p
EPA-520/9-73-003-B
ENVIRONMENTAL ANALYSIS
OF THE URANIUM FUEL CYCLE
m
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PART I - Fuel Supply
U.S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
m
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ENVIRONMENTAL ANALYSIS
OF THE URANIUM FUEL CYCLE
2
PART I - Fuel Supply
October 1973
U.S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
Field Operations Division
Washington,D.C. 20460
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FOREWORD
The generation of electricity by light-water-cooled nuclear power
reactors using enriched uranium for fuel is experiencing rapid growth in
the United States. This increase in nuclear power reactors will require
similar growth in the other activities that must exist to support these
reactors. These activities, the sum total of which comprises the uranium
fuel cycle, can be conveniently separated into three parts: 1) the
operations of milling, conversion, enrichment, fuel fabrication and
transportation that convert mined uranium ore into reactor fuel, 2) the
light-water-cooled reactor that burns this fuel, and 3) the reprocessing
of spent fuel after it leaves the reactor.
This report is one part of a three-part analysis of the impact of
the various operations within the uranium fuel cycle. The complete
analysis comprises three reports: The Fuel Supply (Part I), Light-Water
Reactors (Part II), and Fuel Reprocessing (Part III). High-level waste
disposal operations have not been included in this analysis since these
have no planned discharges to the environment. Similarly, accidents,
although of potential environmental risk significance, have also not been
included. Other fuel cycles such as plutonium recycle, plutonium, and
thorium have been excluded. Insofar as uranium may be used in high-
temperature gas-cooled reactors, this use has also been excluded.
The principal purposes of the analysis are to project what effects
the total uranium fuel cycle may have on public health and to indicate
where, when, and how standards limiting environmental releases could be
effectively applied to mitigate these effects. The growth of nuclear
energy has been managed so that environmental contamination is minimal
at the present time; however, the projected growth of this industry and
its anticipated releases of radioactivity to the environment warrant a
careful examination of potential health effects. Considerable emphasis
has been placed on the long-term health consequences of radioactivity
releases from the various operations, especially in terms of expected
persistence in the environment and for any regional, national or world-
wide migration that may occur. It is believed that these perspectives
are important in judging the potential impact of radiation-related
activities and should be used in public policy decisions for their
control.
Comments on this analysis would be appreciated. These should be
sent to the Director, Criteria and Standards Division of the Office
of Radiation Programs.
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs
iii
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Contents
Foreword ill
1.0 Overview of the Uranium Fuel Industry 1
1.1 Introduction 1
1.2 Industry Operations 5
1.3 Environmental Impact of Fuel Supply 9
1.4 Control Technology Effectiveness 12
1.5 Summary and Conclusions 19
2.0 Uranium Milling 21
2.1 General Description of the Milling Process 21
2.2 The Model Mill 24
2.3 Releases of Radioactive Effluent from Uranium Mills.... 25
2.4 Radiological Impact of Uranium Mills 35
2.5 Health Effects Impact of Uranium Mills 35
2.6 Control Technology 35
2.7 Uranium Mill Tailings Piles 51
2.8 Summary 72
3.0 Conversion Facilities 73
3.1 General Description of the Uranium Conversion Process.. 73
3.2 The Model Conversion Facilities 78
3.3 Release of Radioactive Effluents from Conversion
Facilities 80
3.4 Radiological Impact of Conversion Facilities 82
3.5 Health Effects Impact of a Model Conversion Facility... 82
3.6 Control Technology 82
3.7 Environmental Controls 92
3.8 Summary 95
4.0 Uranium Enrichment Facilities 96
4.1 Description of the Uranium Enrichment Industry 96
4.2 The Model Facility 100
4.3 Release of Radioactive Effluents from Enrichment
Facilities 100
4.4 Radiological Impact of Enrichment Facilities 105
4.5 Health Effects Impact of a Model Enrichment Facility... 105
4.6 Control Technology 110
4.7 Environmental Controls - Enrichment Facilities Ill
4.8 Summary Ill
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Contents (Continued)
Page
5.0 Fuel Fabrication and Scrap Recovery Ill
5.1 Description of the Fuel Fabrication Process Ill
