
source factor emanating to the atmosphere from uranium (radium-226) terrestrial
formations when they are disturbed. Radon emanates naturally from soils at the
average crustal rate of 1.4 pCi/m2-s [Ref. 11]; however, anthropogenic emissions
can be much greater [Ref. 12]. Although there are generally four decay steps
between radon-222 and lead-210, the half -lives of the intermediate radon-222
19
-------
daughters are relatively short, thus lead-210 biochemistry and geochemistry
govern exposure along media transition zones. A case in point is that smokers
are burdened with greater concentrations of polonium-210 in both lungs and
bones than nonsmokers, due to its deposition on the broad leaf tobacco plant
[Ref. 13].
To perform an initial assessment of the magnitude of radon emanation
from both anthropogenic activities (mining-waste disposal) and natural emana-
tion, we should consider the macroscale aspect. Using the following equation
based on the previously given natural emanation rate of radon, we calculate:
where Q = conterminous U.S. natural radon-222 emanation rate, (Ci/yr)
= 1.4 pCi/m2-s
A = 9.4 x 10 12 m2 for the United States
Solving the equation for Q , we obtain:
0^ = 4.13 x 108 Ci/yr
Thus, approximately 400 MCi of radon-222 per year are emitted naturally from
the conterminous United States. In performing an initial assessment, this
natural emission would have to be compared with anthropogenic emissions. An-
thropogenic emissions would involve (1) the emanation of terrestrially en-
trained radon-222 from uranium-238 ores during mining activities, and (2) the
residual emanation of radon-222 from its parent radium-226 in spent mill piles
(see Section IV.E-2).
Table 6 presents a summary of the mode and dose of natural radioac-
tive backgrounds in the United States. Because of radon-220's short half-life,
its longer-lived daughters, lead-212 and bismuth-212, are essentially in equi-
librium and will contribute to higher exposure doses than those listed for
radon-220.
20
-------
TABLE 6. SUMMARY OF DOSE EQUIVALENT RATES (MREM/YEAR) FROM VARIOUS RADIONUCLIDES
COMPOSING THE NATURAL BACKGROUND RADIOACTIVITY IN THE UNITED STATES FOR
EXTERNAL (E), AIRBORNE (A), AND INTERNAL (I) EXPOSURES*
Radionuclide
Carbon-14
Potassium-40
Rubidium-87
Uranium series
Uranium-238 (Uranium-234)
Radium-226
Radon-222
Polonium-218 (polonium-214)
Lead-210 (polonium-210)
Thorium series
Thorium-232
Radium-228
Lead-212 (bismuth-212)
Radon-220
Mode of
Exposure
Gonads Lung
I
E
I
I
E
I
A
I
A
I
A
A
I
A
E
I
I
A
I
0.7
8
19
0.3
6
0.8
—
0.2
—
0.4
—
—
6
—
12
0.0
0.3
—
0.0
0.7
8
19
0.3
6
0.8
0.2
0.2
0.2
0.4
2t
90§
3
11
12
0.0
0.3
3
0.0
Bone
Surfaces
0.8
8
15
0.6
6
4.8
6.6
0.4
24
12
0.7
8.0
Bone
Marrow
0.7
8
15
0.6
6
0.9
1.2
0.4
4.8
12
0.1
1.0
Gastrointestinal
Tractt
0.7
8
19
0.3
6
0.8
0.2
0.4
0.2
0.2
12
0.0
0.3
0.0
^Reference 14.
tDose equivalent rate to the gastrointestinal tract is considered to be the same as for soft tissue, with
no allowance for irradiation by the gut contents.
tDose equivalent rate to bronchial surfaces.
§Dose equivalent rate to the segmental bronchioles would be 450 mrems/year.
-------
Macroregional dose calculations for atmospheric emissions can be
mathematically expressed by equation 7. Although the source factor (Q) is for
air, the dose may also be for ingestion due to terrestrial deposit and bio-
logical accumulation. This general equation can be used to assess the rela-
tive annual doses received from annual releases of various radionuclides. Cum-
ulative dose summations would be required for each year's increase in Q.
A = Qe-Xti fiF1Ci (7)
where A = dose from water (oral) or air (respiration), uCi/yr
Q = release factor, yCi/yr
X = decay constant for radionuclide, yr"1
t « time between emission and consumption, yr
f « environmental dilution factor for exposure compartment i, yr/cm3
F = concentration factor for exposure compartment i, dimensionless
C. - consumption for exposure compartment i, cm3/yr
For macroregional exposures, as previously addressed, we would re-
quire evaluations based upon media dilution factors for appropriate pathways.
These factors are presented in Table 7 for tritium, carbon-14, krypton-85,
iodine-129, and plutonium-239.
4. Local Dose
Local radionuclide air doses from common sources in various compo-
nents of the fuel cycles are calculated from constant environmental exposure
conditions. In this methodology a simple area or point source model is used
to compute average annual concentrations 3 kilometers (km) from the source.
For calculating deposition, the average fallout is computed under the assump-
tion that 80% of nonvolatiles are deposited within 10 km of the plant
[Ref. 15]. Assuming a stack height of 100 m and wind speed of 3 m/s, the dry
deposition velocity term Vd would be approximately
2 cm/s
where x is the downwind distance.
Using the following equation, the downwind ground level concentra-
tion can be computed for various source strengths and downwind distances from
an area source. Turner's [Ref. 16] approximation of the horizontal dispersion
22
-------
TABLE 7. MACROREGIONAL EXPOSURE COMPUTATION VARIABLES
to
u>
Radionuclide
Tritium
Carbon-14
Krypton-85
Iodine-129
Plutonium-239
t
(yr)
1
20
1
1
100
A
(yr-1)
0.056
1.2 x ID'1*
0.64
4.3 x 10-8
2.8 x 10~5
Exposure Compartment
Eastern U.S. rain water
Transpiration
Hemisphere (inhalation)
Biosphere (air -»• food)
Northern hemisphere
(immersion)
Eastern U.S. land area
(grass •> cow -»• milk)
Eastern U.S. land area
(inhalation from
resuspension)
(yr/cm3)
2.53 10~19
2.53 x 10~19
2.51 x 10~25
2.51 x 10~25
2.51 x 10~25
2.6 x 10-17
(yr/cm2)
2.6 x 10"17
(yr/cm2)
Fi
0.5
1
1
1
1
2.8
(cm'1)
ID'7
(cm'1)
Ci
(cm3/yr)
8 x 10s
3.65 x 105
7.3 x 109
1.5 x 105
*
2.5 x 105
7.3 x 109
*Not applicable for immersion.
-------
is used where oyo = s/4.3, and s represents the side of the square area of the
source. A virtual distance can be found, Xy, which varies with stability. Vir-
tual distance Xy is then added to actual distance x for determining oy. With
the area source being ground level, the model can be adjusted to include the
radioactive decay term:
(8)
where x = air concentration, yCi/m3
Q = source strength, pCi/s
X = radioactive decay constant, s"1
t = time between emission and exposure, s
U = average wind speed, m/s
o ,ag = vertical and horizontal diffusion coefficients, m
Assuming area sources of 4 hectares (200 m)2, a = 46.8 m. Thus at Class E
stability: y°
X/Q = 1.45 x 10~5e"Xt:
where the following values have been used:
x = 3000 m, x = 900 m, and a = 170 m at x + x
U = 3 m/s
o_ (3 km) = 43 m
B
From point sources, if we assume that the radionuclide is uniformly distributed
within horizontal sectors, annual average downwind concentrations can be esti-
mated by [Ref. 16]
_ 2.03 Q e-0-5(H/a2)2e,Xt
azUx
where definitions are the same as above and
x • downwind distance, m
H • stack height, m
24
-------
The model facility for the point source assumes Class E (thin overcast with
2-3 m/s winds) stability and would have:
U = 3 m/s
H = 100 m
x - 3000 m
ag = 43 m.
The x/Q term thus reduces to 3.51 x 10-7e~U. The above calculation can also
be used to evaluate the relative atmospheric concentrations for the various
area and point source nuclides in local environments.
Food-chain transport is of considerable importance for certain radio-
nuclide exposures, as well as resuspension factors for others. Thus, the above
models can be used to calculate the exposures via adjustments for deposition
into soil and plants.
Radioactivity deposited on the ground surface can be calculated
[Ref. 8] by
Where Q1 is the annual release rate (yCi/yr), Vj is the deposition velocity
(0.02m/s), and W is the deposition rate (pCi/m2-yr). Thus, the depositions at
3 and 80 km are:
W3 = 7.02 x 10-9e~XtQ
W80 - 7.02 x 10-13e~AtQ
With the appropriate bioconcentration or resuspension factors, W is the deposi-
tion function for exposure estimates between radionuclides. This dose compu-
tation would be the same as for equation (7) but expressed as:
A = W F^ (10)
wljere the FJ routes could be resuspension, ground -*• plant •* man, plant ->• cow -»•
milk -»• man, and plant -*• man and C^ is the annual consumption of the product.
The soil acts as a reservoir for the accumulated deposit, thus increasing dose
over time. The deposit is lost from the system through percolation, resuspen-
sion, or wash-off, as well as the physical decay constant of a particular
25
-------
radionuclide. The accumulated deposit in yCj/m2 can be given as:
(X/Q) Q'VJI - e~Xet]
X
e
where X = the sum of the removal rates for the various processes (i.e.,
6 *a + xb + "')» yr-1> and the other terms are as previously defined.
The rationale in the above methodologies is the computation of con-
stant human exposure concentrations for the principal pathway media, given Q's
and environmental parameters. In estimating human dose, all modes must be taken
into consideration to calculate actual exposures from all consumption; then,
dose factors must be supplied for the critical organs of the exposed populations.
These organ dose factors convert radioactivity per cm3 of the exposure medium to
dose (rem) per year. Thus, given a quantity of consumption per year, the annual
critical organ dose can be computed. Within the limits of our calculations, it
would be inaccurate and misleading to compute estimates of population dose for
other than ranking purposes for the following reasons:
• The results may be erroneously taken to imply health effects
• Variable demographic patterns and habits have not been incorporated
• All exposure media have not been utilized
• There are variable source conditions that are beyond incorporation
in these models
• Environmental transfer and concentration factors are, at best,
order of magnitude estimates
• Our estimates and results appear conservative.
To estimate the organ dose, the calculations would be as follows:
D - Y C F
mj Xji i mj
where D . - dose to organ m from radionuclide j, rem/yr
X.. = annual average concentration of radionuclide j in exposure mode i,
J uCi/cm3
C. = consumption of exposure mode i, cm3/yr
Fj = the exposure mode dependent dose conversion factor for organ m and
^ radionuclide j, rem/vCi
26
-------
These organ dose conversion factors generally assume homogeneous distribution
in organs. To obtain total dose to organs from specific sources all D 's
should be summed. mJ
There are Federal regulations (10CFR20.1) on the maximum permissible
concentrations of radionuclides allowed in effluents to unrestricted areas.
The regulations for air and water are presented in 10CFR20, Appendix B, Ta-
ble II.
A considerable amount of work has been done in the past on calculat-
ing limits for all of the radionuclides. The most comprehensive tabulations of
the background, primarily for internal dose, are found in ICRP #2 (1959) or in
Health Physics (1959) [Ref. 17].
The basic formula that is used to calculate organ dose (rem/yr) is:
1.86 x
D
m
where D = dose, rem/yr
m
B = organ burden,
m
E = energy per disintegration, MeV
Q.F. = quality factor for radiation type
G = organ mass, g
The constant (1.86 x 101*) is derived from the integral of the time period for
energy (rad = 100 ergs/g) absorbed by the tissue per microcurie. The value (E)
is highly dependent on the type of energy, its level, and the nature of the
tissue. This value would hold constant if equilibrium existed in the organ.
Because energy emissions per disintegration are constant, the product of (E)
and (Q.F.) can be calculated for each organ of the body and radionuclide, as
the effective energy (E1), where Q.F. varies with radiation type.
The maximum permissible body burden for each radionuclide is the maxi-
mum amount the body can take without delivering a dose exceeding the maximum
permissible level (rem/yr) for a critical organ. The maximum permissible bur-
dens are the object of continual research.
The EPA has proposed standards on Radiation Protection for Nuclear
Power Operations (FR23432, 1975) limiting dose equivalent rates to the maximum
permissible levels for the critical body organs in Table 8. This table also
included proposed standards for the specific macrodistributed nuclides,
krypton-85, iodine-129, and some actinides.
27
-------
TABLE 8. PROPOSED RULES (FR23421, 1975)
(a)
Organ
Wholebody
Thyroid
Others
mrem/yr
25
75
25
Nuclides
Krypton-85
Iodine-120
Long-lived
actinides
Ci/GW-yr
50,000
0.005
0.0005
Maximum permissible body burdens can, thus, be translated to annual
continuous exposures and to specific radionuclides or mixtures of radionuclides
(the latter contribute to organ dose as a whole) given the fraction of the body
burden that is present in the organ in question. This value, fmj, is specific
for each radionuclide and organ. For example, assuming the thyroid organ mass
for a standard adult is 20 g, fmj for iodine-129 is 0.25 and E1 = 0.046, the
value for the maximum permissible body burden, based on iodine-129 exposure of
75 mrem/yr to the thyroid, is
B
(0.075)(20)
1.86 x 10H(0.046)(0.25)
10
-3
and the burden in the thyroid is
Bm = fmjB
7 x 10'3 = 1.75 x 10~3 yCi
Table 9 presents the maximum permissible concentrations for air (MPCa) and water
(MPCw) formulas for continuous exposures to air and water.
C. Human Exposure Calculation
MPCs will vary according to the specific radionuclides in question and to
the conversion factors (rem/yCi) under consideration. As shown, the conversion
factors will be a function of the maximum permissible organ concentration for a
specific radionuclide based on health effects evaluations, its solubility and
the fraction from air and water reaching the critical organ (fa, L.) , and the
radionuclide 's effective half-life (T£) . For ranking purposes, the isotopes in
Table 10 are considered, with MPCa and MPC^, from 10CFR20, Appendix B, Table II.
The minimum MPCs between soluble and insoluble forms is presented. Based upon*
the organ exposures presented in NBS Handbook #69 [Ref. 18], which generally
show allowable continuous occupational exposures a factor of 10 higher than the
28
-------
current NRC regulations, dose conversion factors are calculated (rem per year/
yCi per year) for the critical organ and are presented in the right-hand col-
umns. Some changes have been made in the table, most notably for radon-222,
radium-226, uranium-235, and uranium-238. These changes were based on dose
conversion factors presented in Ref. 8 and 9 concerning the nuclear fuel cycle.
Modifications for radon-226, uranium-235, and uranium-238 should be used with
considerable caution because of their more than ten-fold differences from the
NBS data. For the latter, the MFCs were changed to correspond to 3 rem/yr for
bone and 1.5 rem/yr for lung. Furthermore, the dose conversion factors were
for the minimum MFC, soluble(s) or insoluble (i). Thus, the use of a given
factor implicitly assumes that all of the nuclide is in either the i or s form.
TABLE 9. MAXIMUM PERMISSIBLE CONCENTRATION FORMULAS
MPCair(yCi/cm3) MPCwater(yCi/cm3)
5 x 10-8(B)(f ) 4.5 x 10^(6) (f )
MFC. ,c~. ;T MFC
i . w
Inhalation = 2 x 107cm3/day Water = 2200 cm3/day
B = body burden which will account for maximum critical organ
exposure, yCi
f = fraction of body burden in critical organ
m
f ,f = fraction of nuclide reaching critical organ
aw T T
B R
T = biological effective half-life , days , = ^ — . T
E *B R
T^ = biological half-life, days
B
T_ - radiological half -life, days
R
(50) •» 50 years in days = 1.8 x 10** days
Thus, for iodine-129, for which Bfm was calculated at 1.75 x 10~3 yCi,
and with fa = 0.23, fw = 0.3, Tfi = 138, and TR = 6.2 x 109,
29
-------
TABLE 10. MAXIMUM PERMISSIBLE CONCENTRATION
Dose Conversion Factor [rem/yiCi]'
OJ
o
Radioisotopes
Tritium
Carbon-14
Potassium-40
Iron-59
Krypton-85
Rubidium-87
Strontium-89
Strontium-90
Ruthenium-103
Ruthenium-106
Iodine-129
Iodine-131
Xenon-131m
Cesium-134
Cesium-137
Samarium-147
Lead-210
Radon-222
Radium-226
Thorium-232
Dranium-235
Uranium-238
Plutonium-238
Plutonium-239
Plutonium-240
Plutonium-241
Polonium-210
m^i
,/ cm-
j
MVP KTO/i
nrij
W
3 x
8 x
7 x
5 x
—
10-3(s)t
10-* (s)
10-6(i)
10-5(i)
10-* (s)
3 x
3 x
8 x
10~b(s)
10-7(s)
10-5(i,s)
IO-5 (i,s)
6 x
3 x
—
9 x
2 x
6 x
10-8(s)
10-7(s)
10-6(s)
10-5(s)
10-5(s)
10-7(s)
— —
2 x
2 x
3 x
3 x
5 x
5 x
5 x
2 x
7 x
10-7(s)
10-6(s)
10-7(s)
IO-7 (s)
10~6(s)
10-6(s)
IO-6 (s)
10-* (s)
10-7(s)
2
10
1.
2
3
2
3
3
3
2
2
1
4
4
5
2
4
4
1
x
-7
5
x
x
X
X
X
X
X
X
X
X
X
X
X
X
X
X
&'U W
a
10-7(s)
s
x 10-
io-9(
io-7
9(i)
i)
10-9(i)
io-20
io-11
10~in(
10- 10
10-"
io-10
io-7
io-10
10-10
io-12
io-12
io-10
10- 13
(i,s)
(s)
i)
(i)
(s)
(s)
(1)
(i)
(s)
(s)
(i)
10-12(i,s)
1
1
7
6
6
3
7
X
X
X
X
X
X
X
10- 1 3
io-13
10- "
10-1*
io-1*
io-12
(i)
(s)
(s)
(s)
(s)
(3)
10-12(s)
Water
Organ
Tissue
Fat
GI
GI
—
Pancreas
Bone
Bone
GI
GI
Thyroid
Thyroid
—
Total body
Total body
Bone
Kidney
Bone
Bone
Bone
Bone
Bone
Bone
Bone
Bone
Spleen
2.
7.
0.
0.
__
0.
1.
12
0.
0.
62
12
—
0.
0.
0.
Factor
1 x 1Q-*
8 x 10~*
26
037
019
2
02
19
.5
.5
07
03
06
18.7
16.4
1.
8
12.3
12.3
0.
0.
0.
0.
2.
7
7
7
017
7
Air
Organ
Tissue
Fat
GI
Lung
Total body
Lung
Bone
Bone
Lung
Lung
Thyroid
Thyroid
Total body
Lung
Lung
Bone
Kidney
Lung
Lungt
Bonet
Lung
Lung
Bone
Bone
Bone
Bone
Lung
Factor
3.5 x
6.8 x
0.14
0.103
2.3 x
0.1
1.3
13
0.07
1.02
20.6
4.1
1.7 x
0.51
0.41
192
51.2
0.54
1506
383
1428
1428
5471
6383
6383
127
29
10-*
10-*
10-*
10-*
*R>r ranking purposes only.
tSoluble(s); insoluble (i).
tAssumed.
-------
the proposed EPA regulations would be:
MPC = 2.7 x 1Q-12 MCi/cm3 of air
a
MPCw = 1.9 x IS'8 uCi/cm3 of water
1. Water Dose
Table 11 presents the variables to be used in computing from equa-
tion 4 the water exposure dose for a man. Columns a, b, and c provide the dose
(yCi/yr), organ, and exposure (rem/yr) to the critical organ when Q (yCi/yr) is
substituted. On checking tritium results of Table 11 with Q data on page 130
of Ref. 9, the rem/year factor calculates Q5E-15,* a factor of approximately
1/20 of column (c)—tritium. This is surprising, in that almost all variables
(including dilution) were similar. The dose conversion factor from Table 10
(2.1 x IQ-4* rem/yCi) for tritium is probably excessive by a factor of 2, as is
the water consumption parameter (2200 cm3/day); however, this should only lead
to a reduction of about one-fifth. Thus, the data should be used with caution
and only for ranking purposes until parameters are firmly reevaluated and reas-
onably established with uncertainty estimates. Furthermore, this data should
not be construed to imply health effects.
2. Atmospheric Dose
a. Macroregional
Table 12 presents the algorithms to be used in computing macro-
regional dose based on equation 7 and the variables presented in Table 7. This
would apply primarily to the United States.
b. Area Source
The only significant radionuclide emitted from a broad area
source would be radon-222. Equation 8 presents the area source calculation
with a x/Q of 1.45 x 10~5 e"^c for a 4 hectare source. A 40-hectare source
would have a reduction effect of about 0.4, and annual averages would have a
reduction effect of 0.1 to 0.5. Thus, using a x/Q of approximately
1.45 x 10~6 e-A* (yCi/m3), the dose algorithms for radon-222 (inhalation) for a
person continuously at 3 km would be
Exposure Dose Organ Dose
(uCi/yr) (rem/yr)
Q 3.4E-10 Q 1.8E-10
*This notation (Q5E-15) is equivalent to Q x 5 x 1Q-15.
31
-------
TABLE 11. WATER TRANSPORT DOSE VARIABLES AND Q-DOSE ALGORITHMS
Equation (4) Radionuclide Variable Values
10
Radioisotopes
Tritium
Carbon-14
Potassium-40
Iron-59
Krypton-85
Rubidium-87
Strontium-89
Strontium-90
Ruthenium- 103
Ruthenium-106
Iodine-129
Iodine-131
Xenon-131m
Cesium-134
Cesium-137
Samarium- 14 7
Lead-210
Radon-222
Radium-226
Thorium-232
Uranium-235
Uranium-238
Plutonium-238
Plutonium-239
Plutonium-240
Plutonium-241
Polonium-210
A (day'1)
1.54 x ID'1*
3.31 x 10-7
1.4 x 10~12
0.0154
1.76 x 10-*
4 x 10-11*
0.0133
6.7 x 10~5
0.017
1.88 x 10-3
1.1 x 10~10
0.086
0.131
9.2 x ID"1*
6.3 x ID"5
1.7 x lO'14
9 x 10~5
0.18
1.2 x 10~6
1.3 x 10~13
2.7 x 10~12
4.2 x 10~13
2.2 x 10-6
7.9 x 10-8
2.9 x ID'7
1.26 x 10-1*
5 x ID-3
(a)
(b)
ri? Exposure Dose, l\,
DF
1
1
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
0.05
F
1
1
(0.5)*
0.2
—
(0.2)
(0.2)
(0.2)
0.2
0.2
0.8
0.8
—
0.8
0.8
(0.2)
(0.2)
—
(0.2)
(0.2)
(0.2)
(0.2)
(0.2)
(0.2)
(0.2)
(0.2)
(0.5)
R
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
0.8
CF
1
1
(500)
1,000
—
(50)
50
50
100
100
1
1
__
1000
1000
(50)
(50)
— —
(10)
(10)
(10)
(10)
(10)
(10)
(10)
(10)
(10)
v»x
s
1
1
(500)
20,000
—
(50)
500
500
1000
1000
25
25
—
1000
1000
(50)
(50)
—
(10)
(10)
(10)
(10)
(10)
(10)
(10)
(10)
(10)
(uCi/yr)
Q4.5E-10
Q4.5E-10
Q8.2E-11
Q8.2E-11
Q1.1E-11
Q1.1E-11
Q1.6E-11
Q2.4E-11
Q2.6E-11
Q4.3E-11
Q2.3E-11
Q1.9E-12
Q1.6E-10
Q1.6E-10
Q1.1E-11
Q1.1E-11
Q6E-12
Q6E-12
Q6E-12
Q6E-12
Q6E-12
Q6E-12
Q6E-12
Q6E-12
Q1.5E-11
Organ
Tissue
Fat
GI(i)f
GI(i)
Pancreas (s)
Bone(s)
Bone(s)
GI(i)
GI(i)
Thyroid(s)
Thyroid (s)
Total body(s)
Total body(s)
Bone(s)
Kidney (s)
Bone(s)
Bone(s)
Bone(s)
Bone(s)
Bone(s)
Bone(s)
Bone(s)
Bone(s)
Spleen (s)
» (c)
Organ Dose, Dm
(rem/yr)
Q9.4E-14
Q3.5E-13
Q2.1E-11
Q3E-10
Q2E-13
Q2E-11
Q3E-10
Q5E-13
Q8E-12
Q1E-9
Q2E-11
Q1E-11
Q5E-12
Q7E-13
Q2E-10
Q9E-11
Q1E-11
Q7E-11
Q7E-11
Q4E-12
Q4E-12
Q4E-12
Q1E-13
Q4E-11
'Estimates ( )
^Soluble (s); insoluble (i).
-------
TABLE 12. MACROREGIONAL DOSE ALGORITHMS
Nuclide
Tritium
Carbon-14
Krypton-85
Plutonium-239
Exposure Dose
(yCi/yr)
Q1E-13*
Q9E-14
Q2E-15
Q2E-25
Q2E-14
Organ
Tissue (water)
Transpiration (air)
Body fat
External
Lung (i)^
or
Bone(s)
Organ Dose
(rem/yr)
Q2E-17
Q3E-17
Q5E-17
Q1.3E-18
Q5E-29
Q2E-11
Q1E-10
Q1E-13 = Q x 1 x 10~13
Insoluble (i); soluble (s)
where Q is in yCi/yr emanating from the source. From the source, an additional
factor of lO'1* can be used for computing concentrations out to 80 km; however,
daughter build-up must be taken into consideration. In the above case, we
have considered a dose at 3 km. At 3 m/sec it would take the effluent about
15 minutes, or 0.01 day, to travel 3 km. During this time, about 0.2% of the
radon-222 activity would have decayed. At 3 km, the daughter, polonium-218,
would have built up to 5.3 x ID"1* of the radon-222 activity, or at this point
polonium-218 exposure would be Q1.2E-12 yCi/yr, where Q is in microcuries of
radon-222 per year.
c. Point Source
Point source atmospheric emissions are presented in equation 9,
which gives a x/Q term of 3.5 x 10~7e~Xt out to 3 km for annual average
concentrations. For average concentrations out to 80 km, a factor of approxi-
mately lO'1* should be applied.
As presented in equation 9, Q is in uCi/s, and x is in uCi/m3.
Adjusting these terms for annual emissions (3.2 x lO'8 yr/s) and annual inhal-
ation (7300 m3/yr), the annual dose (pCi/yr) can be computed at 3 km. Deple-
tion due to fallout has not been taken into consideration; thus, when extra-
polating dose to 80 km, the results should be interpreted with caution as being
conservative. For 3 km, the annual dose (uCi/yr) would be estimated by
8Qe-XtE-ll and to 80 km approximately 8Qe~AtE-15 where t is 0.01 day (3 km)
and 0.31 day (80 km).
33
-------
Table 13 presents the exposure dose and organ dose (rem/yr)
for the radionuclides listed in Table 10.
3. Air-Food Dose
Equation 10 presents the air-food exposure pathway based on W
(yCi/m2-yr), the deposition factor, calculated for the model facility as
Q'7 x 10"9e~xt at 3 km. For up to 80 km, W is Q'7 x lO'13^'.
Exposure will be considered for four modes:
(a) Air*ground-»plant-*inan
(b) air-*plant-*man
(c) air-»plant->x:ow-»milk-»man
(d) air-*ground-»man
For mode (a), the deposit is assumed to uniformly mix in the top
cm of the soil. Assuming a soil density of 2.7 g/cm3, the soil mass density
would be 135 kg/m2. It is also assumed that the average adult consumes 30 kg
of plants/year. Thus: process (a) reduces to:
F Q'1.6 E-9 e~At yCi/yr (a)
SP
where t = 90 days between emission and plant consumption and F = fraction in
plant for unit ground deposit.
For mode (b), 0.25 of the deposit is retained by the plant, it is
"weathered" away with a half-life of 12 days, and only a fraction remains in
the portion ingested; more is lost during preparation (0.1). The expression for
this process is:
Q'2.9 x 10-ne~Xt uCi/yr (b)
For mode (c), 0.25 of the deposit is retained on the plant as above,
0.3 is lost by weathering before cow consumption, the cow forages 0.4 of the
year on 36 m2 per day, the radionuclide transfers to milk at a ratio
S, (uCi/ liter per yd/day), and an adult consumes 84 liters milk per year.
TRe expression for this process is:
S.Q'6.3 x 10"7e-Xt yCi/yr (c)
34
-------
TABLE 13. LOCAL ATMOSPHERIC EXPOSURE ALGORITHMS
Exposure Dose (yCi/yr)
Organ Dose (rem/yr)
in
A (day-1)
1.5 x 10-1*
3.3 x 10-7
1.4 x 10-12
0.0154
1.76 x 10-u
4 x lO-14
0.0133
6.7 x 10~5
0.017
1.9 x ID"3
1 x lO"10
0.086
0.131
9.2 x 10-"
6.3 x 10-5
1.7 x 10-14
9 x 10-5
0.18
1.2 x 10-6
1.3 x 10~13
2.7 x 10-12
4.2 x 10-13
2.2 x 10~6
7.9 x 10-8
2.9 x 10~7
1.3 x 10"1*
5 x 10~3
3 km
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
8QE-11
80 km
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
8QE-15
7 . 8QE-15
7 . 7QE-15
8QE-15
8QE-15
8QE-15
8QE-15
7 . 5QE-15
7 . 5QE-15
7 . 5QE-15
7 . 5QE-15
7 . 5QE-15
7.5QE-15
7 . 5QE-15
7.5QE-15
7.5QE-15
7.5QE-15
Organ
Tissues (s)
Fats(s)
GI(i)
Lung(i)
Total body
Lung(i)
Bone(i.s)
Bone(s)
Lung(i)
Lung(i)
Thyroid (s)
Thyroid (s)
Total body
Lung(i)
Lung(i)
Bone(s)
Kidney (s)
Lung
Lung(i)
Bone(i,s)
Lung(i)
Lung(i)
Bone(s)
Bone(s)
Bone(s)
Bone(s)
Lung(i)
3 km
Q3E-14
Q5E-14
QE-11
Q8E-12
Q2E-14
Q8E-12
QE-10
QE-9
Q5.6E-12
Q8E-11
Q1.6E-9
Q3.3E-10
Q1.4E-14
Q4E-11
Q3.3E-11
Q1.5E-8
Q4.1E-9
Q4.0E-11
Q1.2E-7
Q3.1E-8
Q1.1E-7
Q1.1E-7
Q4.4E-7
Q5.1E-7
Q5.1E-7
Ql.OE-8
Q2.3E-9
80 km
Q3E-18
Q5E-18
QE-15
Q8E-16
Q2E-18
Q8E-16
QE-14
QE-13
Q5.6E-16
Q8E-15
Q1.6E-13
Q3.2E-14
Q1.3E-18
Q4E-15
Q3.3E-15
Q1.5E-12
Q4.1E-13
Q3.7E-15
Q1.2E-11
Q3.1E-12
Q1.1E-11
Q1.1E-11
Q4.4E-11
Q5.1E-11
Q5.1E-11
Ql.OE-12
Q2.3E-13
Soluble (s); insoluble (i)
-------
For mode (d), the resuspension factor is FI = lO^cm'1, and an adult
inhales 7300 m3/yr:
Q'5E-12 yCi/yr
Table 14 presents the exposures resulting from fallout.
36
-------
TABLE 14. FALLOUT DOSE ALGORITHMS
Organ Doae (rcm/yr)
to
Radioiaotopea
Tritium
Carbon- 14
Potasslun-40
Iron-59
Krypton-85
Rubldluo-87
Strontlum-89
Strontlum-90
Ruthenlum-103
Ruthenlum-106
Iodine-129
Iodlne-131
Iodlne-131m
Cesium-134
Cesium- 137
Sanarium-147
Lead-210
Radon-222
Radium-226
Thoriu«-232
Uranlum-235
Uranium-238
Plutonlum-238
Plutonlun-239
Plutonlum-240
Plutonlum-241
Polonium-210
Soluble (a);
t
Estimate
race
2E-1
5.5
--
4E-4
—
4E-4+
2E-1
2E-1
1E-2
1E-2
2E-2
2E-2
~
2E-3
2E-3
4E-3+
2E-lf
—
2E-4f
2E-4+
2E-4*
2E-4f
1E-4
1E-4
1E-4
1E-4
2E-2f
insoluble
ore
C
•1
2E-2
1.5E-2
~
1.2E-3
—
~
1E-3
1E-3
1E-6
1E-6
1E-2
1E-2
—
5E-3
5E-3
~
—
—
—
—
—
—
—
—
—
—
—
(i).
Exposure Dose (uCi/yr) at 3
a
3.1Q'E-10
9Q'B-9
--
2Q'E-13
—
6Q'E-13
1Q'E-10
3.5Q'E-10
3.5Q'E-13
1.3Q'E-12
3.2Q'E-11
1.3Q'E-14
~
3Q'E-12
3.2Q'E-12
6.2Q'E-12
3.2Q'E-10
__
3.2Q'E-13
3.2Q'E-13
3.2Q'E-13
3.2Q'E-13
1.6Q'E-13
1.6Q'E-13
1.6Q'E-13
1.6Q'E-13
2.0Q'E-11
b
2.8Q'E-11
2.9Q'E-11
2.9Q'E-11
7.2Q'E-12
__
2.9Q'E-11
8.7Q'E-12
2.9Q'E-12
6.2Q'E-12
6.JQ'E-12
2.9Q'E-11
1.2Q'E-14
—
2.7Q'E-11
2.9Q'E-11
2.9Q'E-11
2.9Q'E-11
—
2.9Q'E-11
2.9Q'E-11
2.9Q'E-11
2.9Q'E-11
2.9Q'E-11
2.9Q'E-11
2.9Q'E-11
2.8Q'E-11
1.8Q-E-11
c
1.2Q'E-8
9.4Q'E-9
—
6.4Q'E-10
_M
—
S.SQ'E-IO
6.3Q'E-10
5.2Q'E-13
6.1Q'E-13
6.3Q'E-9
2.6Q'E-9
3.2Q'E-9
3.2Q'E-9
—
—
_ _
—
—
—
—
—
—
—
km
d
—
—
—
~
__
—
~
—
—
—
—
~
—
—
5Q'E-12
5Q'E-12
5Q'E-12
5Q'E-12
SQ'E-12
5Q'E-12
5Q'E-12
5Q'E-12
SQ'E-12
5Q'E-12
—
3 km 80 km Organ
a-c d a-c d a-c
A
2.6Q'E-12 — 10-'1 of Tissue(s)
1.4Q'E-11 — ? km Fat(a)
dose
7.5Q'E-11 — CI
2.4Q'E-11 — CI(i)
— — _•
5.7Q'E-13 —
2.4Q'E-10 —
3.4Q'E-8
1.4Q'E-13 —
1.5Q'E-12 ~
4Q'E-7
3.2Q'E-8
2.2Q'E-10
9.6Q'E-11 --
2Q'E-12 9.6Q'E-10
6.5Q'E-9 2.5Q'E-10
-_
4.7Q'E-10 7.5Q'E-9
5.2Q'E-11 1.9Q'E-9
3.5Q'E-10 7.1Q'E-9
3.5Q'E-10 7.1Q'E-9
2Q'E-11 2.7Q'K-8
2Q'E-11 3.2Q'K-8
2Q'E-11 3.2Q'K-8
4.7Q'E-13 6.3Q'E-10
__
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Bone (a)
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CI(i)
GI(i)
Thyrold(s)
Thyroid (a)
—
Total body(a)
Total body(s)
Bono(s)
Kidney (s)
—
Bone(s)
Bone(s)
Bone (a)
Hone(s)
Boni1 (s )
Bone(s)
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Bone (a)
l.OQ'E-10 — Splci-n(s)
d
—
—
~
—
—
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—
--
—
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Kidney (a)
—
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I.unR(i)
Lung(l)
Bone (a)
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Bonc(s)
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— ™
-------
IV ENERGY SYSTEMS
A. Coal for Direct Combustion
1. Coal Characteristics and Locations
The various types of coal are characterized by a number of parameters
including heating value, ash content, moisture content, sulfur content, and the
division of the organic portion of the coal into fixed carbon and volatile
matter (volatile matter is that portion that can be driven off by heating the
coal in a closed vessel; the fixed carbon is that portion which remains behind).
The four main ranks of coal are anthracite, bituminous, subbituminous, and
lignite.
^ Anthracite coals have heating values ranging from 13,000 to 15,000
Btu/lb and are characterized by high fixed carbon (85% to 90%) and low moisture
(2 to 5%) content [Ref. 19]t. Over 96% of the nation's anthracite reserves
are found in Pennsylvania.
The largest quantity of coal reserves is classified as bituminous.
These coals have heating values from 12,000 to 15,000 Btu/lb, and fixed carbon
content ranging from 45% to 78% [Ref. 19]. Moisture content ranges from 5%
to 15%. The bulk of bituminous coal reserves lies in the eastern half of the
United States.
Subbituminous coals occur mainly in the western half of the country.
They have a low fixed-carbon content (37% to 45%) and heating values from 8000
to 11,000 Btu/lb. Moisture content can range from 18% to 35% [Ref. 19].
The youngest coals, geologically, are the lignites that are
characterized by high moisture content (around 40%), low fixed-carbon content
(25% to 30%), and low heating values (6000 to 7500 Btu/lb) [Ref. 19]. The
*
The familiar English units are used in this coal section—one Btu/lb equals
2326 joules/kilogram.
All heating values, fixed carbon contents and moisture contents reported
here are on a mineral-matter-free basis. Mineral matter, or ash content
varies widely from about 3% to over 30%, with typical values being about
5% to 10%.
38
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major lignite deposits occur in eastern Montana, western North Dakota, and
Texas.
The important parameters of coal composition for the purposes of this
study are the uranium and thorium content. These parameters have not been
measured to nearly the same extent as have the parameters previously discussed.
Nevertheless, some measurements have been made, and these can serve as indicators
of the levels of radioactive pollutants to be expected from coal processing and
conversion activities.
In the 1950s, when extensive exploration for new uranium deposits
was being carried out, it was discovered that some coal deposits in the western
United States contained relatively high concentrations of uranium. The highest
concentrations were found in lignite and the second highest in subbituminous
coal. A small lignite deposit in South Dakota was found to contain up to 0.73%
uranium [Ref. 20]. More typical figures for coal deposits containing higher
than usual concentrations of uranium are 0.005% to 0.01% (50 to 100 ppm).
High concentrations of uranium in coal are the exception rather than
the rule, however. In fact, the concentrations of radioactive elements in
most coals are reported to be less than in common sedimentary rock. The uranium
content of eastern bituminous and anthracite coals rarely exceeds 0.001% (lOppm).
The occurrence of higher concentrations of uranium in western coals
is quite dependent on the geological history of the region. The most likely
way in which uranium accumulates in coal is through ground water containing
uranium dissolved from overlying volcanic rocks, or derived from hydrothermal
sources, percolating through the coal beds, which capture it. Consequently,
uranium-bearing coal typically occurs where the structure of adjacent rocks is
permeable to water.
The U.S. Geological Survey Bulletin 1055 [Ref. 21] reported on some
uranium-bearing coals in areas of the west that are considered likely candidates
for future coal development. In southeastern North Dakota, some thin lignite
seams (about 2 feet* thick) were found to contain an average of 0.013% uranium.
The extent of the deposit sampled was about 27 million tons*. In eastern Montana
(Carter County), a deposit of 16.5 million tons of lignite 1.5 to 8 feet thick
was found to contain an average of 0.005% uranium. Neither of these deposits
is considered economically minable at present.
In Sweetwater County, Wyoming, a much larger deposit of subbituminous
coal was surveyed. A deposit lying within 75 feet of the surface, averaging
about 2.5 feet thick and containing 700 million tons of coal, was sampled.
The average uranium content was found to be 0.003%, with some localized concen-
trations ranging as high as 0.051%.
One foot equals 0.3048 m; one ton equals 907.2 kg.
39
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The highest concentrations of uranium in coal reported in Bulletin
1055 were found in the LaVentana Mesa area, Sandoval County, New Mexico. Con-
centrations of up to 0.62% uranium were found, with the average concentration
being 0.1%. This figure applies to a relatively small deposit of 132,000 tons.
As mentioned previously, these high concentrations of uranium are
not typical and were discovered as part of a planned search for uranium-bearing
minerals. If these values are not typical, then the question arises as to what
typical values are. Most coal deposits in the United States have not been tested
for their trace element content. However, over the past several years, concern
over the environmental effects of trace element (including radionuclide) emissions
from coal burning facilities has led to a number of trace element determinations.
Coals from existing mines and from deposits expected to be mined in the future
have been sampled and analyzed. Examples of determinations of uranium and
thorium content for various coals, as reported in the literature, are shown
in Table 15.
The figures in Table 15 represent average values; however, values for
individual measurements within the same coal deposit can vary widely. For
example, the values of uranium and thorium concentration for Powder River coals
varied from 0.3 to 1 ppm and 1 to 3 ppm respectively [Ref. 22].
Because most coal deposits have not undergone testing for radio-
active element concentration, it is impossible to determine how typical the
values reported in Table 15 are. However, as these samples were taken from
actively producing coal areas, or areas marked for future production, they
TABLE 15. URANIUM AND THORIUM CONTENT OF VARIOUS COALS
(ppm by Weight)
Location and Type of Coal Uranium Thorium References
Appalachia (bituminous) 1.1 2.0 23
S. Illinois-West Kentucky 2.2 2.1 24
(bituminous)
Power River Basin-Wyoming 0.7 1.9 22
(subbituminous)
Navajo Reservation-New Mexico 1.2 4.8 25
(subb ituminous)
Kaiparowits Plateau-Utah 0.7 1.6 26
(bituminous)
40
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appear to be satisfactory for the purpose of calculating radionuclide emissions
from coal conversion processes. In carrying out such calculations, it should
be realized that wide excursions from the values reported in Table 15 are possible,
and, in particular, much higher values can be encountered.
Another naturally occurring radioactive isotope of interest in coal
conversion processes is potassium-40 (K-40). Potassium is a commonly occurring
element in coal and potassium oxide (K20) is a minor constituent of coal ash.
Concentrations of ^0 vary widely, as does the ash content itself. However,
some average values can be determined and used for calculational purposes.
Table 16 shows typical 1^0 concentrations in the coals listed, along with the
corresponding K-40 content. The K-40 content of the coal is based on the natural
abundance of this isotope (0.0118%) in potassium.
As in the case of uranium and thorium content, the K-40 concentration
can vary widely in coal samples taken from the same deposit. For example, the
K-40 concentration of Powder River Basin coal samples can vary from 0.015 ppm
to 0.052 ppm [Ref. 22].
TABLE 16. POTASSIUM-40 CONTENT OF VARIOUS COALS
Potassium Potassium-40
Location and Type of Coal Oxide (%) (ppm) References
Appalachia (bituminous) 0.13 0.13 27
S. Illinois-W. Kentucky 0.19 0.18 24
(bituminous)
Powder River Basin-Wyoming 0.031 0.030 22
(subbituminous)
Navajo Reservation-New Mexico 0.13 0.12 28
(subbituminous)
Kaiparowits Plateau-Utah 0.043 0.042 26
(bituminous)
2. Mining Methods
The methods employed to extract coal from the many deposits located
in the United States depend on the geological characteristics of the deposits,
including coal depth, seam thickness, extent of the deposit, and so forth.
Generally, coal mining is classified as either surface mining or underground
mining, depending on whether the overburden is removed prior to recovery of the
coal, or whether the overburden is left in place and the coal removed by under-
ground operations.
41
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a. Surface Mining
Surface mining can be carried out whenever the coal seam lies
close enough to the surface to allow economical recovery of the coal. The
economics of the operation is determined primarily by the thickness of the over-
burden in relation to the thickness of the coal seam. Overburden-to-seam thick-
ness ratios of up to 30 can be economically surface-mined [Ref. 29].
In Appalachia, where coal seams tend to outcrop on hillsides, the
most common surface-mining method is contour mining. With this method, a cut
is made in the hillside above the coal seam, exposing the seam. The coal is
removed by power shovels, which load it onto trucks. The cut is continued into
the hillside until the overburden-to-seam thickness ratio becomes too high. The
cut is continued along the hill in this manner and can extend for many miles.
The relatively flat terrain of the midwestern and northern Great
Plains regions allows a type of surface mining called area mining to be carried
out. With this method, a trench is excavated to expose the coal seam, with
overburden removed by a large device called a walking drag line. The coal is
removed by power shovels or front-end loaders and loaded onto trucks. As the
coal is being removed, an adjacent, parallel cut is excavated. As this proced-
ure is extended, the overburden is returned to mined out areas. Another version
of area mining is open-pit mining in which a large pit is excavated and grad-
ually widened to continually expose new coal. This method is especially appro-
priate to the very thick coal deposits found in the western United States.
Reclamation of surface mines can be carried out by grading over
the back-filled areas, replacing the topsoil, which has been suitably stored,
and either reestablishing a permanent vegetative cover or planting farm crops.
b. Underground Mining
Until recently, underground mining was the predominant method
of coal mining in the United States. Now, however, the total tonnage of
surface-mined coal surpasses that obtained by underground mining [Ref. 4].
Nearly all underground coal mining can be characterized as
room and pillar mining. The name derives from the fact that a series of "rooms"
are excavated in the coal seam, with "pillars" of coal left in place to support
the roof of the mine. The rooms are formed typically by drilling, loading and
blasting the seam and removing the coal by a loading machine. A relatively
new innovation is a machine called a "continuous miner" that both scrapes the
coal from the exposed face and loads it onto a conveyor belt, thus eliminating
the need for blasting. After a room is formed, the roof is further strength-
ened by drilling holes in it and inserting bolts to generate compressive
stresses.
The only other type of mining used to any extent in the United
States is called long-wall mining. To carry out long-wall mining, two tunnels
about 600 feet apart are excavated. Then a shearing machine works the coal
42
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face back and forth between the tunnels. Hydraulic jacks support the mine roof
and are moved forward as mining progresses, allowing the roof to collapse.
Long-wall mining allows a much higher recovery of the coal in place than does
room and pillar mining. However, it accounts for only a small percentage of the
underground coal production. The occurrence of surface subsidence tends to be
much greater with long-wall than with room and pillar mining.
3. Coal Preparation
About 50% of the coal mined in the United States undergoes some sort
of cleaning procedure before it is used. [Ref. 3] Nearly two-thirds of the
coal burned in electric power plants is cleaned.
The first step in coal cleaning is breaking and sizing. In this opera-
tion, the coal is crushed to some preset upper limit of size. Coal is crushed
in rotary mills and then screened to separate the coal that has been sufficiently
crushed. Oversized coal is recycled to the crusher.
After crushing, the coal may be cleaned by a number of methods. The
purpose of cleaning is to remove any foreign matter, such as dust, rock, shale,
and so on, that is associated with the coal. The simplest method of coal clean-
ing is air washing, in which air is simply blown over the coal to remove dust
and small particles.
There are several types of wet washing, all of which essentially make
use of the difference in specific gravity between the coal and any foreign matter
that might be present. Wet-washing methods include water/magnetite slurry
flotation, gravity separation in a pulsating washbox, and froth flotation. On
a nationwide basis, coal washing results in the removal of about 23%, by weight,
of the run-of-mine coal [Ref. 29]. Most of the reject material is foreign
matter, but about 5% of the feed coal is also removed [Ref. 29].
4. Coal Transportation
Once the coal has been mined, and possibly cleaned on-site, there are
several modes of transportation that can be used to deliver it to the point of
use. If the site of the power plant is near the mine, the coal is typically
moved by conveyor belt or truck. For long distance transport, railroads are
the most common method of shipment. About 70% of all coal shipped in the
United States is carried by train [Ref. 29]. When the mine and the power plant
are near navigable waterways, barges can be used. This method accounts for
about 11% of the coal shipped [Ref. 29]. About 11% is shipped by truck [Ref.
29], and the remainder is recounted for by conveyors, tramways, or private
railroads.
Another mode of coal shipment that has not yet come into widespread
use is the coal-slurry pipeline. With this method, coal is first finely ground,
then slurried with water so that It can be shipped through a pipeline much like
crude oil. At the receiving end, the coal is dewatered with centrifuges, where-
upon it is ready for burning. Only a single pipeline is in operation, shipping
coal 270 miles from a mine at Black Mesa, Arizona, to a power plant in Nevada.
Several longer pipelines have been proposed.
43
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5. Electrical Power Generation
About 45% of the electrical power generated in the United States
is produced in coal-fired power stations [Ref. 30]. As the demand for electri-
city grows and the nation attempts to reduce its reliance on petroleum and
natural gas, the use of coal as a fuel in electricity generation will increase
correspondingly.
The technology for producing electricity from the combustion of coal
has been well established for many years. The basic steps involved are the
combustion of the coal in a furnace, the capture of the combustion heat by a
boiler that produces high-temperature steam under pressure, the expansion of the
steam through a turbine that drives a generator, and the condensation of the
steam exhausted by the turbine. The overall thermal efficiency of this process
has increased in recent years to as high as 40% with modern boiler and heat-
exchange technology.
In addition to the main components listed above, additional components
required for electric power stations include heat-dissipation devices, such as
evaporative cooling towers and cooling ponds; stack gas cleanup equipment, such
as electrostatic precipitators and scrubbers; and coal preparation equipment.
These additional units are also consumers of power, and their use may reduce
the coal-to-electricity conversion efficiency by up to 10% [Ref. 29].
Typical sizes of new power-generating stations are on the order of
1000 MWe, although facilities of up to 3000 MWe have been proposed. These large
stations are designed to be base-load facilities, supplying power almost contin-
uously; peak load requirements are met by older, smaller plants and gas turbines.
New source performance standards for coal-fired boilers, promulgated
by the EPA, limit the quantities of pollutants emitted by these facilities. For
participates, sulfur dioxide, and nitrogen oxides, EPA new source performance
standards are 1.2, 0.1, and 0.7 Ib, respectively, per 106 Btu of coal burned.
Because different coals have varying ash and sulfur content and heating value,
the level of control required is dependent on the type of coal used. It also
depends on the type of coal-firing employed. For example, cyclone furnaces
release only about 50% of the ash in the coal as fly ash, whereas conventional
tangentially fired furances release 80% to 90%.
The radiological impact of coal combustion depends to a large extent
on the environmental controls required to limit the emissions of other pollutants,
because it is the release of radionuclides with the stack gases that will have
the greatest potential effect.
The siting of large coal-fired power plants is a complex matter
involving the availability of water for cooling, the location of the coal deposits,
the location of demand centers, the potential for violating air quality standards,
and so forth. In the eastern states, plants have tended to be sited near demand
centers since coal transportation distances are not large, and stringent air
pollution controls have been only recently required. In the west, the recent
trend has been to site coal-fired plants near the resource and transmit the
electricity to load centers via high voltage power lines. Due to existing air
44
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quality problems in may urban areas, the siting of new coal-fired power stations
in or near urban areas is virtually impossible.
Advanced technologies for converting coal to electricity promise
increased efficiency and greater environmental acceptibility. Among these are
the low-Btu gasification of coal followed by combustion in a gas turbine-steam
turbine combined cycle, and fluidized-bed combustion of coal. These and other
technologies are under development by ERDA and private companies, with the
expectation that they will be introduced commercially by the mid to late 1980s.
6. People
Both the number of people employed in the various facilities and the
population density in the surrounding areas are important for assessing the
occupational and general population health effects of radiological emissions
from fossil-fuel technologies.
Employment levels for coal mining vary widely due to the variations
in depth and thickness of the seam, type of mining method employed, and so forth.
The population density adjacent to mining areas varies even more widely from
relatively high densities for eastern mines to relatively low densities for
western mines.
Employment levels required for the operation of coal-fired power plants
and the population densities adjacent to these plants also vary widely. Many
plants are located within or adjacent to large urban complexes and others are
located in sparsely populated regions.
7. Radiological Aspects
The emission of radionuclides from the combustion of coal and the
associated coal fuel cycle is strongly dependent on the type of coal used and
its content of uranium and thorium, their daughter products, and potassium.
As shown in the Tables 15 and 16, these properties vary from one coal region
to the next, so that each region should be considered separately.
In the calculation of radionuclide emissions, it will be assumed
that all uranium and thorium daughter products are in secular equilibrium. This
is generally a good assumption due to the very old geological age of most coal
deposits. In some cases, there has been evidence of selective leaching of some
daughter products with repsect to the parent uranium or thorium. However, in
the absence of any particular evidence for the coals under consideration, secular
equilibrium will be assumed.
The examination of radiological effects will proceed in a step-.by-
step method beginning with coal mining and ending with coal combustion.
45
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a. Mining
The predominant radionuclide emission from coal mining is radon,
which is released during the exposure of coal seams and the breakup of coal
being mined. Any operation that increases the surface area of the coal exposed
to the atmosphere will result in the additional diffusion of radon into the air.
Coal mining can produce individual coal pieces ranging in size from fine powder
to several feet in length. The amount of radon released will vary inversely
with the size of the mined coal pieces.
The amount of radon initially present in the coal is a function
of the uranium and thorium content, because the isotopes radon-222 and radon-220
are members of the uranium-238 and thorium-232 decay series, respectively.
Assuming secular equilibrium, the radon isotopes should have the same activities
as the parent uranium and thorium isotopes. The concentrations of radon-222 and
radon-220 in the coals for which uranium and thorium determinations have been
made are shown in Table 17.
TABLE 17. RADON-222 AND RADON-220 CONCENTRATIONS
IN VARIOUS COALS
Activity
Concentration (pCi/g coal)
Location and Type of Coal Rn-222 Rn-220 Rn-222 Rn-220
Appalachia (bituminous) 2.4 x 10~18 2.4 x 10~22 0.37 0.22
S. Illinois-W. Kentucky 4.8 x 10~18 2.5 x 10~22 0.73 0.23
(bituminous)
Powder River Basin-Wyoming 1.5 x 10"18 2.3 x 10~22 0.23 0.21
(subbituminous)
Navajo Reservation-New Mexico 2.6 x 10~18 5.7 x 10~22 0.40 0.52
(subbituminous)
Kaiparowits Plateau-Utah 1.5 x 10~18 1.9 x 10~22 0.23 0.17
(bituminous)
It is extremely difficult to calculate the exact amount of radon
that is released when coal is mined. However, some rough assumptions can be
made that will give an order-of-magnitude estimate as to how much is released.
First of all, the concentration of radon in coal is extremely low, as can be
In subsequent discussion, only the uranium-238 and thorium-232 decay series
will be included, because the low natural abundance of uranium-235 (0.715%)
results in an activity for this isotope and its daughters of only 4.5% of
uranium-238 and its daughters.
46
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observed from Table 17. These concentrations amount to approximately 1 atom
of radon per gram of coal, or 1400 atoms per 1000 cubic centimeters. At these
low concentrations, the rate of diffusion is likely to be very small, because
the driving force behind diffusion is a concentration gradient. If we assume,
as an upper limit, that all the radon within 1 cm of the surface of a coal chunk
is released during mining, and that the average coal chunk is about a foot (30 cm)
across, then approximately 20% of the radon would be released during mining.
In the calculation of the amount of radon released during the
mining of the coal necessary to support a 1000 MWe power plant, certain values
are assumed for the heating values of different coals. A typical heating value
for Powder River coal is 8200 Btu/lb; for Navajo coal, 8500 Btu/lb; for Utah
coal, 10,800 Btu/lb; for Illinois coal, 11,000 Btu/lb; and for Appalachian coal
12,500 Btu/lb. On the basis of these heating values and an assumed thermal
efficiency of 35%, the amounts of coal that have to be mined to support the
operation of a 1000-MWe power plant are: Powder River coal, 14,300 tons per day;
Navajo coal, 13,800 tons per day; Utah coal, 10,800 tons per day; Illinois coal,
10,600 tons per day; and Appalachian coal, 9400 tons per day.
Table 18 shows the activity of the radon-222 and radon-220
released per day when the above levels of mining occur, for 1%, 5%, and 20%
release of the radon trapped in the coal. For comparison, it should be noted
that the average natural release rate of radon-222 above soil is 150 yCi per
acre* per day.
TABLE 18. RATE OF RELEASE OF RADON FROM COAL MINING
IN SUPPORT OF A 1000-MWe POWER PLANT
Release Rate of 1%, 5%, and 20%
of Radon Trapped in Coal (pCi/day)
1% 5% 20%
Location and Type of Coal Rn-222 Rn-220 Rn-222 Rn-220 Rn-222 Rn-220
Appalachia (bituminous) 32 19 160 94 630 380
S. Illinois-W. Kentucky 70 22 350 110 1400 440
(bituminous)
Powder River Basin-Wyoming 30 27 150 140 600 550
(subbituminous)
Navajo Reservation- 50 65 250 330 1000 1300
New Mexico
(subbituminous)
Kaiparowits Plateau-Utah 23 17 110 83 450 330
(bituminous)
One acre equals 4047 m2.
47
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Radon-222 has a half-life of 3.82 days, so that when it is emit-
ted it will be swept downwind from the mine where it will ultimately decay
through the series polonium-218->lead-214-*bismuth-214->polonium-214-»-lead-210, the
latter product having a half-life of 22 years. Radon-220 has a half-life of only
56 seconds, however, so that most of it will decay while still in the mine
environment. The solid (particulate) daughter products polonium-216-»-lead-212->-
bismuth-212->polonium-212-*-thallium-208 will adhere to the dust particles generated
by mining activities, most of which will settle out near the mine.
In underground mining operations, the presence of radon-222,
radon-220 and their daughters in the mine atmosphere may pose an occupational
health hazard. The concentration of these species in the mine atmosphere de-
pends not only on the uranium and thorium content of the coal, but also on the
ventilation rate of the mine.
In 1975, a survey was made of the presence of radon-222 and
radon-220 daughters in 223 operating underground mines in 15 coal-producing
states [Ref. 31]. According to this survey, there appears to be no significant
health hazard from inhalation of radon-222 and radon-220 daughters in under-
ground mines.
One final possible route of exposure for radionuclides in uranium-
bearing coal is mine acid drainage. This is a phenomenon by which groundwater
drainage through exposed coal seams converts pyrite materials in the coal to
sulfuric acid. In the Appalachian coal region, the sulfuric acid content of
mine drainage is as high as 1700 ppm. Because uranium is capable of forming
soluble sulfate complexes in the presence of sulfuric acid, the leaching of
uranium from coal is a potential problem. Alpha activity in excess of 10 pCi/1,
originating from dissolved uranium, has been reported in the drainage from some
eastern Pennyslvania coal fields where the uranium concentration ranged from
10 to 140 ppm [Ref. 32]. To meet EPA-mandated standards, mine-drainage water
is usually treated with lime to neutralize the acidity. With this treatment,
many of the metals in solution are precipitated out if the pH is allowed to rise
above 7. The formation of a low solubility uranium oxide hydrate in basic
solution could help to reduce the alpha activity of the mine drainage.
b. Coal Preparation
The main effluents from coal preparation are the solid wastes
consisting of rejected noncoal material, and the aqueous effluent termed
"blackwater." Up to 2 tons of blackwater containing 4% to 5% suspended coal
fines can be generated per ton of coal washed [Ref. 29].
The solid waste, averaging 23% by weight of the mined coal, is
typically stored on the surface in large piles. Although runoff from these
piles can pose a serious general water pollution problem (if proper impoundment
techniques are not employed), unless the content 'of radioactive species in these
materials is significantly different from that of the soils in the area, radio-
active contamination of surface waters should not be a problem.
48
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The blackwater from coal washing can be sent to a tailings pond
where the solids are allowed to settle out, and the clarified water can be
recycled to the coal preparation plant. It is possible that leakage from such
tailings ponds could contaminate local groundwater. If the tailings pond water
were acidic due to a high content of pyritic material in the coal, then signi-
ficant quantities of those elements, including uranium, could be dissolved in it.
The extent of possible contamination of ground water is almost impossible to
quantify.
The thermal drying of coal to reduce its moisture content is a
source of emissions to the air. However, the numerical details are left to
those sections in this report in which coal drying is included as an integral
requirement of the coal conversion operation.
c. Coal Combustion
The dominant source of radioactive emissions in the coal-to-
electricity fuel cycle will be the electrical power generating stations.
All the radioactive elements in the coal entering these generating stations will
ultimately exit in the form of waste materials discharged to the environment—
either as emissions to the atmosphere or as solid waste.
With the exception of radon, all the emissions of radionuclides
to the atmosphere will be associated with the discharge of fly ash resulting
from coal combustion. To meet the EPA new source performance standards of 0.1
Ib particulates per million Btu fired, the fly ash discharge must be reduced
by at least 98%. This requirement effectively limits the release of radio-
nuclides as well. The use of electrostatic precipitators or venturi scrubbers
to remove fly ash from stack gases can result in removal efficiencies of 99.5%
or greater. In fact, several proposed new coal-fired power plants have been
designed to use equipment of this efficiency even though the equipment exceeds
EPA requirements. This situation primarily results from stricter state emission
standards, or standards for plume opacity.
Table 19 shows the quantities of particulate matter (fly ash)
emitted in the operation of a 1000-MWe power plant for each type of coal
considered in this chapter. Typical ash contents of the coals are shown along
with the particulate emissions associated with both the EPA standard and a
99.5% removal efficiency. The calculations of the emissions were made under the
assumption of a 35% net power plant thermal efficiency, full capacity operation,
and an 85% conversion of coal ash to fly ash. It is clear from the table
that 99.5% control is sufficient to meet the EPA standard by a large margin for
all coals except Navajo coal from New Mexico. Due to its high ash content, this
coal would require a removal efficiency of about 99.6% to meet the standard.
We will first consider the release of radon-222 and radon-220,
which are the only radionuclides released as gases. In the preparation of
the coal for firing, it is first pulverized, then blown into the boiler with a
stream of air. It is probable that during pulverization most of the trapped
radon is released, and that any remaining would be released during combustion.
49
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TABLE 19. PARTICULATE EMISSIONS FROM A 1000-MWe
COAL-FIRED POWER PLANT
Coal
Appalachia
Illinois-W. Kentucky
Powder River
Navaj o
Utah
Btu/lb
12,500
11,000
8,200
8,500
10,800
% Ash
9
11
6
25
7
Tons of Particulates/Day
0.1 lb/10b Btu99.5% Control
12
12
12
12
12
3.7
5.0
3.6
15
3.2
TABLE 20. RELEASE OF RADON ISOTOPES FROM
A 1000-MWe POWER PLANT
Coal
Radon Release
(yCi/day)
Appalachia
Illinois-W. Kentucky
Powder River
Navajo
Utah
Rn-222
Rn-220
3140 1870
7050 2220
2980 2720
5000 6500
2260 1670
Because these operations take place in a totally enclosed system, most of the
released radon will be swept into the boiler and exit the facility along with
the stack gases. Table 20 shows the quantities of radon isotopes released
for a 1000-MWe power plant, if total release of all radon present in the coal
is assumed. These values should be considered as upper limits, due to the
potential for radon losses prior to combustion. Also, in the case of radon-220,
due to its short half-life, some decay to daughter products will occur during
and after combustion, and these will adhere to ash particles most of which will
be captured by the fly ash control equipment.
50
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The emission of other radionuclides in the uranium-238 and the
thorium-232 decay series, as well as of potassium-40, can be calculated in a
straightforward way by assuming that all these elements adhere to the ash when
the coal is combusted, and subsequently follow the same pathways as the ash
itself—10% to 20% going to bottom ash with the remainder emitted as fly ash,
most of which is captured. Based on these assumptions, the total quantities
of radionuclides emitted from a 1000-MWe power plant are shown in Table 21.
A particulate removal efficiency of 99.5% has been assumed, except in the case
of Navajo coal where a slightly higher efficiency has been assumed so that the
EPA standard of 0.1 pound of particulates per million Btu is met. The emission
levels of uranium and thorium daughters in Table 21 assume, of course, that
secular equilibrium is maintained throughout the combustion and emission pro-
cesses.
In practice, a number of phenomena will cause the actual emission
levels of some radionuclides to differ from those shown in Table 21. The most
obvious of these phenomena is the disturbance of secular equilibrium. In the
case of radon-222 and its daughters, this disturbance will probably be small
because of the length of the half-lives of radon-222 (3.82 d) and its immediate
daughters polonium-218 (3.05 min), lead-214 (26.8 min), and bismuth-214 (19.7
min), compared to the brief time scale of coal combustion and exit via tall
stacks—on the order of a few tens of seconds. Radon-220 (56 s), however, and
its daughter polonium-216 (0.14 s) have relatively short half-lives compared
to this time scale. The observed effect would be values of polonium-216 and
lead-212 enhanced above secular equilibrium, and a reduced value of radon-220
compared to that shown in Table 20. The species below lead-212 in the decay
series will not be appreciably affected because of the relatively longer half-
life of lead-212 (10.6 h).
The other phenomenon that can significantly alter the values
shown in Table 21 is the observed tendency for some elements to become concen-
trated in the fly ash relative to the bottom ash, and to be further concentrated
in the particulate matter exiting the fly ash removal equipment relative to that
entering. Several recent studies have indicated that uranium, lead, and polonium
isotopes tend to be concentrated in fly ash in this fashion [Ref. 24, 33, 34].
Thorium, radium, and potassium do not appear to be concentrated to any appre-
ciable degree. The degree of concentration of any element is dependent on the
type of boiler employed (cyclone or tangential) as well as on the devices used
to remove fly ash from the stack gases (electrostatic precipitator, venturi
scrubber, or cyclone).
The tendency of the elements uranium, lead, and polonium to
concentrate in fly ash could significantly increase the radioactivity of emit-
ted particulates because 7 out of 14 members of the uranium-238 decay series are
isotopes of these elements, as are 3 out of 11 of the thorium-232 series. Iso-
topes of other elements in these series have not been tested. The increase in
emitted radioactivity resulting from this concentration can be calculated using
some average concentration factors. As a rule of thumb, a factor of 5 can be
used to express the concentration of uranium, lead, and polonium in emitted fly
ash relative to the factor of 1 obtained by assuming that these elements are
evenly distributed in all the ash. This rough number averages over several
types of boiler and fly ash collection configurations, and in some configura-
51
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TABLE 21. EMISSION OF RADIONUCLIDES IN PARTICULATE
MATTER FROM A 1000-MWe POWER PLANT:
NO ELEMENTAL CONCENTRATION IN FLY ASH ASSUMED
(pCi/day)
Coal Type
Radionuclide
Uranium-238
Thorium-234
Pro tac t inium-2 34
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
Thoriiim-232
Radium-228
Actinixntt-228
Thorium-228
Radium-224
Radon-220
Polonium-216
Lead-212
Bismuth-212
Polonium-212
Thallium-208
Potassium-40
Total
Appalachia
13.3
13.3
13.3
13.3
13.3
13.3
*
13.3
13.3
13.3
13.3
13.3
13.3
13.3
7.9
7.9
7.9
7.9
7.9
*
7.9
7.9
7.9
5.1
2.8
30.0
290
Illinois-
W. Kentucky
30.0
30.0
30.0
30.0
30.0
30.0
*
30.0
30.0
30.0
30.0
30.0
30.0
30.0
9.4
9.4
9.4
9.4
9.4
*
9.4
9.4
9.4
6.0
3.4
47.2
522
Powder River
12.7
12.7
12.7
12.7
12.7
12.7
*
12.7
12.7
12.7
12.7
12.7
12.7
12.7
11.6
11.6
11.6
11.6
11.6
*
11.6
11.6
11.6
7.4
4.2
10.5
280
Navajo
17.5
17.7
17.5
17.5
17.5
17.5
*
17.5
17.5
17.5
17.5
17.5
17.5
17.5
27.6
27.6
27.6
27.6
27.6
*
27.6
27.6
27.6
17.7
9.9
33.4
509
Utah
9.6
9.6
9.6
9.6
9.6
9.6
*
9.6
9.6
9.6
9.6
9.6
9.6
9.6
7.1
7.1
7.1
7.1
7.1
*
7.1
7.1
7.1
4.5
2.6
11.2
200
See Table 20.
52
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tions uranium would be less concentrated than lead, for example, and in others
the opposite would be true. However, this number will serve to illustrate the
effect.
In calculating increases in emitted radionuclides, it is important
to note that only concentrations of relatively long-lived isotopes will be
affected; concentrations of short-lived isotopes would result only in rapid
decay back to equilibrium values. Thus, concentrations of polonium-214 (164 ys),
polonium-216 (0.15 s), and polonium-212 (0.3 ys) would not increase emitted
radioactivity.
Using the factor of 5 discussed above, along with the preceding
qualifications, the calculated emission of radionuclides from a 1000-MWe power
plant are shown in Table 22. Elements other than uranium, lead, and polonium
are assumed to exhibit no concentration in emitted fly ash.
As can be seen from Table 22, the assumptions made about the
concentration of uranium, lead, and polonium in emitted fly ash result in an
increase in emitted particulate matter by a factor of more than two, for all
the coals considered. An important point to be noted in relation to the
elemental concentration effect is that the particulate matter emitted by collec-
tion devices of various kinds tends to be very fine—about one micron or less
in diameter. It is precisely these particles that are not filtered out during
respiration and that tend to become lodged in the lungs and bronchial passages.
Thus, those particulates on which certain radioactive elements tend to concen-
trate are precisely those most capable of delivering radionuclides to suscept-
ible organs.
The ultimate disposition of the remaining radioactive elements
not emitted from the stacks will be the ash disposal pond, which will receive
the 99+% of the original ash in the coal recovered as bottom ash and removed
from the stack gases. In the case of power plants located near western surface
mines, the ash can be returned to the mine and buried.
Because the ash will be more or less permanently contained in
the disposal ponds, there is little chance of further airborne release of
radionuclides, with the exception of radon. The ash disposal piles will contain
in secular equilibrium all the elements in the uranium-238 and thorium-232
decay series, including radium-226 and radium-224 which decay to radon-222 and
radon 220, respectively. Because it is possible for radon to diffuse out of the
ash piles once it is formed, radon emission could be a problem. This applies
particularly to radon-222, because radon-220 p'robably decays before a signifi-
cant diffusion can take place.
To calculate the maximum impact of radon emissions from ash
storage areas, we assumed that ash was collected and stored over the 30-year
lifetime of a 1000-MWe power plant at a depth of 30 feet with a density of
60 Ib per cubic foot [Ref. 26]. Table 23 shows the resulting amount of radium-
226 present in the ash, assuming secular equilibrium with the parent uranium-
238; the area of the pile; and the rate of radon-222 release if all radon-222
formed from the decay of radium-226 is released. It was also assumed that the
power plant was operated at an average capacity factor of 75%.
53
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TABLE 22. EMISSION OF RADIONUCLIDES IN PARTICULATE
MATTER FROM A 1000-MWe POWER PLANT:
CONCENTRATION OF URANIUM, LEAD, AND POLONIUM IN FLY ASH ASSUMED
(yCi/day)
Coal Type
Radionuclide
Uranium-238
Thorium-234
Protactinium-234
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
Thorium-232
Radium-228
Actinium-228
Thorium-228
Radium-224
Radon-220
Polonium-216
Lead-212
Bisnmth-212
Polonium-212
Thallium-208
Potassium-40
Total
Appalachia
66.5
13.3
13.3
66.5
13.3
13.3
*
66.5
66.5
13.3
13.3
66.5
13.3
66.5
7.9
7.9
7.9
7.9
7.9
*
7.9
39.5
7.9
5.1
2.8
30.0
625
Illinois-
W. Kentucky
150.0
30.0
30.0
150.0
30.0
30.0
*
150.0
150.0
30.0
30.0
150.0
30.0
150.0
9.4
9.4
9.4
9.4
9.4
*
9.4
47.0
9.4
6.0
3.4
v47.2
1280
Powder River
63.5
12.7
- 12.7
63.5
12.7
12.7
*
63.5
63.5
12.7
12.7
63.5
12.7
63.5
11.6
11.6
11.6
11.6
11.6
*
11.6
58.0
11.6
7.4
4.2
10.5
631
Nava j o
87.5
17.5
17.5
87.5
17.5
17.5
*
87.5
87.5
17.5
17.5
87.5
17.5
87.5
27.6
27.6
27.6
27.6
27.6
*
27.6
138.0
27.6
17.7
9.9
33.4
1110
Utah
48.0
9.6
9.6
48.0
9.6
9.6
*
48.0
48.0
9.6
9.6
48.0
9.6
48.0
7.1
7.1
7.1
7.1
7.1
*
7.1
35.5
7.1
4.5
2.6
11.2
459
See Table 20.
54
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The maximum emanation rate of radon-222 as shown in Table 23 is
10.5 Ci/day over 245 acres, or 42,900 yCi per acre per day. The average natural
background emanation from soil is 150 pCi per acre per day [Ref. 35]. Therefore,
it appears that radon emission from ash piles is a rather significant effect.
TABLE 23. MAXIMUM RADON-222 RELEASE FROM 30-YEAR
ASH STORAGE PILE FROM 1000-MWe POWER PLANT
Area of Pile Ra-226 Content Rn-222 Release
Coal Type (acres) (grams) (Ci/day)
Appalachia 176 26.2 4.6
Illinois-W. Kentucky 245 59.5 10.5
Powder River 179 25.4 4.5
Navajo 721 42.0 7.4
Utah 159 19.3 3.4
Even though it is not probable that all the radon formed within
the ash pile is released to the air, the effect could still be significant.
Assuming that the radon release percentage from coal ash piles is similar to
that from uranium mill tailings piles, i.e., about 5% [Ref. 8, 12], the release
rate is still 15 times the natural background rate for the maximum case (Illi-
nois-W. Kentucky coal) and 3.4 times natural background for the minimum case
(Navajo coal).
Another potential long-term problem resulting from ash disposal
is the contamination of groundwater through leaching of the ash piles. Radio-
active elements can be leached along with other constituents of coal ash, re-
sulting in radioactive contamination of groundwater. Although it is nearly
impossible to quantify the extent of such contamination, the problem should be
pointed out so that appropriate monitoring can take place. Steps taken to con-
trol ash pile leaching, such as the lining of disposal ponds with impermeable
material, will effectively control radioactive contamination as well.
B. Coal Gasification and Liquefaction
The use of the nation's vast reserves of coal to produce clean liquid and
gaseous fuels is viewed by many as a potential long-term solution to dwindling
supplies of petroleum and natural gas. Neither coal gasification nor liquefac-
tion are in commercial operation today. However, first generation coal gasifi-
cation technology is available, and two substitute natural gas (SNG) plants
based on Lurgi gasification technology are being planned for the Four Corners
area of New Mexico. Coal liquefaction technology is still in the R&D stages,
and commercial operations are not expected until the mid-1980s, at the earliest.
55
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1. Coal Gasification Technologies
The synthesis of high-Btu gas or SNG (primarily methane) from coal
involves two basic steps: first, the gasification of coal by reaction with
steam and oxygen to form a medium-Btu synthesis gas; and second, the catalytic
reaction of the hydrogen and carbon monoxide in the synthesis gas (after suitable
adjustment of the H2/CO ratio) to form methane. Fifteen separate processes
have been proposed, tested, or are in commercial operation to accomplish the first
step of coal gasification. These processes differ in the temperature and pres-
sure at which gasification takes place, the nature of the medium in which gasi-
fication reactions occur, the method of feeding the coal, steam, and oxygen into
the reactor, and so forth.
Beyond the gasification step, the conversion of synthesis gas to methane
follows basically the same steps for each process. These steps generally
include: (1) quenching or scrubbing of the raw synthesis gas to remove tars
and oil and/or particulate matter formed in the reactor, (2) CO shift, in which
the H2 to CO ratio is adjusted to 3:1 by reacting excess CO with steam to form
H£ and C02, (3) purification of the synthesis gas, involving primarily removal
of the acid gases C02 and ^S, (4) catalytic methanation to form the 900 to 1000
Btu-per-SCF (standard cubic foot) product gas, and (5) removal of water vapor
from the product gas and compression of the gas to pipeline pressure.
Numerous plans for using coal to produce SNG have been announced by a
number of pipeline companies. The plants closest to the construction stage are
those planned by El Paso Natural Gas and WESCO (Western Gasification Company)
for the Four Corners area of New Mexico. These plants are based on the Lurgi
gasification process and will produce approximately 288 million and 275 million
SCF per day, respectively, of pipeline quality gas.
In the Lurgi process, crushed coal is fed in batches into the top of
the reactor, and steam and oxygen are injected at the bottom. Gasification takes
place at high temperature and pressure (1400°F and 400 psi). Ash is removed
through a grate at the bottom of the reactor. The Lurgi gasifier is termed a
"fixed bed" reactor because the coal remains fixed in place as gasification is
carried out.
First generation plants plan to use the Lurgi process because it is
commercially available and its use has been reliably demonstrated over a period
of many years. Because of certain drawbacks in the operation of Lurgi gasifiers,
second or third generation gasification plants will undoubtedly use more ad-
vanced processes such as the ERDA-sponsored Sythane process, or the Institute
of Gas Technology (IGT) Hygas process, as these technologies become available
for commercial application.
2. Coal-Liquefaction Technologies
To date, no commercial coal-liquefaction plants have been planned
because the technology is not at a sufficient level of development to bring
a commercial plant on stream with any reasonable guarantee of success. However,
a number of pilot plants are being operated under sponsorship of ERDA or
56
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private industries, and several demonstration plants are in the planning stages.
It is generally believed in the industry that no full-sized commercial plants
(30,000 bbl/day or larger) will be on-line before 1985.
The processes for converting coal to liquid fuels fall into three
main categories. These are: gasification/Fischer-Tropsch synthesis, pyrolysis,
and coal dissolution. Gasification of coal followed by Fischer-Tropsch synthe-
sis to yield gasoline and other hydrocarbon byproducts is the only one of the
three processes that is currently in commercial operation—a 6600 ton/day plant
is being operated by the South African Gas and Oil Company (SASOL). Due to the
high costs and low efficiency of this technique, it is unlikely that it will be
used to any significant degree in the United States.
Coal pyrolysis is a technique for extracting oil, gas, and char (a
solid product consisting mainly of carbon and ash) from coal by heating it to
high temperatures in the absence of air. The COED (Char-Oil-Energy-Development)
process developed by the FMC Corporation has been tested in a 36 ton/day pilot
plant near Princeton, New Jersey. Other coal pyrolysis methods have been
tested at the pilot plant stage by Garrett Research and Development and The Oil
Shale Corporation. In all pyrolysis methods, the yield of oil is low compared
with the other methods for making liquid fuels from coal.
The process of coal dissolution (also known as solvent refining or
solvent extraction) has the greatest potential for efficiently producing liquid
hydrocarbon fuels from coal. The two basic steps in this process are the
dissolution of the organic matter in the coal in a process-derived solvent, and
hydrogenation of the resulting product to yield a liquid hydrocarbon fuel.
Several variants of the coal dissolution process that have been tested at the
pilot plant level are the Solvent Refined Coal (SRC) process of Pittsburgh and
Midway Coal Company, the Consol Synthetic Fuel (CSF) process of Consolidation
Coal Company, the H-Coal process of Hydrocarbon Research, Incorporated, the Gulf
catalytic coal liquids process, the Exxon donor solvent process, and the ERDA
Synthoil process. The H-Coal process appears to be one of the most promising
for further development, and considerable analysis has been done on it using
information that has been made public.
In the conversion of coal to liquid fuels via the H-Coal process,
dissolution and hydrogenation are carried out in the same step in the presence
of a catalyst. The slurried coal is reacted with hydrogen in an ebullating bed
reactor at 850°F and 2700 psi. A cobalt-molybdenum catalyst is continuously
added to the reactor as spent catalyst is removed. After gases and unreacted
solids are separated from the mixture, synthetic crude oil is recovered from
fractionation of the resulting liquid.
Initial testing of the H-Coal process has been carried out in a 3-ton/
day pilot plant at the Hydrocarbon Research, Inc. (HRI) facilities at Trenton,
New Jersey, under the sponsorship of Ashland, ARCO, Standard of Indiana, and
Exxon. In addition, ERDA and HRI are planning a 600-ton/day pilot plant at
Catlettsburg, Kentucky, to test the commercial feasibility of the H-Coal process.
Industrial sponsors include those mentioned above (except Exxon), the Electric
Power Research Institute, and Sun Oil.
57
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In all of the coal dissolution processes discussed above, large
amounts of hydrogen (15,000 to 20,000 SCF per ton of coal) are consumed. In
general, the hydrogen can be supplied by gasifying unreacted coal solids (char)
or by steam reforming byproduct hydrocarbon gases. If necessary, additional
hydrogen can be provided by gasifying some of the feed coal.
3. People
Although coal gasification and liquefaction are not particularly
labor-intensive processes, the large sizes of the facilities envisioned will
require substantial numbers of employees at any particular site.
El Paso Natural Gas Company estimates that about 880 employees will be
required to operate its proposed 288 million SCF/day SNG plant in New Mexico
[Ref. 36]. Western Gasification Company estimates a somewhat lower figure,
610, for its proposed 275 million SCF/day plant [Ref. 25].
For coal liquefaction, the plant size characteristic of a mature
industry would be on the order of 100,000 barrels of liquid products per day.
This size facility would employ about 1400 people [Ref. 37].
4. Radiological Aspects
a. Coal Gasification
There are two potential radiological problems in coal gasifi-
cation. One is the emission of radionuclides at the plant site, and the other
is possible contamination of the product SNG which will later be used in homes
and industries. We will deal with the latter question first, because it can be
handled with a simple calculation. All the analysis in this section will be
based on the Bureau of Reclamation's Environmental Impact Statement (EIS) for
Western Gasification Company's (WESCO) proposed 275-million-SCF/day plant in
New Mexico [Ref. 36]. Although a similar EIS has been carried out for El Paso's
proposed plant [Ref. 25], the WESCO EIS is somewhat more detailed and more
suitable for the purposes of this section. The coal to be used in the plant is
the Navajo Reservation subbituminous coal whose properties were listed in
Section IV-A. At this time, no other detailed analyses of coal gasification
based on other coal types have been published.
The problem of radioactive contamination of the product SNG is
essentially a problem of radon contamination, because virtually all particulate
matter is removed from the gas during processing. Gasification of the coal is
carried out in an enclosed, pressurized vessel, and probably all the radon in
the coal is released during gasification and enters the synthesis gas. The
various physical and chemical processes that are subsequently employed to clean
the gas and convert it to methane will have little effect on the radon, since
radon is an inert gas. It will be carried through the gasification and synthe-
sis steps, as are the small amounts of nitrogen and argon present in the oxygen
used for gasification.
58
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To calculate the maximum amount of radon present in the product
SNG, we assume that all the radon initially present in the coal fed to the
gasifiers emerges in the product. In the WESCO plant design, 24,820 tons of
coal would be converted to 275 million SCF per day. With a radon-222 activity
of 0.40 pCi/g for Navajo coal (see Section IV-A), 9.0 uCi will enter the gasi-
fiers each day. (Radon-220 is not considered due to its short half-life.) The
resulting maximum concentration in the product SNG will thus be 33 pCi/SCF or
1.2 pCi/1.
The significance of this concentration can be assessed in terms
of the amount of radon-222 present in natural gas reservoirs. Various surveys
of natural gas wells have indicated radon-222 concentrations of 0.2 to 1450
pCi/1, with an average concentration of 37 pCi/1 [Ref. 38]. Because these
levels of radon contamination have been shown to have insignificant health
effects on natural gas users [Ref. 38], we may conclude that the potential health
impact of 1.2 pCi/1 of radon-222 in SNG is negligible.
The emission of radionuclides at the plant site is calculated
in a fashion similar to that for electric power plants, that is by examining
the disposition of the mineral matter (ash) component of the coal, realizing
that most of the uranium-238, thorium-232, their daughters, and potassium-40
will be incorporated in it. The bulk of the emissions of radionuclides to the
air will occur from the coal-fired steam boilers that provide steam to the
gasifiers. These boilers consume 3870 tons/day of washed coal and discharge
0.82 tons/day of particulate matter to the air, after stack gas clean-up opera-
tions. A minor secondary source of emissions is a steam superheater which burns
tar oil—a byproduct of gasification—and emits 0.019 tons/day of particulate
matter. A small amount of coal dust (0.016 tons/day) is emitted from coal-
handling equipment controlled by fabric filters or cyclones.
The rest of the ash that enters the plant is recovered as bottom
ash from the gasifier (6530 tons/day), or bottom ash plus collected fly ash
from the steam boiler (610 tons/day), and is returned to the mine for burial.
Approximately 560 tons of ash that is present in the coal rejected from the
boiler coal-washing facility is removed to a disposal pond rather than buried
with the ash.
Small amounts of ash and coal dust that exit the gasifier along
with the synthesis gas are removed by a scrubber. The scrubber effluent is
treated for removal of phenols, ammonia, and hydrogen sulfide, then sent to a
biological treatment/settling pond where impurities are removed so that the
waste water can be recycled. The ash and coal fines that settle out in this
pond become part of the sludge that is eventually removed to the mine and buried.
It is possible that the scrubber effluent might be slightly enriched with lead
and polonium isotopes, since these volatile elements tend to vaporize to some
extent in the gasifier, and are condensed in the scrubber. The 3.5 million
gallons of treated waste water would contain about 5 ppm of suspended solids
[Ref. 39]. If this water is discharged rather than recycled to the evaporative
cooling towers, its radioactivity level would be 0.2 pCi/1, assuming all solids
are coal ash.
59
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To calculate radon emissions, we can assume that all the radon
in the boiler feed coal is released, part during crushing, and the remainder
during combustion. For the gasifier feed coal, the only source of radon emis-
sion will be crushing, since radon released during gasification will be retained
in the product SNG. Because the gasifier feed coal is crushed (to less than
1-1/2 in) rather than pulverized, we may assume, as an upper limit, that 50%
of the radon in this coal is released.
The emissions of radionuclides resulting from the sources discuss-
ed above are shown in Table 24. The concentration effect in boiler emissions
for isotopes of uranium, lead, and polonium has been incorporated into the
calculations, using the factor of 5 discussed in Section IV-A. The emissions
have been divided into the categories of dust, gases (radon), and particulates.
The reason for this division is that dust emissions consist primarily of heavy
particles that settle out quickly, resulting in exposures that are mainly
occupational. Radon and particulate emission, however, will be carried far from
the plant site, resulting in general population exposures.
Even though the WESCO plan calls for burial of the ash in mined-
out areas, other coal gasification plants operated at long distances from the
mines may find this alternative uneconomical. Therefore, it is important to
calculate the radon emission resulting from disposal of the ash on the surface.
Using the same assumptions as in Section IV-A, a 30-year accumulation of ash
from a 275-million-SCF/day facility operating at 90% capacity would occupy 1850
acres and emit a maximum (assuming 100% release) of 19 Ci/day of radon-222,
or 10,300 uCi per acre per day, compared to the natural background rate of 150
yCi per acre per day.
As was mentioned in the discussion of radioactive emissions from
coal-fired power plants, the disposal of coal ash, either on the surface or in
mined-out areas, is a potential source of groundwater contamination. This
problem will have to be examined on a site-by-site basis, however, due to vary-
ing disposal methods, soil conditions, rate of precipitation, and so on.
b. Coal Liquefaction
Detailed analyses of the operations of [Ref. 40] and pollutant
emissions from [Ref. 37] the H-Coal process have been carried out for two dif-
ferent types of coal—Illinois bituminous, and Powder River subbituminous. The
calculation of radiological effects from coal liquefaction will be based on
these analyses.
An H-Coal plant producing 100,000 bbl/day of synthetic crude
oil will consume 55,177 tons/day of Powder River coal, with 4807 tons of
this amount consumed in steam boilers, and 2717 tons consumed in the coal dryer.
The remaining 47,653 tons is pulverized, dried, and fed to the liquefaction
facility. It is assumed that all the radon contained in the coal entering the
plant is released during either pulverization of the liquefaction feed coal or
combustion of the coal used as fuel.
60
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TABLE 24. RADIONUCLIDE EMISSIONS TO THE AIR
FROM A 275 MILLION STANDARD CUBIC FEET PER DAY COAL
GASIFICATION FACILITY USING NAVAJO COAL
(MCi/day)
Radionuclides
Uranium-238
Thorium-234
Protactinium-234
Uranium-234
Thorium-230
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
Thorium-232
Radium-228
Actinium-228
Thorium-228
Radium-224
Radon-220
Polonium-216
Lead-212
Bismuth-212
Polonium-212
Thallium-208
Potassium-40
Total
Dust
0.006
0.006
0.006
0.006
0.006
0.006
0.006
0.006
0.006
,006
,006
0.006
0.006
Gases
0.
0.
0.008
0.008
0.008
0.008
0.008
0.008
0.008
0.008
0.005
0.003
0.011
0.16
6,200
8,100
14,300
Particulates
6.0
1.2
1.2
6.0
1.2
1.2
6.0
6.0
1.2
1.2
6.0
1.2
6.0
1.6
1.6
1.6
1.6
1.6
1.6
8.0
1.6
1.0
0.6
2.4
68
61
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The other major sources of radionuclide releases to the air are
the emission of particulate matter from the steam boilers and coal dryer, and
the emission of coal dust from the coal dryer. All of these facilities are
assumed to have controls on particulate emissions. The furnaces for the coal
dryer and steam boilers are equipped with electrostatic precipitators which
remove 99.5% of the particulate matter, resulting in the emission of 0.63 and
1.1 tons/day of particulates, respectively. The emission of coal dust from the
dryer is 99% controlled by cyclones and a venturi scrubber, resulting in the
emission of 4 tons/day of coal dust. In addition to these sources, about 0.68
tons/day of particulate matter is emitted as a result of the combustion of
hydrocarbon gases derived from the liquefaction process. This material is
assumed to consist primarily of coal ash. The releases of radionuclides from
the above sources are shown in Table 25.
About 1.2 million gallons/day of waste water are generated in
the coal hydrogenation unit. This water is treated for the removal of hydrogen
sulfide and ammonia and sent to biological treatment ponds for further purifi-
cation. Any suspended solids, such as coal ash, in this waste water would be
largely removed by such treatment. In water-poor regions of the west, the
treated water would be recycled to the evaporative cooling towers where most of
it would be evaporated, with the contaminated blowdown stream being discharged
to evaporation ponds.
If not returned to the mine for burial, the 30-year ash pile
would occupy 830 acres and emit a maximum of 20.9 Ci/day of radon-222, or
25,200 yCi/acre.
It is possible that a small amount of ash will appear in the
syncrude product. This amount has not been quantified for the H-Coal process.
However, based on the data for the SRC process for producing clean boiler fuel
from coal [Ref. 41], the ash content is expected to be no greater than 0.1%.
Thus, the uranium-238, thorium-232, and potassium-40 contents of the product
would be no greater than 0.012, 0.032, and 0.0005 ppm, respectively.
In the production of synthetic crude oil from Illinois coal via
the H-Coal process, 38,241 tons of coal are pulverized, dried, and fed to the
liquefaction plant to produce 100,000 barrels of synthetic crude oil per day.
The major difference, in terms of radionuclide emissions, between using Illinois
and Powder River coal is that in the former case the production of byproduct
hydrocarbon gases is sufficient to supply all the plant fuel needs without add-
itional coal having to be burned.
The only emissions to the air are 1.0 tons/day of coal dust from
the controlled (99.8%) coal dryer, and 1.2 tons/day of particulate matter from
combustion of byproduct gases. All the radon in the feed coal can be assumed
to be released during pulverization and drying. The resulting radionuclide
emissions are shown in Table 26.
About 770,000 gallons/day of waste water will be produced in the
coal hydrogenation unit. If the plant is located in the eastern United States,
then this waste water will probably be discharged after suitable treatment as
62
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TABLE 25. RADIONUCLIDE EMISSIONS TO THE AIR
FROM A 100,000 BBL/DAY COAL-LIQUEFACTION FACILITY
USING POWDER RIVER COAL
(yCi/day)
0.
0.
Radionuclide Dust Gases
Uranium-238 0.84
Thorium-234 0.84
Protactinium-234 0.84
Uranium-234 0.84
Thorium-230 0.84
Radium-226 0.84
Radon-222 — 11,500
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
Thorium-232
Radium-228
Actinium-228
Thorium-228
Radium-224
Radon-220 ~ 10,500
Polonium-216
Lead-212
Blsmuth-212
Polonium-212
Thaliutn-208
Potassium-40
Total 18 22,000
Particulates
.84
,84
0.84
0.84
0.84
0.84
0.84
0.76
0.76
0.76
0.76
0.76
370
63
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TABLE 26. RADIONUCLIDE EMISSIONS TO THE AIR
FROM A 100,000 BBL/DAY COAL-LIQUEFACTION FACILITY
USING ILLINOIS COAL
(yCi/day)
Radionuclides Dust Gases
Uranium-238 0.66
Thorium-234 0.66
Protactinium-234 0.66
Uranium-234 0.66
Thorium-230 0.66
Radium-226 0.66
Radon-222 — 25,300
Polonium-218 0.66
Lead-214 0.66
Bismuth-214 0.66
Polonium-214 0.66
Lead-210 0.66
Bismuth-210 0.66
Polonium-210 0.66
Thorium-232 0.21
Radium-228 0.21
Actinium-228 0.21
Thorium-228 0.21
Radium-224 0.21
Radon-220 — 7,900
Polonium-216 0.21
Lead-212 0.21
Bismuth-212 0.21
Polonium-212 0.13
Thallium-208 0.08
Potassium-40 1.1
Total 12 33,200
Particulates
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.2
2.3
2.3
2.3
2.3
2.3
2.3
2.3
2.3
1.5
0.8
12.0
126
64
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described previously. This treatment is capable of reducing the suspended
solids concentration to low levels, on the order of 10 ppm. Assuming that coal
ash is the sole constituent of the suspended solids component of the waste water,
the total radioactive content of discharged waste water would be 1.2 pCi/1.
The 30-year ash pile from a 100,000-bbl/day plant would occupy
1060 acres and release a maximum of 45 Ci/day of radon-222, or 42,700 pCi per
acre. The maximum concentrations of uranium-238, thorium-232, and potassium-40
in the synthetic crude oil product would be 0.020, 0.019, and 0.0016 ppm,
respectively.
C. Oil Shale
The major attraction of oil shale development is the large quantity of
resource potentially available for recovery. It is estimated that recoverable
reserves of oil shale in Utah, Wyoming, and Colorado containing at least 30
gallons of kerogen per ton of shale, in beds at least 30 feet thick, can yield
over 130 billion barrels of oil [Ref. 42], This is 20 times the current annual
U.S. consumption, and an amount equal to the total estimated conventional
U.S. oil reserves. However, the cost of recovering this oil is high. On the
average, 1.4 tons of shale must be mined, crushed, processed, and disposed of for
each barrel of oil recovered. Thus, an oil shale plant producing 100,000
barrels of oil per day must handle 50 million tons of shale per year. This is
ten times the amount of material mined in the largest underground coal mines
now in operation.
1. Mining
It is anticipated that most of the oil shale lying in underground
deposits will be mined by the room-and-pillar technique described earlier.
With this method, about 60% of the resource in-place can be extracted and 40%
is left in the form of pillars.
When oil shale lies in deposits near the surface, open pit mining
can be carried out. The overburden is first stripped away and stored, then the
shale is recovered, crushed, and retorted. After all the resource is removed
from the mine area, the overburden is replaced, contoured, and revegetated. The
feasibility of surface mining oil shale is determined by the overburden-to-
resource ratio and the availability of an area for overburden storage.
2. Conversion
Conceptually, the technology for obtaining liquid hydrocarbons from
oil shale is simple. The crushed shale is heated in a closed vessel (retort)
to a temperature of 900°F or greater, at which point the kerogen (the organic
portion of the oil shale) vaporizes and is separated from the solid inorganic
portion of the rock. After retorting, the shale oil is upgraded by means of
hydrotreating (chemically reacting with hydrogen) to yield a synthetic crude
oil that is suitable for transport via pipeline and can be used as a refinery
feedstock.
65
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The various methods for retorting oil shale differ in the manner in
which heat is generated and transferred to the shale. The simplest method is
the Fischer assay technique in which heat from an external source is transferred
to the shale through the wall of the retort. Any fuel may be used to supply
the heat. Due to large capital and operating costs, this method is unsuitable
for commercial development. However, it is commonly used on a laboratory scale
to measure the kerogen content of the shale.
There are four additional methods for retorting oil shale; they are in
various stages of development and have the potential for commercial application.
These four methods are discussed in the following paragraphs.
a. Gas-to-Solids Heating/Internal Gas Combustion Method
In this method, crushed shale is fed to the top of a vertical
retort and low-Btu byproduct gas is injected at the bottom. The gas is com-
busted in the retort along with residual carbon on the spent shale, and the
hot combustion gases heat the shale, driving off the oil vapors that are con-
densed at the top of the retort. The noncondensible gases are recycled for
combustion. Due to the lack of external heating equipment, this method is less
costly than other types of retorts. Energy recovery efficiency is somewhat
lower, however.
The Union Oil Company version of the process uses a unique "rock
pump" that injects shale at the bottom of the retort while combustion gases are
drawn down from the top by blowers, and retorted shale oil is collected at the
bottom. An advanced version of this retort, called the steam gas recirculation
(SGR) process, was recently announced, and a 1500-ton/day demonstration plant
based on this process will be built on private land in Colorado.
Another variation on the process has been constructed by
Development Engineering, Inc. (DEI), the operating arm of Faraho Development
Corporation (a consortium of 17 firms). This process, usually referred to as
the Paraho retort, uses patented shale feed and spent shale discharge grates,
which provide a uniform flow of shale through the retort. Multilevel gas injec-
tors are also used to carefully control the level of incoming gases. DEI has
completed a successful run on its 500-ton-per-day test plant near Rifle,
Colorado, as part of a 30-month R&D program. Faraho has also proposed the con-
struction and testing of a commercial-size retort on the Naval Oil Shale Reserve
in Colorado. Both of the planned commercial operations on federally leased
tracts in Utah have proposed using the Faraho retort.
b. Gas-to-Solids Heating/External Heat Generation Method
In this method, recirculated byproduct gas is used as the medium
of heat transfer; however, heating of the gas is carried out in an external
furnace, rather than by combusting the gas and spent shale within the retort.
Some of the byproduct gas, carbon residue on the spent shale, or any other
suitable fuel may be combusted to supply heat to the furnace. Paraho will
soon begin testing a version of its retort which operates with externally
heated gases.
66
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The Union Oil SGR retort mentioned earlier is actually an example
of the external heat generation method, although the original Union Oil techno-
logy was based on internal gas combustion.
c. In-Situ Retorting
In this process, shale rock is fractured in place by explosives
to form an underground retorting chamber. Air is injected to combust part of
the shale, and retorting is carried out by heat transfer from the hot combustion
gases. Shale oil is collected from a hollow mined at the bottom of the shale
column.
Numerous tests of this method have been made by various companies.
Commercial feasibility has not yet been demonstrated, although recent tests by
Garrett Research and Development, a subsidiary of Occidental Petroleum, appear
promising. If the Garrett or other tests demonstrate the commercial feasibility
of in-situ retorting, the use of this method could considerably reduce water
consumption, spent shale disposal, and other problems presently associated with
aboveground retorting. However, new problems, such as surface subsidence and
the release of large quantities of combustion gases, would be created, and these
would need to be carefully managed. This method is expected to be less costly
than any above-ground retorting technique.
d. Hot Solids or Solids-to-Solids Heating Method
The TOSCO II process developed by The Oil Shale Corporation
(TOSCO) is the most advanced version of this technique. In this process,
ceramic balls are heated by the combustion of byproduct gases and liquids and
transferred to the retort where they are mixed with crushed, preheated shale.
Shale oil vapor is driven off and recovered. The ceramic balls are separated
from the spent shale (on the basis of size) and subsequently reheated. A high
efficiency of energy recovery is achieved; however, capital and operating costs
are high.
The TOSCO II process is essentially ready for commercial appli-
cation. Colony Development Operation (a joint venture of ARCO, Ashland, Shell,
and The Oil Shale Corporation) has successfully completed tests on a 25-ton/day
test unit and an 1100-ton/day semiworks plant at Parachute Creek, Colorado.
Colony had announced plans to begin construction in April 1975, of a 50,000-
bbl/day commercial plant based on the TOSCO II process. These plans were later
postponed, with Colony citing rapidly inflating construction costs and uncer-
tainties in U.S. energy policy as the basis for its decision.
There are several other planned commercial operations in which
the TOSCO II retort will be used. These include the following: a 50,000-bbl/day
plant planned by ARCO, TOSCO, Ashland, and Shell as a joint venture on Colorado
Tract C-b; the Rio Blanco Oil Shale Project, a joint venture on Colorado Tract
C-a by Gulf Oil and Standard of Indiana; and the 75,000-bbl/day Sand Wash Project
in Utah planned by TOSCO.
67
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The TOSCO II process is the most advanced retorting method for
which a sufficient amount of information is available to provide the emission
factors required for analysis. Thus, we have chosen to use it in our analysis
of oil shale conversion.
Subsequent to retorting, described previously, the shale-derived
gases and liquids must be processed to remove sulfur and nitrogen, and produce
a syncrude that is suitable as a refinery feedstock. The raw shale oil is
separated into naphtha, gas oil, and residual fractions. The naphtha and gas
oil are sent to separate hydrotreaters where they are upgraded and desulfurized.
The residual oil is sent to the coker unit, where.coke is produced along with
additional naphtha and oil, which are sent to the hydrotreaters. During hydro-
genation of the naphtha and gas oil, sulfur and nitrogen compounds are converted
to hydrogen sulfide and ammonia, which are separated in the sour water waste
stream and subsequently recovered as ammonia solution and elemental sulfur.
The hydrogenated naphtha and gas oil are recombined and leave
the plant as synthetic crude oil or fuel oil. The high-Btu byproduct gases
from the retort are purified to remove hydrogen sulfide and ammonia, and to
remove uncondensed liquids (naphtha). All of these gases are then consumed on
site, either as plant fuel to provide steam and heat, or as feed to the steam
reforming furnaces, where they are reacted to form hydrogen for the hydro-
treaters.
3. People
The mining, retorting, and upgrading of oil shale must be considered
to be part of a single integrated operation. This is because it is far too
costly to ship the raw shale any significant distance from the mine, and the
retorted shale oil is too viscous to be shipped by pipeline to be upgraded
elsewhere. Therefore, these operations will be carried out at or near a single
site. The sites will be located primarily in the oil shale-rich Piceance Basin
in northwestern Colorado, although some development will take place in north-
eastern Utah as well.
Colony Development Operation has estimated that a 50,000-bbl/day oil
shale mining retorting and upgrading operation will employ 900 to 1000 people
[Ref. 44]. A 100,000-bbl/day operation, which is the unit size analyzed in
the next section, would probably employ something less than twice this number,
or 1600 to 1800 people.
4. Radiological Aspects
The emissions of radionuclides from oil shale mining, retorting
and upgrading are calculated in the same manner as for coal conversion techno-
logies, and depend on the radionuclide content of the resource as well as the
emission of dust and particulate matter during conversion operations. The basic
data on the emissions of various kinds was taken from the Environmental Impact
Statement by the Bureau of Land Management for Colony Development Operation's
proposed oil shale facility in Colorado [Ref. 44].
68
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The uranium, thorium, and potassium concentrations in spent (retorted)
shale have been measured at 0.99 ppm, 0.77 ppm, and 2.72%, respectively [Ref.
44]. Allowing for the fact that the mineral matter content of raw oil shale of
35-gallon/ton grade (which would be used in the Colony project) is 82.6%, and
assuming that all uranium, thorium, and potassium remain in the spent shale during
retorting, the uranium-238, thorium-232, and potassium-40 concentrations in raw
oil shale would be 0.82, 0.64, and 2.7 ppm, respectively.
To support a 100,000-bbl/day oil shale facility, approximately 132,000
tons of raw shale would have to be mined, crushed, and retorted daily. It can
be assumed that all the radon trapped in the shale is released to the atmosphere
during these processes. The radon released into the closed retort would be
withdrawn along with the byproduct retort gases, and would be later released when
these gases are burned as plant fuel.
The emission of raw shale dust will occur at several points around the
processing area, including the mine vent, the primary and secondary crushers,
the crushed shale storage area, and the conveyor transfer points. The control
of dust release with fabric filters will result in the emission of 2.6 tons of
raw shale dust per day.
The main sources of emission of fine particulate matter will be in
the shale preheat system, spent shale moisturizer, and the ellutriator system.
In these areas, fine shale particles, along with hot gases, are mobilized and
vented from stacks controlled by venturi scrubbers. This process results in
the emission of 18.7 tons/day of particulate matter. In addition, the combus-
tion of byproduct hydrocarbon gases and liquids as fuels in various sections of
the operation will result in the emission of 0.50 tons/day of particulate matter,
for a total of 19.3 tons/day. All of this particulate matter may be assumed to
have the composition of raw shale.
The emission of radionuclides to the air resulting from the processes
discussed above are shown in Table 27.
All of the waste water generated in the oil shale operation will be
treated and reused on the site. No discharges to streams or rivers are planned.
This probably will be the case for all oil shale operations due to the scarcity
of water in the western Colorado region, and the resulting necessity of making
full use of the water that is available.
Approximately 113,000 tons of spent shale will be disposed of each
day. This material will be dumped into canyons, compacted and graded, and
ultimately revegetated. Due to the similarity of this material to the composi-
tion of the rock and soil of the surrounding area, it is probable that very
little additional radon emanation above that of the natural background of the
area will be induced by spent shale disposal operations.
Although saline runoff and leaching of salts from spent shale dis-
posal piles present potentially serious environmental hazards, the low levels
of radionuclides in these piles compared to those of neighboring soils should
not result in significant activity levels in the runoff and leachate.
69
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TABLE 27. RADIONUCLIDE EMISSIONS TO THE AIR
FROM A 100,000 BBL/DAY OIL SHALE
MINING, RETORTING, AND UPGRADING OPERATION
(yd/day)
Radionuclides Dust Gases
Uranium-238 0.64
Thorium-234 0.64
Protactinium-234 0.64
Uranium-234 0.64
thorium-230 0.64
Radium-226 0.64
Radon-222 ~ 32,800
Polonium-218 0.64
Lead-214 0.64
Bismuth-214 0.64
Polonium-214 0.64
Lead-210 0.64
Bismuth-210 0.64
Polonium-210 0.64
Thorium-232 0.16
Radium-228 0.16
Actinium-228 0.16
Thorium-228 0.16
Radium-224 0.16
Radon-220 — 8,300
Polonium-216 0.16
Lead-212 0.16
Bismuth-212 0.16
Polonium-212 0.10
Thallium-208 0.06
Potassium-40 40
Total 50 41,100
Particulates
4.9
4.9
4.9
4.9
4.9
4.9
4.9
4.9
4.9
4.9
4.9
4.9
4.9
1.3
1.3
1.3
1.3
1.3
1.3
1.3
1.3
0.9
0.4
310.0
385
70
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The presence of some of the trace elements occurring in raw shale
has been detected in the upgraded shale oil [Ref. 43]. The concentrations of
these elements have not been quantitatively assessed, however. It is probable
that uranium and thorium, if present in the shale oil product, would be found
only in very small concentrations—less than 0.1 ppm.
D. Geothermal Energy Systems
1. Introduction
The principal present commercial use of geothermal energy is the
production of electricity at The Geysers in California [Ref. 29,45]. Current
production capacity is approximately 502 MWe from 11 units, and four additional
units totaling 415 MWe are in the process of being approved or constructed.
This geothermal resource consists of dry steam, and has been estimated by
various sources to have an ultimate capacity ranging from 2000 to 10,000 MWe.
A reliable estimate of the extent of this resource is not available.
In addition to power generation, hot water resources are used for
space heating at Klamath Falls, Oregon, and Boise, Idaho. Additional uses
now under consideration include the desalination of water and food processing.
Geothermal resources may consist of dry steam, hot water, or a mixture
of the two depending on pressure and temperature conditions at the wellhead.
Resources are also classed according to the descriptions: hydrothermal,
geopressured, and hot rock. Hydrothermal reservoirs consist of a permeable
formation containing water and overlying a magma deposit. The molten rock
furnishes heat to the groundwater in the formation, and the heated water or
steam commonly appears naturally at the surface in the form of a hot spring or
fumarole. The formation may be recharged periodically with rainfall.
The heat in geopressured resources results when the clays in a rapidly
subsiding area trap heat in underlying permeable formations containing water.
In hot rock systems, no permeable formation overlies the heat source. Conse-
quently, the rock must be fractured and water injected to recover heat. Most
geothermal resources are in the western third of the country.
Because neither the geopressured nor the hot rock systems have been
exploited, no direct data are available on radioactive elements that may be
contained in the geothermal fluids. Until such data are available, experience
and data on hydrothermal systems may serve as an indicator of the potential
radioactivity in these fluids.
The fate of any radioactive contaminants in the geothermal fluids
will depend on their ultimate disposition. Because many fluids under consider-
ation contain other contaminants that may be harmful to the environment, re-
injection of the fluid into the formation is considered. The feasibility of
this step has to be established for each formation. Consequently, the ultimate
disposition of spent geothermal fluids cannot be determined. In the absence of
specific process plans, it may be assumed that any water-borne contaminants
produced would be discharged into surface waters, and that any gases produced
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would be discharged into the atmosphere. For example, in some present space
heating applications, the water is discharged into the city storm sewer system.
On the other hand, the present practice at The Geysers is an example of the
likely fate of radioactive contaminants released in the generation of electricity
from dry steam resources.
In the sections that follow, the energy system specifically described
will be an electric power plant. Many of the steps in the use of geothermal
power for electricity will apply equally to other applications. The major
functional difference is that a heat exchanger would typically replace the
turbine and condenser.
2. Exploration and Development
The exploration and development of geothermal resources resemble to
a large degree similar phases for petroleum or natural gas resources [Ref. 29,
45]. The various techniques for determining the location, size, quality,
density, and extent of geothermal resources will not be described in detail,
since radiation exposure from these activities is negligible in comparison to
activities involving very large quantities of geothermal fluids. Briefly,
test corings, electrical conductivity tests, thermal gradient surveys, surface
water analysis, seismic measurements, gravity and magnetic surveys are examples
of geothermal exploration activities. However, little direct knowledge of thermal
reservoir characteristics is obtainable without drilling, once a reservoir is
located.
The extraction system for geothermal fluids is similar to that for oil
and gas in that a well is drilled to sufficient depth, and is cased and completed
to provide a safe and stable conduit for the fluid. However, there are several
differences in details.
Once a well is completed, it may be tested at relatively high flow
rates in order to establish the productivity and suitability of the well for
commercial use.
After a steam well is completed and tested, a period of several
years may intervene before it is put into production, because of construction
and other delays. During this period it is slowly bled to prevent condensation
and the accumulation of noncondensible gases. This bleed rate is negligible
compared to production flow rates.
Geothermal wells may experience blowouts, which occur at the well-
head during drilling, or below the surface where the well intersects a permeable
channel. Blowouts are controlled with blowout preventers at the wellhead and
with the installation of suitable casing and cement fill in permeable zones.
The possibility of a blowout is greatest early in the exploration phase when
least is known of subsurface conditions. When a well is blowing out of control,
especially near the surface, flow rates may approach production levels.
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3. Production
If a suitable number of proven and tested wells are established in a
sufficiently small area, it will be economic to construct pipelines connecting
these wells to a facility that can absorb the thermal load and produce the
desired output. In considering radiological hazards, it is unimportant whether
the geothermal fluids are channeled through heat exchangers to provide space
heating, process drying, or electricity generation in a binary (two fluid) cycle
process, or through a turbine and condenser to generate electricity. The con-
trolling feature will be the disposition of the geothermal fluids after the
heat has been extracted.
An electric generating plant uses low pressure turbines. The resource
at The Geysers, dry steam, is characterized as 350°F (177°C) and about 100 psi
(6.9xl05 Pa). This compares to 1000°F (538°C) and 3,500 psi (2.4xl07 Pa) in
a modern fossil fuel plant. Consequently, overall efficiency is down to about
15%, compared to 35 to 38% for the fossil fuel plant. In other terms, the heat
rate is about 22,000 Btu (2.32xl07 J) per kWhe instead of about 10,000 Btu.
This implies that the now standard 55-MW generators each require about one million
pounds of steam per hour. About seven wells, averaging about 150,000 pounds
per hour, are required to produce this output.
If a unit goes out of production for short periods, the steam is vented
for operational reasons, rather than closed off at the wells. Consequently,
the flow of geothermal fluids is approximately constant regardless of electricity
production. During venting, the entire flow with all contaminants is injected
directly into the atmosphere at a site near the power plant.
During normal production, the steam flows in large pipes, to minimize
the pressure drop, from the wells to the turbine inlet. Centrifugal separators
in the lines near the well remove particulate matter that would exacerbate
erosion of the turbine blades. Present units at The Geysers have direct contact
condensers. The condensate is cycled through cooling towers and returned to the
condensers or reinjected. Approximately 80% of the fluids produced are evaporated
in the cooling process, and the remaining 20% are reinjected into the reservoir.
Future units will use surface condensers to facilitate control of the hydrogen
sulfide. The proportions of the fluids evaporated, however, would not materially
change.
In the present units, which have direct contact condensers, noncon-
densible gases present in the steam are removed with gas ejectors that exhaust
to the atmosphere. The relatively small quantities that remain in solution in
the condensate and are not stripped out in the cooling tower cycle would be
reinjected. In effect, essentially all the noncondensibles are emitted into
the atmosphere. In two of the present Geysers units, experimental hydrogen
sulfide control systems are installed. In these units, the gas ejectors vent
into the cooling towers, where an hydrogen sulfide absorber is added to the
cooling water circuit. Other noncondensibles continue to be emitted. In future
units, the gas ejector stream will be treated in a Stretford unit to control
the hydrogen sulfide. About 10% of the hydrogen sulfide might remain in the
condensate, and be stripped out in the cooling tower unless plans to treat the
condensate are altered. In summary, any gases present in the steam, other than
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hydrogen sulfide, end up in the atmosphere. They are emitted by the gas ejectors
or the cooling tower. A very small amount remaining in solution would be
reinjected into the reservoir.
Power plants that may be constructed to exploit hot water resources,
such as in the Imperial Valley in California, would differ from The Geysers
in several respects. In one scheme, depending on pressure and temperature
conditions at the wellhead, the geothermal fluids would flash into a mixture
of water and steam, perhaps in two or more stages depending on design. A typical
Imperial Valley well might have a mass flow rate three times as high as a Geysers
well, but less than a third would flash to steam at, say, 70 psi. Overall
plant efficiency might be 11%, compared to 15% at The Geysers. Due to the very
high dissolved-solids content (250,000 ppm), and other reasons such as land
subsidence, it would be desirable to reinject the water. However, the feas-
ibility of reinjection over long periods remains to be proven. In any event,
if experience at The Geysers is indicative, noncondensible gases would be
emitted to the atmosphere. Alternatively, a binary cycle process may be used
with hot water resources. Again, the disposition of the spent fluid is the
critical factor.
4. Radiological Aspects
The only radionuclide of interest in dry steam resources is radon-222
[Ref. 46, 47]. An average value for producing wells at The Geysers is 20,000
pCi per liter of condensed steam [Ref. 47]. For the standard 55-MW generating
plant and the conditions described above, the radon release would be 0.24 Ci/day,
or 4.4 Ci/1000 MW-day. Put another way, the release for a 55-MW plant is equiv-
alent to the natural release of radon from about 2.7 square miles (7xl06m2) of
land surrounding The Geysers [Ref. 47]. Samples of the ambient air around
present power plants and developmental fields have been measured for radon
content, and compare with normal levels found over the continental land area
[Ref. 47].
Hot water resources vary widely in radionuclide content, but 136
sites throughout the western United States have been recently [Ref. 48] sampled
for six nuclides—radon-222, radium-226, uranium-234 and -238, and thorium-230
and -232. Generally, the content of radon in pCi/1 was at least an order of
magnitude larger than those of the other nuclides. The importance of the other
nuclides would depend on the disposal plans for the waste water—specifically,
on the opportunity for the nuclides to accumulate selectively. It is difficult
to speculate on this issue, since disposal is already a problem because of the
high concentrations of dissolved solids. In addition, large scale development
of any of the sampled reservoirs could easily change the indicated concentrations.
The highest concentrations of radon were measured in the Imperial
Valley of California, an extensive area likely to undergo development sooner
than most. The concentrations there ranged up to 14,000 pCi/1. If this figure
is taken as representative for hot water resources, and we assume that essen-
tially all of the radon separates into the flashed steam, then the radon re-
leased in the generation of electricity in a plant like The Geysers would be
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17 Ci/1000 MW-day. Here, we have assumed that 25% of the hot water flashes to
steam, and we have used the power plant efficiencies cited above. This radon
emission rate would be nearly four times the rate at The Geysers for similar
plant size and emission assumptions.
Estimates of emission rates for other applications of hot water re-
sources depend directly on the assumed or planned fluid disposal practice. If
closed cycle containment and 100% reinjection can be demonstrated and maintained
for a site, emissions would be nil.
E. Nuclear Systems
The extraction of energy from nuclear fission involves several major pro-
cesses, each of which could be the source of unacceptable releases of radio-
active materials to the environment. The processes, referred to as the nuclear
fuel cycle, include uranium mining, uranium milling, conversion to uranium
hexafluoride, enrichment, fuel fabrication, reactor power plant operations, and
fuel reprocessing. In addition, the nuclear system requires transportation
services and has the problem of radioactive waste management. The throughput
requirements of the nuclear fuel cycle and the associated radioactive materials
generated depend on the combined efficiency of all the system components. For
example, the mining requirements per unit quantity of electrical power output
depends not only on the uranium content in the ore but also on the extraction
efficiency of the mill, the thermal efficiency of the reactor, and so on. In
general, the reported variations in these factors cause only minor quantitative
differences in the analytical results. Therefore, for computational purposes
the following typical values are used for light water reactors [Ref. 8,9,49].
The U-jOg content in the uranium ore is 0.2%
The mill extraction efficiency is 95%
• The conversion process loss is 0.25%
• The enrichment process loss is 0.05%
The enriched uranium is 3.3% uranium-235
• The depleted uranium is 0.3% uranium-235
The fuel fabrication process loss is 1%
• The uranium-235 fission energy output is 65%
• The plutonium fission energy output is 35%
• The thermal efficiency is 32.5%
• All the uranium-235 except process losses is recycled
• The reprocessing uranium loss is 1%.
The plutonium recovered in reprocessing is consigned for high temperature
gas reactor (HTGR) and liquid metal fast breeder reactor (LMFBR) use. If the
plutonium is recycled for light water reactor fuel, the uranium requirements
will be fractionally decreased.
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With regard to the use of natural thorium, thorium-232, in high temperature
gas reactors to produce fissile uranium-233 as a nuclear fuel, the processes and
problems associated with thorium-232 are similar to those of uranium-238; the
differences are generally quantitative only. For example, the thorium content
in thorium ores varies widely but is generally on the order of a few percent,
that is, about an order of magnitude higher than the uranium content in uranium
ore. The radioactivity associated with thorium-232 is also only half as much
as that associated with uranium-238. Because the uranium-233 produced must be
extracted through fuel reprocessing prior to use, it is anticipated that the use
of thorium in the next few decades will be minimal compared with the use of
uranium.
1. Uranium Mining
The Colorado Plateau (Arizona, Colorado, New Mexico, and Utah), is
the largest uranium source area containing approximately 54% of the resource
[Ref. 50]. (Another publication has estimated the uranium in this area to be
in excess of 90% of the United States total [Ref. 51].) An additional 42% is
located in the basins of Wyoming and the coastal plains of Texas. Uranium
mines are usually located in remote areas where the population density is about
2 to 4 people per square kilometer (5 to 10 people per square mile) [Ref. 51].
The uranium mining methods used in the United States are open pit
mining and underground mining. Although underground mines currently far out-
number open pit mines, open pit mines generally have higher production capa-
cities, and slightly more than half of the uranium produced is from open pit
mines. This near equality in total output is not anticipated to change
significantly in the future as about one-half of the uranium reserves is in
shallow deposits suitable for open pit mining and about one-half is located
deep underground where underground mining is more suitable. About 96% of the
reasonably assured uranium is in sandstone deposits.
Both open pit and underground mine sizes vary over a wide range.
The average capacity of operating open pit mines in the United States is about
1.5 x 108 kg of ore per year and the average capacity of operating underground
mines is about 2x 10' kg of ore per year. For comparison purposes, the ore
requirements per 1000 MWe year (9.7 x 1016 J thermal equivalent) are 1.3 x
108 kg.
The overburden for open pit mines varies but for calculative purposes
it is estimated at 30 kg per kilogram of ore. The ore is sandstone and carno-
tite (K2[U02]2[V04]2 ' 3H20). The total uranium content in 1.3 x 108 kg of ore
is 2.2 x 10s kg (0.2% ^Og). The radioactivity of this amount of uranium and
its daughters is 1040 Ci. Radionuclides other than uranium and its daughters
are not reported, but the total activity for three likely major radionuclides
(potassium-40, vanadium-50, and rubidium-87) is estimated to be relatively
insignificant. The radioactive contents in the overburden are not reported but
the total radioactivity for 3.9 x 109 kg of overburden is estimated as follows:
4.4 Ci of potassium-40, 7.4 mCi of rubidium-87, and insignificant vanadium-50.
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The solid waste from underground mining is estimated to be about
equal to the volume of the processed ore. The radioactivity associated with the
solid waste is not reported. The water pumped from both open-pit and underground
mines will contain dissolved and suspended uranium and its daughters. The
radioactivity in the mine water is estimated at 1% of that in the ore, but it is
merely returned to the mine by ion exchange as the water is permitted to seep
through the soil.
The airborne radioactive effluents from uranium mines are particulates
with uranium and daughter products and radon gas. The atmospheric releases of
radioactivity from open-pit mining are not reported, but because dust is
created and radon gas is released when the ore body is exposed and broken up by
open-pit mining as in underground mining, the atmospheric releases by the two
mining methods can be expected to be comparable. However, although the radon
gas emitted from a 750-ton/day (6.8 x 10^ kg/day) underground mine was calcu-
lated to be about 3 Ci/day [Ref. 52], open-pit mining atmospheric releases were
estimated to be minimal [Ref. 51]. The radon released in open-pit mining has
been estimated to be less than 0.01% of the total in the ore or about 7.5 mCi
per 1.3 x 108 kg of ore. For the average sized open pit mine with a capacity
of 1.5 x 10 8kg per year, the radon release rate would be about 3 nCi/s. This
may be compared to a calculated release rate of 20 yCi/s for an underground mine
with the same capacity. The large discrepancy in the estimated radon releases
bears investigation.
All the particulates in the effluent released to the atmosphere
eventually return to the earth's surface. The radon gas decays to radon daugh-
ters and these eventually return to the earth's surface as particulates.
An open-pit mine area is restored for other use by covering the
excavated area with the overburden that was previously removed. After restora-
tion, radiological limitations on the future use of the area can be established
by an assessment of its radiological state. The restoration of underground
mine areas for other use includes the sealing of all access to underground areas.
The accumulated waste-pile materials could be used for this purpose. Limitation
on future use of the site can be established by radiological assessment.
2. Uranium Milling
The relatively low concentrations of uranium in the ore makes it
economically imperative to minimize ore haulage distances; therefore, uranium
mills are located near uranium mines in relatively sparsely populated areas.
Mechanical and chemical processes are used to extract the uranium from the ore.
The major features of a uranium mill are an ore storage area, an ore crushing
building, an ore grinding building, chemical processing equipment including a
solvent extraction building and leaching and precipitation tanks, a building
for drying and packaging the output product, and a prepared location for receiv-
ing the mill waste. The problem of radiological controls exists throughout
the entire milling process from ore storage to final product packaging and mill
waste disposal.
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At the ore storage site, small amounts of uranium and uranium daughter
products in the form of dust and radon gas (mainly radon-222) are released to
the atmosphere. The amounts released are small because most of the radioactive
materials are bound within the individual pieces of ore. The ore is also wetted
to control the dust. The milling process is initiated with the crushing and
the grinding of the ore. Because the ore is still wet during crushing and the
grinding is a wet process, the dust created by these processes is also minimal.
Dust is also controlled by a dust collection system. The reduction of the ore
to small particle sizes, however, permits an increased rate of radon gas release.
The waste waters of these operations are piped to the retention pond. The chem-
ical processes that follow (leaching, solvent extraction, precipitation, liquid
removal, and calcining) all release radon to the atmosphere. The pulverizing
and packaging operations also produce radioactive dust. Control measures are
used to reduce the amount of radioactive dust escaping to the atmosphere.
The relative amounts of radioactive materials released by each milling
process are not reported. The radon released during the entire milling operation
is estimated to be between 20% and 80% of that in the ore. Most of the uranium
and daughter products, other than radon, that escape to the atmosphere, do so
during the drying process. Almost all of the other daughter products of uranium,
in solution or as part of the mill tailings, are pumped to the retention pond.
At the retention ponds, the solids settle, the liquid evaporates, and mill tail-
ings piles are created. These mill tailings piles are the major sources of
radioactive materials released to the environment from mill sites.
Uranium mill capacities vary widely. A mill output sufficient to
supply the consumption requirements of five 1000-MWe reactors at 100% utiliza-
tion or an output of 1.5 x 106 kg of yellowcake per year can be considered
tyoical. As specified, the milling requirements per 1000 MWe-year (9.7 x
10*6 J) of power is 1.3 x 108 kg of ore and the radioactivity of the uranium
and uranium daughters in this ore is 1040 Ci. The milling process converts
this input into the following outputs.
Yellowcake (NH4)2 U20? or (85% U30g)
Amount 2.9 x 105 kg
Uranium content 2.1 x 105 kg
Radioactivity 150 Ci (uranium-234, uranium-235,
uranium-238)
Effluents: 890 Ci
Solids
Amount 1.29 x 108 kg
Radioactivity 810 Ci (uranium and uranium daughters)
Fate Tailings pond
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Liquids
Amount 1.7 x 108 kg
Radioactivity ~4 Ci
Fate Tailings pond
Airborne
Amount
Radon ~75 Ci
Uranium and
uranium daughters <2 Ci (mostly from dryer)
Fate Return to earth
At the mill tailings site radon gas, radon-222, is formed from the
decay of radium-226, which has a half-life of 1622 years. Radium-226 in turn
is formed from thorium-230 and is in secular equilibrium with thorium-230,
which has a half-life of 8 x 101* years. Because this half-life is extremely
long relative to the tailings accumulation rate and the radon half-life of
3.8 days, the radon gas production rate at the tailings pile is essentially
proportional to the amount of radium-226 or thorium-230 that is pumped to the
tailings pile. The radon production rate at the tailings site will therefore
increase with each day of mill operations, and the radon escaping from the
tailings site can be expected to increase accordingly. After the cessation
of mill operations the radon escape can be expected to remain relatively con-
stant at the peak rate. It is estimated that about 5% of all the radon generated
at the tailings site is released to the atmosphere [Ref. 8,12]. The radon gas
then decays to its daughter products and is returned to earth with precipita-
tion.
The rate of radon gas escape to the atmosphere from the tailings
site of a 1.5 x 106 kg yellowcake mill after 10 years of operations is esti-
mated to be about 6 yCi/s. Besides the airborne releases, radioactive materials
in the liquid effluent at the mill tailings retention site could also make its
way into the environment through seepage. Although the movement of radioactive
material through seepage can be expected to be extremely slow, seepage surveil-
lance is required to prevent unacceptable contamination of local water sources.
Radon-222 is not the only radon gas that is generated at the tailings
site. A much smaller amount of radon-219, a daughter product originating from
uranium-235, is also generated. If the uranium ore contains significant amounts
of natural thorium, thorium-232, most of this thorium and its daughter products
will also be pumped to the tailings pond. In this case, radon-220 is also gen-
erated at the tailing site and some of this radon will also escape to the atmo-
sphere. The half-life of radon-220 is only 55 seconds, however, and the frac-
tion escaping the tailings can be expected to be small. The radon-220 that
escapes will quickly decay to stable lead.
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3. Conversion to Uranium Hexafluoride
The yellowcake, packaged at the mill, is shipped to a conversion plant
where the coarse l^Og is converted to uranium hexafluoride, UFg. Two different
processes are used to produce UFg. They are the "hydrofluor" process and the
wet solvent extraction process. UFg conversion plants are currently located
in low population density areas (5 to 10 people per square kilometer). A
typical UFg conversion plant has a capacity of about 7 x 106 kg of UFg per year.
This capacity is sufficient to produce 24,000 MWe-year of electrical energy
(2.33 x 1018 J thermal energy equivalent).
The inputs to the conversion plant are new uranium in the form of
yellowcake and recycled uranium in the chemical form U02(NOo)o " BH^O. The
total uranium requirements per 1000 MWe-year are 2.1 x 105 kg of make-up uranium
and 3.2 x 101* kg of recycle uranium. The radioactivity of this total amount
of uranium is about 170 Ci. Because the uranium conversion loss is only 0.25%,
all but about 0.4 Ci of the radioactivity is associated with the uranium con-
verted to UFg. Additional radioactive materials associated with the impurities
of the coarse U^Og inputs and the recycled uranium will also be part of the plant
effluents. The radioactive solid effluent is buried and the liquids are pumped
to storage ponds. A small fraction of the total radioactive effluents, about
0.01% of the throughput activity or about 0.016 Ci per 2.4 x 105 kg of uranium
processed, is released to the atmosphere. This is equivalent to a release
rate of about 11 nCi/s for the typical size conversion plant.
4. Uranium Enrichment
Enriched uranium fuel is required by light water reactors. The
uranium-235 enrichment required is about 3%. Highly enriched uranium is
required by high-temperature gas reactors for their initial fuel load. For
this application and for the start-up of liquid metal fast breeder reactors
the uranium-235 content in the total uranium is greater than 90%.
The uranium-235 enrichment method in use is isotopic separation by
gaseous diffusion. Other technologies such as centrifuge separation and laser
separation are being developed. The United States currently has three gaseous
diffusion plants; one at Oak Ridge, Tennessee; one at Paducah, Kentucky; and
one at Portsmouth, Ohio. The combined capacity of the three plants is about
1 x 107 kg of separative work units per year. This total capacity is equiva-
lent to an input of about 3.2 x 107 kg of UFg and an output of 4.5 x 106 kg
of 3.3% enriched UFg per year (sufficient to produce about 90.000 MWe-year of
electrical energy from light water reactors or about 8.7 x 10*8 J of thermal
energy). Current plans are to increase the capacities of these plants by a
factor of 2.5 by 1980 to meet projected domestic and foreign demands.
The enrichment process losses are estimated to be less than 0.05%.
This is equivalent to about 0.085 Ci per 1000 MWe-year of uranium fuel processed.
The total effluent radioactivity for all three enrichment plants operating at
capacity can therefore be estimated at about 8 Ci per year.
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The enrichment inputs per 1000 MWe-year fuel requirement are 3.55
x 105 kg of UF6 with uranium contents as follows: 1.75 x 103 kg uranium-235,
135 kg uranium-236, and 2.4 x 10s kg uranium -238. The electrical energy
consumed by the enrichment process per 1000 MWe-year fuel requirement is about
44 MWe-year. The outputs for the above inputs are 5.05 x 101* kg of enriched
UFe with 1.13 x 103 kg of uranium-235 and 3.3 x 101* kg of uranium-238, and
3.05 x 105 kg of depleted UFg with 610 kg of uranium-235, 136 kg of uranium-236,
and 2.05 x 105 kg of uranium-238. The enriched UF6 is shipped to fuel fabrica-
tion and the depleted UF6 is stored. The bulk of the radioactive effluents
is discharged to holding ponds as a liquid and then discharged to streams. A
small fraction of the effluent is airborne uranium and this is vented to the
atmosphere. The total atmospheric release rate for all three plants operating
at capacity is estimated at less than 25 nCi/s.
5. Fuel Fabrication
The fuel fabrication process converts the enriched UFg to complete
fuel assemblies for reactor installation. The current industry includes
plants that convert enriched UFg to U02 in either powder or pellet form,
plants that use the U02 as input to produce completed fuel assemblies, and
total process plants. The capacity of a relatively large fuel fabrication
facility is on the order of 1 x 106 kg per year. The uranium losses for fuel
fabrication from UFg conversion to the production of completed fuel assemblies
are estimated at 1% of the total throughput uranium. About two thirds of the
uranium losses are solid effluents that are buried within the plant site, and
about one third is liquid and is discharged to a waste lagoon. The atmos-
pheric releases of uranium are estimated to be about 0.003% of the total through-
put uranium. Based on this percentage, the atmospheric releases for 1000 MWe-
year of nuclear fuel are calculated to be about 400 yCi. The annual atmos-
pheric releases for a plant with an annual throughput of 1 x 106 kg of uranium
are calculated to about 0.4 nCi/s. All uranium effluent activity will be
accompanied by the activity of the uranium daughter products thorium-231 and
thorium-234.
With respect to the fabrication of plutonium and uranium-233 nuclear
fuels, there has been only limited experience with plutonium and essentially
no experience with uranium-233.
6. Reactor Operations
Up to the point of reactor operations, the radioactive materials
of the nuclear fuel cycle are all of natural origin. If secular equilibrium
between uranium and its daughters is assumed for the mined materials, then the
radioactivity brought into the biosphere essentially remains at its original
level, unless removed, even though each process and nature redistribute the
individual radionuclides within the biosphere. Removal occurs through natural
decay or by various control procedures, for example, encapsulation and burial.
Because of the extremely long half-lives of the parent radionuclides—7.18 x
108 years for uranium -235 and 4.51 x 109 years for uranium-238—decay removal
is negligible and the total disintegration rate of all the radionuclides is
essentially constant.
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This situation is changed with the initiation of reactor operations.
Within the reactor, fissile materials are fissioned, fertile materials are
converted to fissile materials, and lighter elements in the coolant and struc-
tural materials are made radioactive. The fissioning of the fissile materials,
uranium-235 and plutonium-239 in light water reactors, uranium-235 and uranium-
233 in high-temperature gas reactors, and plutonium-239 and plutonium-241 in
breeder reactors, produces the reactor output energy. The fission process also
produces fission products whose radioactivity is several orders of magnitude
greater than that of the original fissile fuel. In the operation of nuclear
reactors, some of these created radioactive materials are vented to the atmos-
phere and some are discharged to local waters, but most of the radioactive
materials are contained within the fuel rods. After a period of operations
during which most of the fissile materials are fissioned, the "spent fuel rods"
are removed from the reactor, stored for a cooling period and shipped to a
fuel reprocessing plant. The amount of radioactive pollutants released to the
environment at the reactor site depends on a variety of factors. Such factors
include the type and size of the reactor, reactor design, operations, percent
of fuel failure, and the control measures used. Nuclear power plants are
generally located not too distant from the communities they serve, where there
is adequate cooling water, but population density in the immediate surroundings
is generally low. Construction permits have been issued, however, for nuclear
power plants at locations where the population density is high [Ref. 53].
The size of nuclear power plants coming on line, those under construc-
tion, and those planned for future installation are generally in the 800 MWe to
1200 MWe range. A 1000 MWe nuclear power plant can therefore be considered
representative, and it will be used as the reference capacity for specifying
the estimated radioactive effluents from nuclear reactors.
a. Light Water Reactors
Current nuclear power production is entirely from light water
reactors and it is projected that light water reactors will continue to produce
most of the nuclear power through the year 2000. The radioactive materials
created in light water reactors are those created in the coolant water, those
created in the structural materials, and those created within the fuel. The
radioactive materials created in the coolant are generally gases, such as
argon-41, fluorine-18, nitrogen-13, nitrogen-16, and oxygen-19. These radio-
active gases all have relatively short half-lives and most of these radioactive
materials are decayed in place. The portion that is released to the atmosphere
with the discharge of gases from the coolant is quickly decayed to stable nu-
clides. Tritium is also created by neutron activation of boron-10, lithium-7,
and deuterium, but the amount created is small when compared with that produced
by fission. Other radioactive materials such as carbon-14 and sodium-24 are
also created.
The radioactive materials created in the structural materials
normally remain in the structural materials; small amounts do enter the coolant
water. These radioactive materials include chromium-51, manganese-54, man-
ganese-56, cobalt-58, cobalt-60, iron-59, nickel-63, copper-64, zinc-65,
zirconium-95 and many others. Most of these relatively longlived radionuclides
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are normally removed from the coolant by the coolant purification system.
The radioactive materials created in the fuel are fission pro-
ducts and actinides. Most of these radioactive materials will remain in the
uranium dioxide fuel pellets; however, the volatile radionuclides will diffuse
from the fuel. These radioactive materials (and their daughter products)
enter the coolant through defects in the fuel cladding. The nonvolatile
radioactive materials are removed by the coolant purification system. The
separation and storage of the gaseous and volatile materials, on the other
hand, is difficult, and after a period of holdup to reduce radioactivity through
the decay of the shorter half-life radionuclides, these gaseous and volatile
materials are released to the atmosphere. The major radionuclides released
to the environment are the gases: tritium, radiokrypton, and radioxenon;
radioiodine, which is highly volatile at reactor temperatures; and carbon-14
as carbon dioxide.
Tritium is released to the environment as a liquid and as a
gaseous effluent. The release rate of tritium from nuclear reactor operations
has been variously estimated to be on the order of 100 Ci per 1000 MWe-year for
boiling water reactors to 10,000 Ci per 1000 MWe-year for some pressurized
water reactors [Ref. 9]. Roughly 90% of the material released is estimated
to be released as a liquid effluent and roughly 10% in the form of tritiated
gaseous water vapor is estimated to be released to the atmosphere although the
partitioning of these releases vary and can be controlled. On the other hand,
the estimated annual stack releases of tritium from Yankee Nuclear Power
Station, a 185 MWe PWR, was only 13 Ci [Ref. 54].
The amount of krypton escaping to the coolant and subsequently
released to the environment depends on the rate of krypton escape and the hold-
up time prior to release. The escape rate to the coolant is normally very slow,
permitting substantial decay of the shorter half-life radiokryptons. In PWRs
the gas volume buildup in the coolant is relatively slow, and consequently the
extracted gases can be held up for a relatively long time (on the order of
30 days) prior to release. A 30-day delay would reduce all the short-lived
radiokryptons to insignificance and only krypton-85, with a half-life of 10.76
years, would be released. The gas production rate in the BWR coolant, on the
other hand, requires continuous removal and, because of its large volume, the
delay time between gas removal and release must necessarily be shorter. Earlier
practice had been to discharge these gases after only a short holdup time of
about 30 minutes. The current holdup time for krypton from BWRs is about one
day. A delay of one day would reduce krypton-89 to insignificance and the
other short-lived radiokryptons by several orders of magnitude. The estimated
krypton releases per 1000 MWe are on the order of 1 Ci per day for PWRs to
about 100 Ci per day for BWRs [Ref. 9]. However, the estimated annual stack
releases of krypton-85 from the Yankee Nuclear Power Station were only 3 Ci
[Ref. 54].
The carbon-14, released as a gas from light water reactors, is
that which is produced in the coolant. Without controls to capture the carbon
containing substances in the off-gas, the estimated reactor carbon-14 releases
are on the order of 6 Ci per 1000 MWe-year for PWRs and 16 Ci per 1000 MWe-year
for BWRs [Ref. 55].
83
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As with the case of radiokrypton, the amount of radioxenon
escaping to the coolant and subsequently released to the environment depends
on the rate of xenon escape and the holdup time prior to release. The escape
rate to the coolant is normally very slow, permitting substantial decay of
the shorter half-life radioxenon. The slow escape rate and the holdup times
(15 to 30 days) results in xenon-133 being the only radioxenon released in
significant quantities at the reactor site. The estimated reactor xenon
releases per 1000 MWe are on the order of 20 Ci per day for PWRs and 150 Ci
per day for BWRs [Ref. 9]. The estimated annual stack releases of xenon-133
from the Yankee Nuclear Power Station, however, were 0.1 Ci per year [Ref. 54].
The radioiodine that is released from nuclear reactors with
coolant discharges also occurs as the result of fuel-cladding leaks and the
fission of trace quantities of uranium in the coolant. Only small quantities
of radioiodine, as gaseous and liquid effluent, are normally released at the
reactor site. The estimated and measured releases of iodine-131 vary widely,
but are in the order of several millicuries to several curies per 1000 MWe-year
of power production.
Small quantities of other radionuclides are also released at
the reactor site. These less volatile radionuclides are more amenable to
decontamination and conversion to solid waste. The quantities released depend
on the decontamination treatment used. The radionuclides that may be released
in the liquid effluent in significant quantities besides tritium and iodine-131
are: chromium-51, manganese-54, iron-55, iron-59, cobalt-58, cobalt-60, stron-
tium-89, yttrium-90, zirconium-95, niobium-95, molybdenum-99, ruthenium-103,
ruthenium-106, cesium-134, cesium-137, barium-140, cerium-141, and cerium-144.
b. High-Temperature Gas Reactors
The use of a gas instead of water to cool the reactor core
permits higher working temperatures. Helium is currently the gas used in
HTGRs. At present there are only two HTGRs in the United States, a 40-MWe
prototype and the Fort St. Vrain 330-MWe reactor. The latter plant is graphite
moderated, helium cooled, and fueled with thorium and highly enriched uranium.
The thermal efficiency of the plant is about 39%.
Plant operation is initiated with the fission of uranium-235,
but as operations continue, the thorium-232 is converted to uranium-233 and
an increasing fraction of the energy will result from the fission of uranium-233.
As with the light water reactors, most of the radioactivity generated by the
HTGR are the fission products. The distribution of the fission products from
the fission of uranium-233 is not significantly different from that produced
by the fission of uranium-235 or plutonium-239. Most of the fission products
will remain in the fuel, but some can be expected to leak to the helium coolant.
These are removed by the helium purifier and released to the environment under
controlled conditions. A very small fraction of helium-3 in the helium is
neutron activated to form tritium, and this is also released to the atmosphere.
The major releases, however, are the gaseous fission products.
84
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The radiokrypton releases are estimated at about 3000 Ci per
1000 MWe-year, and the radioxenon releases are estimated at about 100 Ci per
1000 MWe-year. Only small quantities of radioactive liquid wastes are gener-
ated by the HTGR, and these are decontaminated as necessary prior to discharge
to the environment.
c. Liquid-Metal Fast Breeder Reactors
The fast breeder reactor uses liquid sodium for the primary
coolant and plutonium and depleted uranium for fuel. It not only uses the
uranium more efficiently in its own operations than do light water reactors
but it also produces more fissionable materials than it uses. Because all
but a small fraction of natural uranium is uranium-238, the LMFBR maximizes
the use of the potential energy in uranium. Not only is enrichment unneces-
sary but the mining and milling requirements are considerably lower than those
for light water reactors.
The fast breeder reactor program is still in a developmental
stage, however, and only relatively low levels of LMFBR energy are projected
in Scenarios III and V for the year 2000. Estimated radionuclide releases
for the LMFBR also vary widely, but they are on the order of a magnitude or
more lower than those estimated or measured for light water reactors.
7. Spent Fuel Reprocessing
The reprocessing of spent light water reactor fuels permits the
maximum use of the scarce natural resource uranium-235. Because the recovered
uranium is slightly enriched, about 0.8% uranium-235, reprocessing also de-
creases enrichment requirements. A small amount of fissile plutonium is also
recovered. Spent fuel reprocessing is also the fuel recovery process for the
thorium and plutonium fuel cycle, where the relatively plentiful resources
thorium-232 and uranium-238 are converted to fissile uranium-233 and plutonium-
239 in HTGRs and LMFBRs.
Currently, however, there is no spent fuel reprocessing plant in
operation in the United States [Ref. 56]. The Nuclear Fuel Services Plant has
been shut down for process alteration and expansion after 6 years of operation
(1966 to 1972). Its capacity was 1000 kg of uranium per day. A second re-
processing plant, the Midwest Fuel Recovery Plant, was constructed but was
never operational and is currently abandoned. The third and latest reprocess-
ing plant, the Barnwell Nuclear Fuel Plant, was scheduled to be operational in
1976, but it has not been licensed to operate. This plant is designed to
process 5000 kg of uranium per day. A plant this size has the capacity to
service about forty-five 1000-MWe light water reactors.
After the fuel rods with the spent fuel are removed from the reactor,
they are stored before they are shipped to reprocessing. A storage time of
150 days permits significant reduction of the radioactivity of the shorter
half-life radionuclides. Even so, vast quantities of natural and reactor
generated radioactive materials are present in the fuel rods at the time of
85
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reprocessing. The total radioactive effluents from fuel reprocessing is large
when compared with the rest of the nuclear fuel cycle. The radioactive materials
present in spent fuel are fission products, light element activation products,
and the actinides. To recover the uranium, all these radioactive materials
must be removed from the uranium. Plutonium is also recovered by chemical
separation. The recovered uranium is converted to uranium hexafluoride and
shipped to enrichment plants. The gaseous radionuclides and small amounts of
the volatile radionuclides are released to the atmosphere. Other low level
radioactive wastes are discharged to local waters. The remaining actinides,
fission products, and the activated light elements, constituting the bulk of
the radioactivity, are retained, converted to solids, and stored. Represent-
ative quantities of the radioactive materials in the spent fuel per 1000 MWe-
year uranium reprocessing requirement (3.3 x 101* kg uranium) are listed in
Table 28.
The Nuclear Fuel Service Plant, when in operation, released to the
atmosphere all the krypton-85 and carbon-14, about 60% of the tritium, and
minute quantities of radioiodine and other fission products and actinides in
the spent fuel [Ref. 51]. The remainder of the tritium, small amounts of
ruthenium-106, and other radioactive materials were released from the site as
a liquid effluent. The radioactive effluents that will be released from future
spent fuel reprocessing plants will depend upon the control systems used to
limit these releases.
The estimated airborne radionuclide releases based on past experience
per 1000 MWe-year of uranium processed are as follows [Ref. 51,55,56]:
Radionuclide Ci/1000 MWe-yr
Krypton-85 3.4 x 105
Tritium 2.2 x 101*
Carbon-14 10
Iodine-129 1 x 10~3
Iodine-131 6 x 10~2
Other fission products 0.9
Actinides 4 x 10~3
The estimated liquid radionuclide releases to streams per 1000 MWe-
year of uranium processed are as follows:
Radionuclide Ci/1000 MWe-yr
Tritium 4 x 103
Ruthenium-106 4
86
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TABLE 28. REPRESENTATIVE QUANTITIES OF RADIOACTIVE MATERIALS
IN SPENT FUEL PER 1000 MWe~YEAR
URANIUM REPROCESSING REQUIREMENT
Radionuclide
Tritium
Carbon-14
Manganese-54
Iron-55
Iron-59
Cobalt-58
Cobalt-60
Krypton-58
Strontium-89
Strontium-90
Yttrium-91
Zirconium-93
Zirconium-95
Niobium-95
Technetium-99
Ruthenixim-103
Ruthenium-106
Antimony-125
Tellurium-125m
Tellurium-127m
Tellurium-129m
Iodine-129
Iodine-131
Cesium-134
Curies
(1000 MWe yr)
2.6 x 101*
10
9.7 x 105
6.5 x 105
1.6 x 10"
9.7 x 105
6.5 x 10U
3.4 x 105
3.2 x 106
1.9 x 106
6.1 x 106
65
1.3 x 107
2.6 x 107
480
5.8 x 106
2.6 x 107
4.2 x 105
2.1 x 105
8.1 x 105
4.2 x 105
1.3
65
3.2 x 106
Radionuclide
Cesium-135
Cesium-137
Cerium-141
Cerium-144
Promethium-147
Europium-155
Uranium-235
Uranium-236
Uranium-238
Neptunium-237
Plutonium-238
Plutonium-239
Plutonium-240
Plutonium-241
Plutonium-242
Americium-241
Americium-243
Curium-242
Curium-244
Curies
(1000 MWe yr)
39
3.4 x 106
2.6 x 106
2.6 x 107
6.5 x 106
1.3 x 106
0.6
9
10
15
1.3 x 105
1.6 x 101*
2.1 x lO4
4.8 x 106
65
2.4 x 101*
650
1.1 x 106
6.5 x 10"
87
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The two newer and, as yet, inoperative plants are designed, however,
to eliminate liquid-effluent discharges to the environment and all the tritium
would be released to the atmosphere as a tritiated water vapor. The actual
quantities of radioactive gaseous and liquid effluents that will be discharged
to the environment once reprocessing operations are resumed will depend upon
the control systems installed in the plants. Several decontamination systems
are under development to reduce gaseous as well as particulate releases.
Except for that small fraction of the radioactive materials released
to the environment, it is planned that the remainder will be converted to
solids and these solids will eventually be consigned to perpetual storage.
The current plans for this waste, that consist of about 1.3 x 108 Ci of fission
products and 2 x 101* Ci of cladding hulls per 1000 MWe-year of energy produced,
is to place it in interim storage for a few years and then to ship it to a
repository for permanent storage. The problems associated with high level
radioactive waste management are yet to be resolved, however, and a permanent
disposal method has not been adopted.
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V RADIOLOGICAL PROJECTIONS
A. Energy Scenario II Emissions and Effects
1. Coal
a. Mining and Preparation
The primary radionuclide emissions of concern during coal mining
and coal preparation are the radon gases, radon-222 and radon-220. The radon
release rates from these two activities are not known. To facilitate projec-
tions of annual releases and population exposure doses the following is
assumed: (1) the combined mining and preparation operations release 20% of
the radon in the coal, (2) the area source dose conversion factor applies,
and (3) the average population within 80 km of these operations is 2 x 105
people.
The estimated radon release rates for all coal requirements
are as follows:
Ci/yr
Year Radon-222 Radon-220 Total
1985 75 55 130
2000 135 99 234
The projected population exposure doses from the radon-222
releases are 1.7 man-rems per year in 1985 and 3 man-rems per year in 2000.
b. Direct Combustion for Electric Power
The primary radionuclides released to the environment in
direct combustion of coal at electric power plants are uranium-238 and its
daughters, thorium-232 and its daughters, and potassium-40. To facilitate
projections of annual releases and population exposure doses, the following
is assumed: (1) coal combustion releases all the radon in the coal, (2)
5% of the radon-222 generated from the radium-226 accumulated at the ash piles
is released to the atmosphere, and (3) the average population within 80 km
of coal-burning power plants is 4 x 106 people.
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The estimated radioactive material releases are as follows:
Ci/yr
Radionuclide 1985 2000
Stack releases
Radon-222 190 230
Radon-220 160 200
Uranium-238 0.9 1.1
Uranium-238 + daughters 13 15
Thorium-232 0.7 0.8
Thorium-232 + daughters 6 7
Potassium-40 1.4 1.7
Ash pile releases
Radon-222 6.5 x 103 1.9 x 101*
The projected population exposure doses from the above releases
are as follows:
Man-Kerns/Year
Radionuclide 1985 2000
Radon-222 470 1.4 x 103
Uranium-238 40 48
Thorium-232 9 10
Potassium-40 6 x 10"3 7 x 10~3
Polonium-210 0.8 1
Additional exposure doses from uranium daughters and thorium daughters are not
listed. Also, if the uranium concentration is increased by a factor of 5,
then the resulting exposure doses will be correspondingly increased. Even if
the uranium exposure doses were increased by a factor of 5, however, they
would be over an order of magnitude lower than the radon exposure doses.
c. Gasification
The projected coal gasification energy levels for Scenarios
11, IV, and V are the same and are at a relatively low level when compared
to the levels projected for other energy systems. No coal gasification is
projected for Scenario III. Using the data in Table 25, the projected
radioactive material releases are as follows:
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Ci/yr
Radionuclides 1985 2000
Stack releases
Radon-222 15 57
Radon-220 20 75
Uranium-238 1.5 x 10~2 5.5 x 10~2
Uranium-238 + daughters 0.11 0.41
Thorium-238 4 x 10~3 1.5 x 10~2
Thorium-238 + daughters 2.8 x 10~2 0.1
Potassium-40 6 x 10~3 2.2 x 10~2
Ash pile releases
Radon-222 580 3,300
The projected population exposure doses using a local population
of 2 x 105 people within 80 km of all gasification facilities are as follows:
Man-Kerns/Year
Radionuclide 1985 2000
Radon-222 2 12
Uranium-238 3 x 10~2 1 x 10"1
Thorium-232 2 x 10"3 9 x 10~3
Polonium-210 7 x 10'4 3 x 10~3
Radium-226 7 x 10~3 3 x 10~2
Potassium-40 1 x 10~6 4 x 10~6
As in the case of direct coal combustion, additional doses from uranium and
thorium daughters are not listed above. The sum of these exposure doses,
however, will be small relative to the radon-222 exposure doses.
d. Liquefaction
The projected coal liquefaction energy levels for Scenarios
II, IV, and V are the same. The projected energy level for 1985 is very low
but for 2000 the projected level exceeds that of direct coal combustion for
electrical power. No coal liquefaction is projected in Scenario III. Using
an average of the emissions listed in Tables 26 and 27, the projected radio-
active material releases are as follows:
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Ci/yr
Radionuclide 1985 2000
Stack releases
Radon-222 5 x 103 1.1 x 105
Radon-220 2.7 x 103 5.9 x 101*
Uranium-238 5 110
Uranium-238 + daughters 46 1 x 103
Thorium-232 1.4 30
Thorium-232 + daughters 16 160
Potassium-40 — —
Ash pile releases
Radon-222 180 5,000
The projected population exposure doses assuming a local pop-
ulation of 2 x 105 people within 80 km of all liquefaction facilities are as
follows:
Man-Kerns/Year
Radionuclide 1985 2000
Radon-222 4 100
Uranium-238 11 240
Thorium-232 0.9 19
Polonium-210 0.2 5
Radium-226 5.5 120
Potassium-40 5.6 x 10'1* 1.2 x lO"2
The uranium and thorium daughters not listed above will also provide exposure
doses. The sum of these exposure doses could very well exceed the radon-222
exposure doses. As the ash piles accumulate with operations beyond the year
2000, however, radon-222 will be the major contributor to exposure doses.
2. Oil Shale
The primary radionuclide releases to the environment are those re-
leased to the atmosphere during processing. These radionuclides and their
estimated release rates are listed in Table 27. The projected levels of shale
oil utilization in Scenarios II, IV, and V are the same, i.e., 1 quad in 1985
and 8 quads in the year 2000. No shale oil utilization is projected in Scenario
III.
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For the projected utilization rates of Scenarios II, IV, and V, the
calculated radionuclide releases are as follows:
Ci/yr
Radionuclide
Uranium-238
Radon-222
All other uranium-238 daughters
Thorium-232
Radon-220
All other thorium-232 daughters
Potass ium-40
1985
1.9
1.3 x 101*
25
0.5
3.2 x 103
4
120
2000
15
1 x 105
200
4.1
2.6 x 104
33
970
To rank the relative importance of these emissions they need to be
converted to radiation exposure doses. To do this, an exposed population is
required and in this case a population of 80,000 is assumed to reside within
80 km of oil shale processing facilities. The calculated population exposure
doses for the release rates listed above for the assumed population within
80 km of oil shale processing facilities are listed below for selected
radionuclides.
Man-Rems/Year
Radionuclide
Radon-222
Radium-226
Uranium-238
Thorium-232
Polonium-210
Potassium-40
1985
3.8
1.8
1.7
0.1
3.5 x 10-2
9.6 x 10~3
30
14
13
1
0.3
7.8
2000
x 10~2
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3. Geothermal
The primary radionuclide of concern ±s the release of radon gas,
principally radon-222, to the atmosphere. The fate of the other radionuclides
in source fluid is dependent on the waste water disposal plan adopted. The
level of geothermal utilization in Scenario II for 1985 and 2000 is extremely
low when compared to the projections for other energy systems. Consequently,
even though the radon release rates per unit of electric energy produced is
relatively high, the annual projected releases are relatively low. For pro-
jection purposes, the radon release rates are estimated to range between
8,000 and 20,000 Ci per quad (1015 Btu). The projected releases for Scenario
II are, therefore, 5,500 to 14,000 Ci per year in 1985 and 11,000 to 28,000 Ci
per year in 2000.
To convert the radon release rates to population exposure doses, an
exposed population is required. The assignment of an average population in
the region of geothermal sites is difficult since only one site is currently
being developed. For the purpose of this exercise a population of 1 x 106
people is assumed to be within 80 km of geothermal developments. Also because
there are many atmospheric release points within a geothermal development area,
the area source dose conversion factor is used. The projected population
exposure doses are 100 to 250 man-reins per year in 1985, and 200 to 500 man-rems
per year in 2000.
4. Nuclear Systems
a. Mining
The primary radionuclide emission of concern is the release of
radon gas, principally radon-222, to the atmosphere. Based on the underground
mining release rate of 574 Ci per ore requirement of 1.3 x 108 kg per 1000 MWe-
year, the radon release rates of 2 mCi/s and 8 mCi/s are estimated for years
1985 and 2000, respectively. Because the half-life of radon-222 is only 3.8
days, it will decay through its even shorter half-life daughters to lead-210
rather rapidly. That is, all the radon-222 released to the environment would
be decayed to lead-210 which in turn decays with a half-life of 22 years.
Again, based on the underground mining release rate to the atmosphere, the
radioactivity of the lead-210 distributed in the environment from the release
of radon-222 is estimated at 150 Ci in 1985 and about 1100 Ci in the year
2000. It should be noted that the estimated radon releases associated with
open-pit mining were several orders of magnitude lower than that for under-
ground mining.
For ranking purposes, all radioactive emissions are converted
to radiation exposure doses. The conversion requires an assumed exposed
population. For uranium mining, a population of 80,000 people is assumed to
reside within 80 km of all mining sites. Using this average population, the
annual radon emission rates, and the area source conversion factor (Q x 1.8
x 10"14), the population exposure doses of 90 man-rems/year and 360 man-rems/
year are calculated for the years 1985 and 2000, respectively. It should be
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noted that the average individual exposure rate is not equal to the man-reins/
year divided by 80,000, but is equal to the man-r ems/year divided by 80,000 N,
where N is equal to the number of mining sites.
b. Milling
The primary radionuclide emission of concern from the mill sites
is the release of radon gas, principally radon-222, to the atmosphere. The
sources of radon releases at the mill site are from the milling operations and
from the mill tailings. Roughly half of the radon in the ore is released
during mill operations and the remainder is discharged to the mill tailings
pond. It is estimated that about 5% of the radon present in the pile escapes
at the mill tailings and the remainder decays to daughter products within the
pile. Because the shorter half-life radon-222 is continuously generated by
the radium-226 in the tailings and 5% of this radon-222 is assumed to escape to
the atmosphere, the radon-222 escape rate from all the mill tailings is esti-
mated at 2,200 Ci/year in 1985 and 20,200 Ci/year in 2000. The radon escape
rates from all milling operations are estimated at 4,400 Ci/year in 1985 and
16,500 Ci/year in 2000. The total radon release rates to the atmosphere from
all milling sites are therefore estimated at 6,600 Ci/year in 1985 and 36,700
Ci/year in 2000. The released radon-222 decays to lead-210. The lead-210
activity in the environment created by the release of radon-222 is estimated
at 11 Ci by 1985 and 110 Ci by 2000.
Assuming an exposed population of 80,000 people per mill site,
the mill site radon releases convert to 10 man-rem/year in 1985 and 50 man-rem/
year in 2000.
It was also estimated that less than 2 curies of uranium and
uranium daughter radioactivity, besides radon, are released to the atmosphere
per 1.3 x 108 kg of ore processed. This is equivalent to uranium and uranium
daughter atmosphere release totals of about 200 Ci/year in 1985 and about
800 Ci/year in 2000. The total activity of the uranium and uranium daughters
released to the environment is projected to be about 1000 Ci in 1985 and about
9000 Ci in 2000. In this case, because the uranium is released from a low-
point source, the conversion factor in Ref. 8 was used. The uranium releases
convert to 7000 man-rems/year and 28,000 man-r ems/year for the years 1985 and
2000, respectively.
c. Conversion to Uranium Hexafluoride
In the conversion of yellowcake to uranium hexafluoride, it
was estimated that about 0.017 Ci per 2.4 x 105 kg of uranium processed is
released to the atmosphere. This projects to uranium release rates of about
2 Ci/year in 1985 and 7.5 Ci/year in 2000. The total activity of the uranium
released is projected to be about 9.4 Ci in 1985 and 81 Ci in 2000. Because
the half-lives of uranium-235 and uranium-238 are extremely long compared
with their daughter products thorium-231 and thorium-234 respectively, insig-
nificant daughter product radioactivity will be generated by the released
uranium.
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An average of 1.5 x 106 people are assumed to reside within
80 km of conversion facilities. Assuming this average population, the uranium
emission rates convert to 1300 man-rems/year in 1985 and 4300 man-rems/year in
2000.
d. Uranium Enrichment
The estimated radioactive effluents from uranium enrichment
amount to about 0.085 Ci of uranium per 1000-MWe light water reactor enrich-
ment requirement. Of this amount, about 90% is discharged as a liquid to
holding ponds and then subsequently discharged to streams. About 10% or less
of the total radioactive effluents are estimated to be atmospheric releases.
This projects to uranium release rates of 1 Ci/year in 1985 and 3.7 Ci/year
in 2000. The total activity of the uranium released in the environment is
estimated to be about 47 Ci in 1985 and 400 Ci in 2000.
Assuming an exposed population of 1.5 x 106 people per enrich-
ment facility, the projected population exposures are 600 man-rems/year in
1985 and 2400 man-rems/year in 2000.
e. Fuel Fabrication
The uranium losses for fuel fabrication from uranium hexafluoride
to the production of completed fuel assemblies are estimated at 1% of the total
uranium throughput. About two-thirds are buried within the plant site as
solids and about one-third is discharged as a liquid to a waste lagoon. About
0.003% of the total throughput uranium is released to the atmosphere. The
projected atmospheric release rates from fuel fabrication are 0.05 Ci in 1985
and 0.2 Ci in 2000. The total activity of the atmospherically released uranium
to the environment is estimated to be about 0.24 Ci in 1985 and 0.9 Ci in 2000.
Assuming an exposed population of 1.5 x 106 people per fuel
fabrication facility, the projected population exposures are 30 man-rems/year
in 1985 and 130 man-rems/year in 2000.
f. Reactor Operations
The major radioactive materials released from nuclear reactor
sites are the fission products krypton-85 and xenon-133, which are noble gases
and are released to the atmosphere, tritium which is released to local streams
and to the atmosphere, and iodine-131 which is also released to the atmosphere.
Small quantities of many other radionuclides are also released at the reactor
site. The estimated radionuclide releases from reactor operations vary widely.
The krypton-85 atmospheric releases are estimated at about 100 Ci per day for
a 1000-MWe boiling water reactor (BWR) and 1 Ci per day for the pressurized
water reactor (FWR); however, measured release rates equivalent to 15 Ci per
year were reported for a BWR and 3 Ci per year were reported for a PWR. The
xenon-133 atmospheric releases are estimated at about 150 Ci per day for a
1000-MWe BWR and 20 Ci per day for the PWR; however, measured release rates
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equivalent to 4.5 x 101* Ci per year were reported for a BWR and 0.5 Ci per
year were reported for a PWR. The carbon-14 atmospheric releases are estimated
at 16 Ci per year for a 1000-MWe BWR and 6 Ci per year for the PWR. The tritium
atmospheric releases are estimated at 10 Ci per year for a 1000-MWe BWR and
1000 Ci per year for a PWR. The tritium releases as liquid effluent to streams
are estimated at 90 Ci per year for the BWR and 9000 Ci per year for the PWR.
The iodine-131 atmospheric releases are estimated to range from several milli-
curies to several curies per year.
The annual radioactive noble gas releases of a 1000-MWe high-
temperature gas-cooled reactor (HTGR) are estimated at 3000 Ci for krypton-85
and 100 Ci for xenon-133. The annual releases from a 1000-MWe liquid-metal
fast breeder reactor (LMFBR) are estimated to be considerably lower than the
HTGR. Because the estimated rates for these reactor types are speculative
and projected energy levels for them are relatively minimal, all projected
reactor radioactive material releases will be based on BWR and PWR release
rates.
Because the estimated radioactive releases are significantly
different between the BWR and the PWR, the projected releases are based on
estimated LWR installed capacities of 40% BWR and 60% PWR.
The projected krypton-85 release rates are 1.7 x 106 Ci per
year in 1985 and 6.5 x 106 Ci per year in 2000. Because the half-life of
krypton-85 is 10.76 years, a considerable buildup of the krypton-85 released
to the environment will result. The total activity of the released krypton-85
is estimated at 7.1 x 106 Ci in 1985 and 4.7 x 107 Ci in 2000.
The projected xenon-133 release rates are 3.1 x 106 Ci per year
in 1985 and 1.2 x 107 Ci per year in 2000. Because the half-life of xenon-133
is only 5.3 days, there will not be any buildup of xenon-133.
The projected carbon-14 atmospheric release rates are 1,200
Ci per year in 1985 and 4,400 Ci per year in 2000. Because the half-life of
carbon-14 is 5700 years, a buildup of carbon-14 activity will result. The
total activity of the released carbon-14 is estimated at 5,500 Ci in 1985 and
48,000 Ci in 2000.
The projected tritium atmospheric release rates are 7.1 x 101*
Ci per year in 1985 and 2.7 x 105 Ci per year in 2000. The projected tritium
stream release rates are 6.4 x 105 Ci per year in 1985 and 2.4 x 106 Ci per
year in 2000. The projected total tritium release rates to the environment
are 7.1 x 105 Ci per year in 1985 and 2.7 x 106 Ci per year in 2000. The
tritium released to the atmosphere in the form of water vapor and its residence
time in the atmosphere is relatively short. The tritium half-life of 12.26
years is sufficiently long, however, for a considerable buildup of released
tritium activity in the environment. The projected total activity of the
released tritium is estimated at 3 x 106 Ci in 1985 and 2 x 107 Ci in 2000.
The projected iodine-133 atmospheric release rates are estimated
at less than one curie to a few hundred curies per year in 1985, and from
one to one thousand curies per -year in 2000. Because of its relatively short
97
-------
half-life of 8 days, the buildup of iodine-131 with time is Insignificant.
The population density near reactor sites varies widely; however,
for calculative purposes an average of 4 x 106 people are assumed to reside
within 80 km of nuclear power plants and an average of 80,000 people per plant
consume downstream water. For macroregional effects, representative populations
of 1.25 x 108 and 1.5 x 108 are assumed for the eastern part of the United
States in 1985 and 2000, respectively. The projected population exposure
doses from reactor operations in man-rems/year are as follows:
Man-Rems/Year
Radionuclide 1985 2000
Krypton-85 (local air) 14 52
Krypton-85 (macroregional) 1 x 10~8 5 x 10~8
Xenon-133 (local air) 25 96
Carbon-14 (local air) 0.02 0.09
Carbon-14 (macroregional) 0.2 0.9
Tritium (local air) 0.9 3
Tritium (local fallout) 70 280
Tritium (local water) 5,000 18,000
Tritium (macroregional) 440 2,000
Iodine-131 (local air) 5 17
g. Spent-Fuel Reprocessing
Currently there is no spent-fuel reprocessing facility in
operation in the United States. Although the regulations governing future
operations may be more stringent than those in the past, the releases of past
operations are used for the projections that are listed below.
Ci/yr
Radionuclide 1985 2000
Atmospheric releases
Krypton-85 4 x 107 1.5 x 108
Carbon-14 1,200 4,400
Tritium 2.6 x 106 1 x 107
Iodine-129 0.1 0.4
Other fission products 100 400
Actinides 0.5 2
Stream releases
Tritium 4.7 x 10s 1.76 x 106
Ruthenium-106 470 1,760
98
-------
The total activity of the released krypton-85 is estimated at
1.7 x 108 Ci in 1985 and 1.1 x 109 in 2000. The total activity of the released
carbon-14 is estimated at 5,500 Ci in 1985 and 48,000 Ci in 2000. The total
activity of all the released tritium is estimated at 1.3 x 107 Ci in 1985
and 8.7 x 107 Ci in 2000. The ruthenium-106 has a half-life of 1 year and
buildup of ruthenium-106 with time is minimal, that is, the ruthenium-106
activity for any year is approximately equal to the sum of the annual releases.
The population density near current inoperative facilities
varies widely, and it is difficult to project the population densities near
future facilities. For calculation purposes a population of 4 x 106 people
is assumed to reside within 80 km of each facility, and 80,000 people per
facility consume the downstream water.
The projected population exposure doses from reprocessing
facility emissions in man-rems per year are as follows:
Man-Kerns/Year
Radionuclide 1985 2000
Krypton-85 (local air) 320 1200
Krypton-85 (macroregional) 2.5 x 10~7 1.1 x 10~6
Carbon-14 (local air) 0.02 0.09
Carbon-14 (macroregional) 0.2 0.9
Tritium (local air) 50 190
Tritium (local fallout) 2700 10,000
Tritium (local water) 3500 14,000
Tritium (macroregional) 1.6 x 10U 7.5 x 104
Iodine-129 (local air) 0.06 0.3
Iodine-131 (local air) 0.9 3
Ruthenium-106 (local water) 300 1200
Actinides (local air) 2 8
The radioactive waste materials that are not released to the
environment are viewed as a source of potential radioactive environmental
pollution. This waste management problem is yet to be completely resolved.
Within the first few years, the decay of the radioactive waste can be expected
to be relatively rapid. However, with the depletion of the shorter lived
radionuclides, the decay rate will be relatively slow. The fission product
activity of the nuclear waste produced is estimated at 1.5 x 1010 Ci per year
in 1985 and 5.7 x 1010 Ci per year in 2000. The total activity of the accumu-
lated waste fission products is estimated at 3 x 1010 Ci in 1985 and 1.4 x 1011
Ci in 2000. The estimated actinide activity produced is 2.4 x 106 Ci per year
in 1985 and 9 x 106 Ci per year in 2000. The total accumulated actinide
activity is estimated at 1 x 107 Ci by 1985 and 7 x 107 Ci by the year 2000.
99
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B. Energy Scenario III Emissions and Effects
1. Coal
a. Mining and Preparation
In Scenario III, the coal mined is a fraction less in 1985
and about 40% less in the year 2000 than the amount mined in Scenario II. The
amount used for direct combustion to produce electric power, however, is the
same in 1985 and increased by about 20% in the year 2000, compared with Scenario
II. The projected radon release rates for all coal requirements are as follows:
Ci/yr
Year Radon-222 Radon-220 Total
1985 68 50 118
2000 104 77 181
The projected population exposure doses from the radon-222
releases are 1.5 man-rem per year in 1985 and 2.3 man-rem per year in 2000.
b. Direct Combustion for Electric Power
The projected radioactive material releases are as follows:
Ci/yr
Radionuclide 1985 2000
Stack releases
Radon-222 190 280
Radon-220 160 240
Uranium-238 0.9 1.3
Uranium-238 + daughters 13 17
Thorium-232 0.7 1
Thorium-232 + daughters 6 13
Potassium-40 1.4 2.1
Ash pile releases
Radon-222 6,500 2 x
100
-------
The projected population exposure doses from the above releases
are as follows:
Man-Reins/Year
Radionuclide 1985 2000
Radon-222 470 1,400
Uranium-238 40 57
Thorium-232 9 1.2
Potassium-40 6 x 10~3 8.4 x 10~3
Polonium-210 0.8 1.1
c. Gasification
No coal gasification is projected in Scenario III.
d. Liquefaction
No coal liquefaction is projected in Scenario III.
2. Oil Shale
No shale oil is projected in Scenario III.
3. Geothermal
The projected geothermal developments in Scenario III are roughly
twice that of Scenario II in 1985 and roughly 5 times that of Scenario II
in 2000. Nevertheless, the projected geothermal energy levels are very low
with respect to projected total energy levels. The projected radon releases
from geothermal plants are 13,000 to 32,000 Ci per year in 1985 and 53,000
to 130,000 Ci per year in 2000.
The projected population exposure doses from these releases are 230
to 580 man-rems per year in 1985 and 950 to 2,300 man-rems per year in 2000.
4. Nuclear Systems
Of the four energy scenarios, the growth rate of the use of nuclear
energy is fastest in Scenario III. Nevertheless, the growth rate of Scenario
III is only 22% greater than that of Scenario II in 1985 and only 10% greater
than that of Scenario II in the year 2000. The projected radioactive pollu-
tants released to the environment will therefore be similarly increased.
101
-------
a. Mining
The projected radon atmospheric release rates for the uranium
mining requirements are about 2.6 mCi/s in 1985 and about 9 mCi/s in 2000.
The amount of radioactivity from the lead-210, distributed in the environment
from the radon-222 released, is estimated at about 170 Ci to 180 Ci by 1985
and about 1200 Ci by the year 2000.
The projected radon population exposure doses are 120 man-rems
per year in 1985 and 410 man-rems per year in 2000.
b. Milling
The total radon-222 escape rates from milling operations and
from the accumulated tailings piles are estimated at 7,900 Ci per year in
1985 and 40,500 Ci per year in 2000. The lead-210 activity in the environment
created by the release of radon-222 from the mill site is estimated at about
13 Ci by 1985 and 120 Ci by the year 2000. The projected activity of the
uranium released to the atmosphere is about 250 Ci per year in 1985 and about
900 Ci per year in the year 2000. The total activity of the uranium and uranium
daughters released to the environment is estimated at about 1,200 Ci by 1985
and about 10,000 Ci by the year 2000.
The projected population exposure doses are 11 man-rems per
year in 1985 and 70 man-rems per year in 2000 for radon; and 9,000 man-rems
per year in 1985 and 32,000 man-rems per year in 2000 for uranium.
c. Conversion to Uranium Hexafluoride
The projected uranium release rates from the uranium conversion
facilities are about 3 Ci per year in 1985 and about 8.2 Ci per year in the
year 2000. The total activity of the uranium released is projected to be
about 11 Ci by 1985 and about 92 Ci by the year 2000.
The project population exposure doses are 2,000 man-rems per
year in 1985 and 5,400 man-rems per year in 2000.
d. Uranium Enrichment
The projected atmospheric releases of uranium from the uranium
enrichment facilities are about 1 Ci per year in 1985 and about 4 Ci per year
in the year 2000. The projected uranium releases to streams are about 11 Ci
per year in 1985 and about 37 Ci per year in 2000. The total activity of the
uranium released to the environment is estimated at 53 Ci by 1985 and 460
Ci by the year 2000.
102
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The projected population exposure doses are 660 man-rems per year
in 1985 and 2,600 man-rems per year in 2000.
e. Fuel Fabrication
The projected atmospheric release rates of uranium from fuel
fabrication facilities are about 0.06 Ci per year in 1985 and 0.2 Ci per
year in the year 2000. The total activity of the atmospherically released
uranium is estimated at 0.25 Ci by 1985 and 2.1 Ci by the year 2000.
The projected population exposure doses are 40 man-rems per year
in 1985 and 130 man-rems per year in 2000.
f. Reactor Operations
The projected krypton-85 release rates for reactor operations
are estimated at 2 x 106 Ci per year in 1985 and 7 x 106 Ci per year in the
year 2000. The total activity of the released krypton-85 is estimated at
8 x 106 Ci by 1985 and 5.3 x 107 Ci by the year 2000.
The projected xenon-133 release rates are 3.8 x 106 Ci per
year in 1985 and 1.3 x 107 Ci per year in the year 2000.
The projected carbon-14 release rates are 1,400 Ci per year in
1985 and 4,800 Ci per year in 2000. The total activity of the released
carbon-14 is estimated at 6,300 Ci in 1985 and 54,000 Ci in 2000.
The projected tritium atmospheric release rates are 8.7 x 101*
Ci per year in 1985 and 3 x 105 Ci per year in the year 2000. The projected
tritium stream release rates are 7.8 x 105 Ci per year in 1985 and 2.6 x 106
Ci per year in 2000. The projected total tritium release rates to the environ-
ment are 8.7 x 105 Ci per year in 1985 and 3 x 106 Ci per year in 2000. The
projected total activity of the tritium releases is estimated at 3.4 x 10° Ci
by 1985 and 2.3 x 107 Ci by the year 2000.
Because the range of the estimated iodine-131 release rates was
extremely broad, the projected values are the same as those projected for
Scenario II, that is, one curie to a few hundred curies per year in 1985, and
one curie to one thousand curies per year in the year 2000.
The projected population exposure doses in man-rems per year
are as follows:
103
-------
Man-Rems/Year
Radionuclide
Krypton-85 (local air)
Krypton-85 (macroregional)
Xenon-133 (local air)
Carbon-14 (local air)
Carbon-14 (macroregional)
Tritium (local air)
Tritium (local fallout)
Tritium (local water)
Tritium (macroregional)
Iodine-131 (local air)
1985
16
1.3 x 10~8
30
0.03
0.2
1
90
6,000
540
5
2000
56
5.3 x 10"8
100
0.1
0.9
4
310
20,000
2,300
18
g. Spent-Fuel Reprocessing
The projected radionuclide releases of fuel reprocessing
operations are as follows:
Ci/yr
Radionuclide 1985 2000
Atmospheric releases
Krypton-85 4.9 x 107 1.6 x 108
Carbon-14 1,400 4,800
Tritium 3.2 x 106 1.1 x 107
Iodine-129 0.14 0.5
Iodine-131 9 29
Other fission
products 130 430
Actinides 0.6 2
Stream releases
Tritium 5.8 x 105 1.9 x 106
Ruthenium-106 580 1,930
The total activity of the released krypton-85 is estimated at
2 x 108 Ci by 1985 and 1.2 x 109 Ci by the year 2000. The total activity of
the released carbon-14 is estimated at 6,300 Ci in 1985 and 54,000 Ci in 2000.
The total activity of all the released tritium is estimated at 1.5 x 107 Ci
by 1985 and 9.8 x 107 Ci by the year 2000.
104
-------
The projected population exposure doses in man-reins per year
are as follows:
Man~Rems/Year
Radionuclide
Krypton-85 (local air)
Krypton-85 (macroregional)
Carbon-14 (local air)
Carbon-14 (macroregional)
Tritium (local air)
Tritium (local fallout)
Tritium (local water)
Tritium (macroregional)
Iodine-129 (local air)
Iodine-131 (local air)
Ruthenium-106 (local water)
Actinides (local air)
1985
390
3 x 10-7
0.03
0.2
60
3,300
4,400
2 x 104
0.09
1
370
2.4
2000
1,300
1.2 x 10~6
0.1
0.9
210
11,000
14,000
8.3 x 10"
0.3
4
1,200
8
The projected production rates of the radioactive fission
products that are not released to the environment are 1.9 x 1010 Ci per year
in 1985 and 6.3 x 1010 Ci per year in the year 2000. The total activity of
accumulated waste fission products is estimated at 3.6 x 1010 Ci by 1985 and
1.6 x 1011 Ci by the year 2000. The estimated actinide activity produced is
2.9 x 106 Ci per year in 1985 and 9.7 x 106 Ci per year in the year 2000. The
total accumulated actinide activity is estimated at 1.1 x 107 Ci by 1985 and
8 x 107 Ci by the year 2000.
C. Energy Scenario IV Emissions and Effects
1. Coal
a. Mining and Preparation
In Scenario IV, the coal production in 1985 and 2000 is slightly
lower than that of Scenari II. The amount of coal used for direct combustion
electric power generation is also slightly lower. The projected radon release
rates for all coal requirements are as follows:
Ci/yr
Radon-222
64
121
Radon-220
47
89
Total
_»^B_^^BB_
111
210
105
-------
The projected population exposure doses from the radon-222
releases are 1.4 man-reins per year in 1985 and 2.7 man-rems per year in 2000.
b. Direct Combustion for Electric Power
The projected radioactive material releases are as follows:
Ci/yr
Radionuclide 1985 2000
Stack releases
Radon-222 174 209
Radon-220 147 176
Uranium-238 0.8 1
Uranium-238 + daughters 11 13
Thorium-232 0.6 0.7
Thorium-232 + daughters 5.5 6.6
Potassium-40 1.3 1.6
Ash pile releases
Radon-222 6,000 1.7 x 10**
The projected population exposure doses from the above releases
are as follows:
Man-Reins /Year
Radionuclide
Radon-222
Uranium-238
Thorium-232
Potassium-40
Polonium-210
1985
430
35
7
5 x 10~3
0.7
2000
1,200
44
9
6 x 10~3
0.9
c. Gasification
The proposed coal gasification energy levels are the same as
Scenario II. The projected radioactive material releases are as follows:
106
-------
Radionuclide
Cl/yr
Stack releases
Radon-222
Radon-220
Uranium-238
Uranium-238 + daughters
Thorium-232
Thorium-232 + daughters
Potassium-40
Ash pile releases
Radon-222
1985
15
20
1.5 x 10~2
0.11
4 x 10~3
2.8 x 10~2
6 x 10~3
580
2000
57
75
5.5 x 10~2
0.41
1.5 x 10~2
0.1
2.2 x 10"2
3,300
The proj ected population exposure doses, assuming a local pop-
ulation of 2 x 105 people within 80 km of all gasification facilities, are
as follows:
Man-Rems/Year
Radionuclide
Radon-222
Uranium-238
Thorium-232
Polonium-210
Radium-226
Potassium-40
1985
2
3 x 10~2
2 x 10~3
7 x 10-1*
7 x ID"3
1 x 10~6
2000
12
1 x 10"1
9 x 10~3
3 x 10~3
3 x 10~2
4 x 10~6
Scenario II.
Liquefaction
The projected coal liquefaction energy levels are the same as
The projected radioactive material releases are as follows:
Radionuclide
Stack releases
Radon-222
Radon-220
Uranium-238
Uranium-238 '+ daughters
Ci/yr
1985
5 x 103
2.7 x 103
5
46
2000
1.1 x 10s
5.9 x
110
1 x 103
107
-------
Ci/yr
Radionuclide 1985 2000
Stack releases (continued)
Thorium-232 1.4 30
Thorium-232 + daughters 16 160
Potassium-40
Ash pile releases
Radon-222 180 5,000
The projected population exposure doses, assuming a local
population of 2 x 105 people within 80 km of all liquefaction facilities,
are as follows:
Man-Rems/Year
Radionuclide
Radon-222
Dranium-238
Thorium-232
Polonium-210
Radium-226
Potassium-40
1985
4
11
0.9
0.2
5.5
5.6 x 10~u
2000
100
240
19
5
120
1.2 x 10~2
2. Oil Shale
The projected shale oil energy levels are the same as Scenario II.
The projected radioactive material releases to the atmosphere are as follows:
Ci/yr
Radionuclide
Uranium-238
Radon-222
All other uranium-238 daughters
Thorium-232
Radon-220
All other thorium-232 daughters
Potassium-40
1985
1.9
1.3 x 10"
25
0.5
3.2 x 103
4
120
2000
15
1 x 105
200
4.1
2.6 x 10"
33
970
108
-------
The projected population exposure doses, assuming a local
population of 80,000 people within 80 km of all oil shale processing facilities,
are as follows:
Man-Reins /Year
Radipnuclide 1985 2000
Radon-222 3.8 30
Radium-226 1.8 14
Uranium-238 1.7 13
Thorium-232 0.1 1
Polonium-210 3.5 x 10~2 0.3
Potassium-40 9.6 x 10~3 7.8 x 10~2
3. Geothermal
The geothermal development projected in, Scenario IV is the highest
of the four scenarios. The projected geothermal energy level of 15 quads
(1015 Btu) per year for the year 2000 is roughly one-tenth of the total
projected energy level for that year. The projected radon releases from
geothermal plants are 2.6 x 101* to 6.4 x 101* Ci per year in 1985 and 1.2 x
105 to 3 x 105 Ci per year in 2000.
The projected population exposure doses are 470 to 1,200 man-rems
per year in 1985 and 2.2 x 103 to 5.4 x 103 man-rems per year in 2000.
4. Nuclear Systems
The growth rate of the use of nuclear energy in Scenario IV is the
same as that of Scenario II up to the year 1985. It then levels off at 11.37
quads per year (1.2 x 1019 J). For light water reactors this is approximately
equal to 124 GWe-year per year. The projected radioactive material emission
rates for the entire nuclear fuel cycle for Scenario IV through 1985 is the
same as that projected for Scenario II. After 1985, the projected emission
rates are based on a constant nuclear power utilization rate of 124 GWe.
a. Mining
The projected radon atmospheric release rates for the uranium
mining requirements are about 2 mCi/s in 1985 and 2.1 mCi/s in the year 2000.
The amount of radioactivity from the lead-210, distributed in the environment
from the radon-222 released, is estimated at about 150 Ci by 1985 and about
560 Ci by the year 2000.
The projected radon population exposure doses are 550 man-rems
per year in 1985 and 580 man-rems per year in 2000.
109
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b. Milling
The total radon-222 escape rates from milling operations and
from the accumulated tailings piles are estimated at 6,600 Ci per year in
1985 and 15,000 Ci per year in the year 2000. The lead-210 activity in the
environment created by the release of radon-222 from the mill site is estimated
at about 11 Ci by 1985 and 60 Ci by the year 2000. The projected activity
of the uranium released to the atmosphere is about 200 Ci per year in 1985
and about the same in the year 2000. The total activity of the uranium and
uranium daughters released to the environment is estimated at about 1000 Ci
by 1985 and about 4800 Ci by the year 2000.
The projected population exposure doses are 10 man-rems per
year in 1985 and 22 man-rems per year in 2000 for radon; and 7000 man-rems
per year in 1985 and about the same in the year 2000 for uranium.
c. Conversion to Uranium Hexafluoride
The projected release rates from the uranium conversion
facilities are about 2 Ci per year in 1985 and about 2.1 Ci per year in the
year 2000. The total activity of the uranium released is projected to be
about 8 Ci by 1985 and about 40 Ci by the year 2000.
The projected population exposure doses are 1,300 man-rems per
year in 1985 and 1,400 man-rems per year in 2000.
d. Uranium Enrichment
The projected atmospheric releases of uranium from the uranium
enrichment facilities are about 1 Ci per year in 1985 and about the same in
the year 2000. The projected uranium releases to streams are about 9 Ci per
year in 1985 and about 9.5 Ci per year in 2000. The total activity of the
uranium released to the environment is estimated at 47 Ci by 1985 and about
200 Ci by the year 2000.
The projected population exposure doses are 600 man-rems per
year in 1985 and about the same in 2000.
e. Fuel Fabrication
The projected atmospheric release rates of the uranium from
fuel fabrication facilities are about 0.05 Ci per year in 1985 and about the
same in the year 2000. The total activity of the atmospherically released
uranium is estimated at 0.24 Ci by 1985 and about 1 Ci by the year 2000.
The projected population exposure doses are 30 man-rems per year
in 1985 and about the same in 2000.
110
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f. Reactor Operations
The projected krypton-85 release rates for reactor operations are
estimated at 1.7 x 106 Ci per year in 1985 and 1.8 x 106 Ci per year in the
year 2000. The total activity of the released krypton-85 is estimated at
7.1 x 106 Ci by 1985 and 1.9 x 107 Ci by the year 2000.
The projected xenon-133 release rates are 3.1 x 106 Ci per year
in 1985 and 3.3 x 106 Ci per year in the year 2000.
The projected carbon-14 release rates are 1,200 Ci per year in
1985 and about the same in 2000. The total activity of the released carbon-14
is estimated at 5,500 Ci in 1985 and 24,000 Ci in 2000.
The projected tritium atmospheric release rates are 7.1 x 101*
Ci per year in 1985 and 7.5 x 104 Ci per year in the year 2000. The projected
tritium stream release rates are 6.4 x 10^ Ci per year in 1985 and 6.7 x 105
Ci per year in the year 2000. The projected total tritium release rates to
the environment are 7.1 x 105 Ci per year in 1985 and 7.5 x 10s Ci per year
in the year 2000. The projected total activity of the tritium releases is
estimated at 3 x 106 Ci by 1985 and 8.3 x 106 Ci by the year 2000.
The estimated iodine-131 release rates are one curie to a few
hundred curies per year for the years from 1985 through 2000.
The projected population exposure doses in man-rems per year
are as follows:
Man-Reins/Year
Radionuclide
Krypton-85 (local air)
Krypton-85 (macroregional)
Xenon-133 (local air)
Carbon-14 (local air)
Carbon-14 (macroregional)
Tritium (local air)
Tritium (local fallout)
Tritium (local water)
Tritium (macroregional)
Iodine-131 (local air)
1985
14
1 x 10~8
25
0.02
0.2
0.9
70
5,000
440
5
2000
14
1.4 x 10~8
26
0.02
0.2
0.9
80
5,000
560
5
g. Spent-Fuel Reprocessing
The projected"radionuclide releases of fuel reprocessing
operations are as follows:
111
-------
Ci/yr
Radionuclide
Atmospheric releases
Krypton-85
Carbon-14
Trititon
Iodine-129
Iodine-131
Other fission products
Actinides
1985
4 x 107
1,200
2.6 x 106
0.1
7
100
0.5
2000
4.2 x
1,200
2.7 x
0.1
7.4
110
0.5
107
106
Stream releases
Tritium 4.7 x 105 5 x 10s
Ruthenium-106 470 500
The total activity of the released krypton-85 is estimated at
1.7 x 108 Ci by 1985 and 4.1 x 108 Ci by the year 2000. The total activity
of the released carbon-14 is estimated at 5,500 Ci in 1985 and 24,000 Ci in
2000. The total activity of all the released tritium is estimated at 1.3 x 107
Ci by 1985 and 2.9 x 107 Ci by the year 2000.
The projected population exposure doses in man-reins per year
are as follows:
Man-Kerns/Year
Radionuclide
Krypton-85 (local air)
Krypton-85 (macroregional)
Carbon-14 (local air)
Carbon-14 (macroregional)
Tritium (local air)
Tritium (local fallout)
Tritium (local water)
Tritium (macroregional)
Iodine-129 (local air)
Iodine-131 (local air)
Ruthenium-106 (local water)
Actinides (local air)
1985
320
2.5 x 10~7
0.02
0.2
50
2,700
3,500
1.6 x 10**
0.06
0.9
300
2
2000
340
3 x 10~7
0.02
0.2
50
2,800
3,800
2 x 10**
0.06
0.9
320
2
112
-------
The projected production rates of the radioactive fission
products that are not released to the environment are 1.5 x 1010 Ci per year
in 1985 and 1.6 x 1010 Ci per year in the year 2000. The total activity of
the accumulated waste fission products is estimated at 3 x 1010 Ci by 1985 and
4.5 x 1010 Ci by the year 2000. The estimated actinide activity produced
is 2.4 x 106 Ci per year in 1985 and 2.5 x 106 Ci per year in the year 2000.
The total accumulated actinide activity is estimated at 1 x 107 Ci by 1985
and 2.7 x 107 Ci by the year 2000.
D. Energy Scenario V Emissions and Effects
1. Coal
a. Mining and Preparation
The projected coal utilization is lower for Scenario V than for
any other energy scenario. The projected radon releases for all coal require-
ments are as follows:
Ci/yr
Year Radon-222 Radon-220 Total
1985 57 42 99
2000 97 71 168
The projected population exposure doses from the radon-222
releases are 1.3 man-rems per year in 1985 and 2.1 man-rems per year in 2000.
b. Direct Combustion for Electric Power
The projected radioactive material releases are as follows:
Ci/yr
Radionuclide 1985 2000
Stack releases
Radon-222 145 124
Radon-220 122 105
Uranium-2 38 0.7 0.6
Uranium-238 + daughters 8.8 7.8
Thorium-232 0.5 0.5
Thorium-232 + daughters 4.6 4.1
Potassium-40 1.1 1
Ash pile releases
Radon-222 5,500 1.4 x 10"
113
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The projected population exposure doses from the above releases
are as follows:
Man-Rems/Year
Radionuclide
Radon-222
Uranium-238
Thorium-232
Potass ium-40
Polonium-210
1985
400
31
6
4.4 x 10~3
0.6
2000
1,000
26
6
4 x 10~3
0.6
c. Gasification
The projected coal gasification energy levels are the same as
Scenario II. The projected radioactive material releases are as follows:
Ci/yr
Radionuclide 1985 2000
Stack releases
Radon-222 15 57
Radon-220 20 75
Uranium-238 1.5 x 10"2 5.5 x 10~2
Uranium-238 + daughters 0.11 0.41
Thorium-232 4 x 10~3 1.5 x 10~2
Thorium-232 + daughters 2.8 x 10~2 0.1
Potassium-40 6 x 10~3 2.2 x 10~2
Ash pile releases
Radon-222 580 3,300
The projected population exposure doses, assuming a local
population of 2 x 10s people within 80 km of all gasification facilities,
are as follows:
Man-Rems /Year
Radionuclide
Radon-222
Uranium-238
Thorium-232
Polonium-210
Radon-226
Potassium-40
1985
2
3 x 10~2
2 x 10"3
7 x 10"1*
7 x 10~3
1 x 10~6
114
2000
12
1 x 10"1
9 x 10"3
3 x 10~3
3 x 10~2
4 x 10~6
-------
d. Liquefaction
The projected coal liquefaction energy levels are the same as
Scenario II. The projected radioactive material releases are as follows:
Ci/yr
Radionuclide
Stack releases
Radon-222
Radon-220
Uranium-238
Uranium-238 + daughters
Thorium-232
Thorium-232 + daughters
Potassium-40
Ash pile releases
Radon-222
1985
5 x 10 3
2.7 x 103
5
46
1.4
16
180
2000
1.1 x 10
5.9 x 10
110
1 x 103
30
160
5,000
5
J*
The projected population exposure doses, assuming a local pop-
ulation of 2 x 105 people within 80 km of all liquefaction facilities,
are as following:
Man-Rems/Year
Radionuclide
Radon-222
Uranium-238
Thorium-232
Polonium-210
Radium-226
Potassium-40
1985
4
11
0.9
0.2
5.5
5.6 x 10"1*
2000
100
240
19
5
120
1.2 x 10"2
115
-------
2. Oil Shale
The projected shale oil energy levels are the same as Scenario II.
The projected radioactive material releases to the atmosphere are as follows:
Ci/yr
Radionuclide
Uranium-238
Radon-222
All other uranium-238 daughters
Thorium-232
Radon-220
All other thorium-232 daughters
Potassium-40
1985
1.9
1.3 x 101*
25
0.5
3.2 x 103
4
120
2000
15
1 x
200
4.1
2.6
33
970
105
x 10"
The projected population exposure doses, assuming a local popu-
lation of 80,000 people within 80 km of all oil shale processing facilities,
are as follows:
Man-Reins/Year
Radionuclide 1985 20000
Radon-222 3.8 30
Radium-226 1.8 14
Uranium-238 1.7 13
Thorium-232 0.1 1
Polonium-210 3.5 x 10~2 0.3
Potassium-40 9.6 x 10"3 7.8 x 10~2
3. Geothermal
The projected geothermal developments for this scenario are the same
as those for Scenario III. The projected radon releases are 13,000 to 32,000
Ci per year in 1985 and 53,000 to 130,000 Ci per year in 2000. The projected
population exposure doses from these releases are 230 to 580 man-rems per year
in 1985 and 950 to 2,300 man-rems per year in 2000.
116
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4. Nuclear Systems
The growth rate of the use of nuclear energy for Scenario V is
the same as that for Scenario III, up to the year 1985. After 1985, the number
of additional power plant units per year starts to decrease, leveling out at
24.3 quads per year (2.56 x 1019 J) in the year 2000. The nuclear energy
utilization rate in 2000 is equivalent to 264 GWe.
a. Mining
The projected radon atmospheric release rates for the uranium
mining requirements are about 2.6 mCi/s in 1985 and about 4.8 mCi/s in 2000.
The amount of radioactivity from the lead-210, distributed in the environment
from the radon-222 released, is estimated at about 170 Ci to 180 Ci by 1985 and
about 840 Ci by the year 2000.
The projected radon population exposure doses are 120 man-rems
per year in 1985 and 220 man-rems per year in 2000.
b. Milling
The total radon-222 escape rates from milling operations and
from the accumulated tailings piles are estimated at 7,900 Ci per year in 1985
and 27,000 Ci per year in the year 2000. The lead-210 activity in the environ-
ment created by the release of radon-222 is estimated at 13 Ci by 1985 and 95
Ci by the year 2000. The projected activity of the uranium released to the
atmosphere is about 250 Ci per year in 1985 and about 500 Ci per year in the
year 2000. The total activity of the uranium and uranium daughters released
to the environment is estimated at about 1,200 Ci by 1985 and about 8,000 Ci
by the year 2000.
The projected population exposure doses are 11 man-rems per
year in 1985 and 40 man-rems per year in 2000 for radon; and 9,000 man-rems
per year in 1985 and 18,000 man-rems per year in 2000 for uranium.
c. Conversion to Uranium Hexafluoride
The projected release rates from the uranium conversion facil-
ities are about 3 Ci per year in 1985 and about 5 Ci per year in the year
2000. The total activity of the uranium released is projected to about 11 Ci
by 1985 and about 68 Ci by the year 2000.
The projected population exposure doses are 2,000 man-rems per
year in 1985 and 3,300 man-rems per year in 2000.
117
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d. Uranium Enrichment
The projected atmospheric releases of uranium from the uranium
enrichment facilities are about 1 Ci per year in 1985 and about 2 Ci per year
in the year 2000. The projected releases to streams are about 11 Ci per year
in 1985 and about 20 Ci per year in 2000. The total activity of the uranium
released to the environment is estimated at 53 Ci by 1985 and 340 Ci by the
year 2000.
The projected population exposure doses are 660 man-rems per
year in 1985 and 1,300 man-rems per year in 2000.
e. Fuel Fabrication
The projected atmospheric release rates of the uranium from
fuel fabrication facilities are about 0.06 Ci per year in 1985 and 0.1 Ci
per year in 2000. The total activity of the atmospherically released uranium
is estimated at 0.25 by 1985 and 1.6 Ci by the year 2000.
The projected population exposure doses are 40 man-rems per
year in 1985 and 70 man-rems per year in 2000.
f. Reactor Operations
The projected krypton-85 release rates for reactor operations
are estimated at 2 x 106 Ci per year in 1985 and 3.8 x 106 Ci per year in
the year 2000. The total activity of the released krypton-85 is estimated at
8 x 106 Ci by 1985 and 3.5 x 107 Ci by the year 2000.
The projected xenon-133 release rates are 3.8 x 106 Ci per year
in 1985 and 7 x 106 Ci per year in the year 2000.
The projected carbon-14 release rates are 1,400 Ci per year in
1985 and 2,600 Ci per year in 2000. The total activity of the released
carbon-14 is estimated at 6,300 Ci in 1985 and 40,000 Ci in 2000.
The projected tritium atmospheric release rates are 8.7 x 101*
Ci per year in 1985 and 1.6 x 105 Ci per year in 2000. The projected tritium
stream release rates are 7.8 x 105 Ci per year in 1985 and 1.4 x 106 Ci per
year in 2000. The projected tritium release rates to the environment are
8.7 x 105 Ci per year in 1985 and 1.6 x 106 Ci per year in 2000. The projected
total activity of the tritium releases is estimated at 3.4 x 106 Ci by 1985
and 1.5 x 107 Ci by the year 2000.
The estimated iodine-131 release rates are one curie to a few
hundred curies per year in 1985, and one curie to several hundred curies
per year in the year 2000.
118
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The projected population exposure doses in man-reins per year
are as follows:
Man-Kerns/Year
Radionuclide
Krypton-85 (local air)
Krypton-85 (macroregional)
Xenon-133 (local air)
Carbon-14 (local air)
Carbon-14 (macroregional)
Tritium (local air)
Tritium (local fallout)
Tritium (local water)
Tritium (macroregional)
Iodine-131 (local air)
g. Spent-Fuel Reprocessing
The projected radionuclide
operations are as follows:
Radionuclide
Atmospheric releases
Krypton-85
Carbon-14
Tritium
Iodine-129
Iodine-131
Other fission products
Actinides
Stream releases
Tritium
Ruthenium-106
1985
16
2000
30
1.3 x 10~8 2.9 x 10~8
30
0.03
0.2
1
90
6,000
540
5
releases
1985
4.9 x
1,400
3.2 x
0.14
9
130
0.6
5.8 x
580
60
0.05
0.5
2
170
11,000
1,200
10
of fuel reprocessing
Ci/yr
2000
107 9 x 107
2,600
106 5.8 x 106
0.26
16
240
1.1
105 1.1 x 106
1,100
The total activity of the released krypton-85 is estimated to
be 2 x 108 Ci by 1985 and 8.2 x 108 Ci by the year 2000. The total activity
of the released carbon-14 is estimated at 6,300 Ci in 1985 and 40,000 Ci in
2000. The total activity of all the released tritium is estimated at 1.5 x
107 Ci by 1985 and 6.6 x 107 Ci by the year 2000.
119
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The projected population exposure doses in man-reins per year
are as follows:
Man-Kerns/Year
Radionuclide 1985 2000
Krypton-85 (local air)
Krypton-85 (macroregional)
Carbon- 14 (local air)
Carbon-14 (macroregional)
Tritium (local air)
Tritium (local fallout)
Tritium (local water)
Tritium (macroregional)
Iodine-129 (local air)
Iodine-131 (local air)
Ruthenium-106 (local water)
Actinides (local air)
390
3 x 10~7
0.03
0.2
60
3,000
4,400
2 x 101*
0.09
1
370
2.4
720
7 x 10~7
0.05
0.5
110
6,000
7,500
4.4 x 101*
0.2
2
700
4
The projected production rates of the radioactive fission
products that are not released to the environment are 1.9 x 1010 Ci per year
in 1985 and 3.4 x 1010 Ci per year in the year 2000. The total activity of
the accumulated waste fission products is estimated at 3.6 x 1010 Ci by 1985
and 9.8 x 1010 Ci by the year 2000. The estimated actinide activity produced
is 2.9 x 106 Ci per year in 1985 and 5.3 x 106 Ci per year in the year 2000.
The total accumulated actinide activity is estimated at 1.1 x 107 Ci by 1985
and 5.2 x 107 Ci by the year 2000.
120
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VI DISCUSSION OF RESULTS
The radioactive materials release rates and the population exposure doses
for coal, oil shale, geothermal, and nuclear resources utilization in four pro-
jected expanded energy programs were calculated for the years 1985 and 2000
and presented in the previous section. These results were based on radio-
active material concentrations that varied widely, release rates that also
varied widely, and exposure dose calculations where many of the equation input
values were speculative rather than established. Where data were not available,
estimates were made to facilitate the calculation of results. For example, it
was assumed that 5% of the radon generated from radium-226 at the coal ash pile
escapes to the atmosphere. This assumed escape rate may be several times too
high or it may be too low. Also, because geothermal, coal gasification and
liquefaction, and shale oil extraction energy systems are not yet established
industries, their future locations with respect to surrounding population
densities are not known. To fill the calculation gap, rough estimates of
potential exposed populations were necessary. The calculated population expos-
ure doses, therefore, are not for the purpose of projecting health effects,
but rather for projecting the relative importance of the radioactive materials
released to the environment.
For nonnuclear energy systems, the dominant radionuclide contributing to
radiation exposure doses is radon-222. The energy system contributing a major
portion of the radon-222 population exposure doses is the combustion of coal
to produce electricity. The major source of radon-222 from this system is the
coal ash piles—the stack releases of radon-222 from coal combustion being
relatively insignificant. The coal ash piles would be the major radon-222
release source for the coal combustion for electricity energy system even if
the radon release were 1% rather than the assumed 5%.
On a per unit of energy basis, the geothermal energy system is also a
major contributor to radon-222 exposure doses. For example, in Scenario IV
where the projected geothermal and the coal combustion for electricity levels
were similar, the population exposure doses attributed to the release of
radon-222 from these energy systems were of the same order of magnitude.
Radon escapes also from the coal ash piles of coal gasification and coal
liquefaction facilities; however, the accumulated ash from both these facilities
is less than that from coal combustion. In addition, it was assumed that the
coal gasification and liquefaction facilities were located in more sparsely
populated areas. For example, if the coal liquefaction expansion rate indicated
in Scenarios II, IV, and V were to continue, and these liquefaction plants
were located in more densely populated areas, the population radon-222
exposure doses from these plants could exceed those from coal-burning power
plants.
121
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The other radioactive materials released from these nonnuclear energy
facilities are the uranium series radionuclides, the thorium series radio-
nuclides, and potassium-40. Because these radionuclides are either in liquid,
solid, or particulate form, the waste or emissions control systems of these
facilities limit their escape to the environment and the population radiation
exposure doses from these radionuclides are relatively insignificant.
The population exposure doses derived from the extraction of shale oil are
relatively insignificant. Radon-222 will escape from the waste shale piles, but
the radium-226 content in oil shale is very low, and hence, the radon escape
rate from the massive amounts of spent shale is expected to be slow.
For nuclear energy systems, the ranking order of the radionuclides contri-
buting to radiation exposures is subject to the control technologies adopted.
Based on the radionuclide release rates given in Section IV, the release of
tritium from reactor operations and from spent-fuel reprocessing provided the
greatest population exposure doses. This was followed by the population ex-
posure doses from the escape of particulate uranium from milling operations,
conversion facilities, enrichment facilities, and fuel fabrication facilities.
Other radionuclides and sources contributing to the population doses in de-
creasing order of importance are: radon-222 from mining and milling operations,
krypton-85 from spent fuel reprocessing and reactor operations, ruthenium-106
from spent fuel reprocessing operations, xenon-133 from reactor operations,
and iodine-131 from reactor and spent-fuel reprocessing operations.
The relative importance of these radioactive emissions can be perceived
by the following summation of the exposure doses derived from these radio-
nuclides and their sources from Scenario IV.
Man-rems/year
Radionuclide and Source 1985 2000
Tritium reprocessing 2.2 x 101* 2.7 x 101*
reactor 5.5 x 103 5.6 x 103
Tritium total 2.7 x 101* 3.3 x 101*
Uranium-238 milling 7 x 103 7 x 103
conversion 1.3 x 103 1.4 x 103
enrichment 660 660
fabrication 30 30
Dranium-238 total 9 x 103 9.1 x 103
Radon-222 mining 550 580
milling 10 22
Radon-222 total 560 600
Krypton-85 reprocessing 320 340
reactor 14 14
Krypton-85 total 330 350
Ruthenium-106 reprocessing 300 320
Xenon-133 reactor 25 26
Iodine-133 and Iodine-129 total 6 6
Actinides reprocessing 2 2
Carbon-14 reactor 0.2 0.2
reprocessing 0.2 0.2
122
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For comparative purposes the population exposure doses for nonnuclear
energy systems from Scenario IV may be used. For direct combustion of coal,
the projected energy levels in 1985 and 2000 are sufficiently similar for
direct comparisons. The 1985 projections for the remaining nonnuclear energy
systems are omitted because of the relatively low energy levels projected.
The coal liquefaction and the geothermal energy levels in 2000 can be consid-
ered comparable. The projected energy level for coal gasification is lower by
a factor of 3 and the projected energy level for shale oil is lower by a factor
of 2. The projected population exposure doses for comparable projected energy
levels for nonnuclear energy systems in Scenario IV are summarized as follows:
Man-rems/year
Energy System 1985 2000
Coal, direct combustion
Radon-222 430 1200
Uranium-238 35 44
Thorium-232 7 9
Polonium-210 0.7 0.9
Coal gasification
Radon-222 — 12
Coal liquefaction
Radon-222 — 100
Uranium-238 — 240
Radium-226 — 120
Polonium-210 — 5
Shale oil
Radon-222 — 30
Uranium-238 — 13
Radium-226 ~ 14
Geothermal
Radon-222 — 2.3 x
As can be seen, the population exposure doses derived from nuclear energy
systems, the direct combustion of coal, and the extraction of geothermal
energy are comparable in magnitude. Although the projected coal liquefaction
energy level for the year 2000 is comparable to that of direct coal combustion,
the projected population exposure dose for coal liquefaction is considerably
lower. The reasons for this lower dose are that the projected late development
of the coal-liquefaction industry makes the accumulated coal ash less, and it
was assumed that the population in the region of coal liquefaction plants
will be lower than that in the region of coal-burning power plants.
123
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The exposure doses received by individuals within 3 km of the accumulated
ash from a 100,000 bbl/day liquefaction plant, however, would be relatively
high. For example, if 5% of the radon escapes from the 30-year accumulation
of ash for the case cited in the text, the annual radon release rate would be
about 800 Ci/year. At this release rate the estimated lung exposure dose
is about one rem per year at 3 km.
In the case of the direct combustion of coal, the total population expos-
ure dose is divided among a large population because a large number of power
plants are required to produce the projected energy levels. For this reason,
the average individual exposure doses are relatively low. For example, if the
number of people exposed were 30 million people, then the average individual
exposure would be 0.04 millirem per year.
On the other hand, the exposure dose to people within 3 km of the ash
pile of a 1000-MWe coal-combustion power plant could be relatively high. For
example, if 5% of the radon escapes from the 30-year accumulation of ash for
the 1000-MWe case cited in the text, the annual release rate would be about
200 Ci per year. At this release rate, the estimated lung exposure dose is
about 0.04 rem per year at 3 km.
For the geothermal energy extraction case, the number and locations of
the projected geothermal developments are not known. However, it can be
anticipated that because the geothermal areas are generally located in the
less populated western states, fewer people than in the case of direct coal
combustion would be exposed. The result is that for the same total population
exposure dose, the average individual exposure dose would be higher. For
example, if only half the number (15 million) of people were exposed, the
average individual exposure dose for geothermal radon exposures would be 1.5
millirem per year. For close-in exposure doses, the potential radon releases
from each geothermal site must be separately assessed.
In the case of nuclear energy, a major part of the population exposure
dose is from the release of tritium during spent-fuel reprocessing. The
capacity of spent-fuel reprocessing plants is relatively large, and only a few
would be necessary to meet the reprocessing requirements of the projected
Scenario IV nuclear energy level for the year 2000. Thus, the average individ-
ual exposure doses among the local population could be relatively high. If it
is assumed that the number of spent-fuel reprocessing facilities in the year
2000 for this scenario is five, then the average individual exposure dose for
the projected 400,000 people drinking the downstream waters would be 9.5 milli-
rems per year. The tritium total exposure dose received by individuals within
3 km of a reprocessing plant would be 1.4 rems per year. The average individual
exposure dose among the local (within 80 km) population of 20 million people
(5 x 4 x 106) from atmospheric tritium releases would be only 0.14 millirems
per year.
The population exposure doses from the release of tritium from reactor
operations are about a factor of 5 lower than that from spent-fuel reprocessing.
Also because the total energy output is spread among a large number of reactors,
the local exposure doses are lower by more than 2 orders of magnitude. Exposure
124
-------
doses from the shorter-lived xenon-133 and iodine-131 radionuclide releases,
on the other hand, are higher, but the population exposure doses from these'
two radionuclides are, nevertheless, relatively insignificant.
The radon releases from uranium mill tailings piles (totalling 1.5 x 101*
Ci/yr for the year 2000) are comparable to the coal combustion ash pile release
(totaling 1.7 x 10U Ci/yr for the year 2000). The higher population exposure
doses derived for coal ash pile releases are due to the greater number of
people that are exposed by the radon releases from coal burning power facilities.
The uranium mill is also the facility releasing the greatest quantity
of uranium particulates to the atmosphere and the source of the major uranium
population exposure dose. Most of the atmospheric uranium releases are from
the drying process. Although the amount of uranium radioactivity released
is substantially lower than that of radon, the dose conversion factor for
uranium is over two orders of magnitude higher; this results in the relatively
high uranium population exposure doses.
On a local basis, the size of a mill tailings pile is independent of the
energy scenarios and a large mill with a long plant life will accumulate a
large tailings pile. The exposure doses received by individuals close-in to
these large tailings piles could be relatively high.
The other radionuclides listed do not contribute significantly to popu-
lation exposure doses. Other natural radionuclides that are not listed here
are also released to the environment by nonnuclear energy systems, but their
contribution to population exposure doses are even less significant. The same
could be said about the other radionuclides released from nuclear facilities
that are not listed here.
The findings with respect to radiological quality assurance aspects are
the wide differences in the data (both estimated and measured) on the release
rated of the various radionuclides in nuclear energy systems and the sparseness
of the data on radioactive material releases from nonnuclear energy systems.
For nuclear systems, better assessments of the tritium generation rates in the
coolant and in the fuel, and the subsequent release rates to the atmosphere
and to streams are required. Better assessments of the escape rates during
mining (particularly open-pit mining) and milling operations, and from the
mill tails are also required. Assessments of the uranium atmospheric releases
from uranium mills, conversion plants, enrichment plants, and fabrication plants
are required, in addition to assessments of reprocessing radioactive releases
once reprocessing operations are reestablished.
For nonnuclear energy systems radon releases appear to be the potential
predominant source of population exposure, and for this reason measurements of
radon releases for coal ash piles, spent oil shale piles, and from geothermal
resources need to be assessed.
Finally, the problems associated with the high-level radioactive wastes
from spent-fuel reprocessing are yet to be resolved, and they were not addressed
in this study.
*
125
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D.C., May 1975).
30. Bureau of National Affairs, Inc., Energy Users Report, No. 86 (Washington,
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ment and Safety Administration Information Report 1025 (Washington, D.C.,
1975).
32. R. P. Caldwell et al., "Radioactivity in Coal Mine Drainage," in
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128
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130
*
*U.S. GOVERNMENT PRINTING OFFICE: 1977 - 784-870/137 Region No. 9-1
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TECHNICAL REPORT DATA
(Please read Instructions on the reverse before completing)
1. REPORT NO.
EPA-600/7-77-082
3. RECIPIENT'S ACCESSION NO.
». TITLE AND SUBTITLE
POTENTIAL RADIOACTIVE POLLUTANTS RESULTING FROM
EXPANDED ENERGY PROGRAMS
5. REPORT DATE
August 1977
6. PERFORMING ORGANIZATION CODE
7. AUTHOR(S)
Hong Lee, Thomas 0. Peyton (Greenfield Attaway and
Tyler, Inc.), Robert V. Steele, and Ronald K. White
8. PERFORMING ORGANIZATION REPORT NO.
CRESS No. 6
9. PERFORMING ORGANIZATION NAME AND ADDRESS
Center for Resource and Environmental Systems Studies
Stanford Research Institute
333 Ravenswood Avenue
Menlo Park, California 94025
10. PROGRAM ELEMENT NO.
1NE 625
11. CONTRACT/GRANT NO.
68-03-2375
12. SPONSORING AGENCY NAME AND ADDRESS
U.S. Environmental Protection Agency-Las Vegas
Office of Research and Development
Environmental Monitoring and Support Laboratory
Las Vegas, Nevada 89114
13. TYPE OF REPORT AND PERIOD COVERED
Final Contract Report
14. SPONSORING AGENCY CODE
EPA/600/07
IS. SUPPLEMENTARY NOTES
16. ABSTRACT
An effective environmental monitoring program must have a quality assurance
component to assure the production of valid data. Quality assurance has many
components: calibration standards, standard reference materials, standard reference
methods, interlaboratory comparison studies, and data validation. The purpose
of this document is to identify and document the potential radioactive pollutants
that could result from the expanded energy program and for which quality assurance
programs must be provided.
The radionuclide releases and the resulting population exposure doses from
several energy systems for four projected energy utilization scenarios were cal-
culated and compared. The energy system components examined were: coal mining,
processing, combustion, and ash disposal; coal gasification and liquefaction;
oil shale mining, processing, residue disposal and product utilization; geo-
thermal and development and operations; uranium mining, milling, conversion,
enrichment and fabrication; nuclear reactor operations; and fuel reprocessing
and waste disposal.
17.
KEY WORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
b.lDENTIFIERS/OPEN ENDED TERMS C. COSATI Field/Croup
coal
oil shale
geothermal
nuclear systems
radiation exposures
radiation dosage
alternative energy pro-
grams
energy systems
human exposure calculatio
radioactive pollutants
radiological projections
08 G,I
18 F,H
20 H
18. DISTRIBUTION STATEMENT
RELEASE TO PUBLIC
19. SECURITY CLASS (TMs Report)
UNCLASSIFIED
21. NO. OF
142
20. SECURITY CLASS (Thispage)
UNCLASSIFIED
22. PRICE
EPA Form 2220-1 (R«v. 4-77) PREVIOUS EDITION is OBSOLETE
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