5.2 The Model Facility 114
5.3 Radionuclide Effluents from Fuel Fabrication
Facilities 118
5.4 Radiological Impact of a Model Fuel Fabrication
Facility 124
5.5 Health Effects Impact of a Model Fuel Fabrication
Facility 124
5.6 Control Technology 124
5.7 Environmental Controls - Fuel Fabrication 132
5.8 Summary 136
6.0 Transportation 135
6.1 Description and Growth Patterns 136
6.2 Shipping Containers 140
6.3 Exposure Levels 140
6. 4 Radiological Impact 141
6.5 Health Effect Impact 145
6.6 Cost Effectiveness of Reducing Transportation
Exposure 145
Appendixes
Appendix A. Exposure Pathway, Radiation Dose and Health
Effects A_l
Appendix B. Costs of Control Technology B-l
References ^ 1
vi
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Figures
Page
Section 1
1-1 Model facility relationships in the uranium fuel cycle
for LWR power plants 3
1-2 Proj ected nuclear fuel cycle needs 6
Section 2
2-1 Uranium mill process diagram 23
Section 3
3-1 UF^ production-hydrofluor process block diagram 75
3-2 UF6 production-wet solvent extraction-fluorination block
diagram 76
Section 4
4-1 Mode of operation for gaseous diffusion plants 99
4-2 Model plant characteristics 101
Section 5
5-1 Fuel fabrication-chemical processing (ADU) block diagram 115
5-2 Model fuel fabrication plant 119
Section 6
6-1 Simplified schematic of transportation requirements for the
LWR nuclear power industry 137
6-2 Cost effectiveness for segregation of low level wastes with
an additional shielding compartment on the truck (1 truck per
reactor year) 150
6-3 Impact reduction available with route control through various
areas of population density (impact from stops remains
constant) 150
vii
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Tables
Page
Section 1
1-1 Number of LWR's supported by model fuel supply
facilities
1-2 Radiological impact of model fuel supply facilities -
current effluent control procedures ..... . ................. 11
1-3 Radiological impact of model fuel supply facilities -
optimum effluent control procedures ....................... 13
1-4 Cost effectiveness of control technology for the fuel
supply .......................... . ......................... 14
Section 2
2-1 Predicted airborne releases of radioactive materials from
the Highland Uranium Mill 27
2-2 Concentrations of radioactive effluents in waste liquor
from a uranium mill 29
2-3 Estimates of quantities of radionuclides seeping through
the impoundment dam of a uranium mill initially and at
2-1/4 years 31
2-4 Analysis of plant tailings effluents from the Humeca
Uranium Mill (alkaline leach process) 33
2-5 Discharge of radionuclides to the environment from a model
uranium mill (acid leach process) 34
2-6 Radiation doses to individuals in the general population
in the vicinity of a model mill, from inhalation 36
2-7 Radiation doses to individuals in the general population
in the vicinity of a model mill, from drinking water...... 37
2-8 Aggregate dose to the general population in the vicinity
of a model mill 38
2-9 Committed health effects to the general population in the
vicinity of a model mill (acid leach process) 39
viii
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Page
2-10 Cost and reduction of effluents for control technology
for mills 42
2-11 Radiological impact of airborne effluents vs controls for
a model uranium mill 52
2-12 Radiological impact of waterborne effluents vs controls
for a model uranium mill 53
2-13 Radiological impact of airborne effluents vs controls
for a model uranium mill tailings pile (250 acres) 61
2-14 Experimental and predicted values of radon-222 emanating
from tailings piles 62
2-15 Radon-222 decay scheme 65
2-16 Calculated alpha dose rates in mrad/yr from inhalation of
short-lived 222Rn daughter products to the basal cell
nuclei of segmental bronchi. 69
2-17 Health effects resulting from the 100-year dose commitment
from radon-222 emanation from a uranium mill tailings pile
(250 acres) 73
Section 3
3-1 Model uranium conversion plant 79
3-2 Discharges of radionuclides to the environment from a
model conversion facility using the wet solvent extraction
process 83
3-3 Discharges of radionuclides to the environment from a
model conversion facility using the hydrofluor process.... 84
3-4 Radiation doses to individuals in the general population
in the vicinity of a model conversion plant, through
inhalation 85
3-5 Radiation doses to individuals in the general population
in the vicinity of a model conversion plant, through
drinking water 86
3-6 Aggregate dose to the general population in the vicinity
of a model conversion facility 87
ix
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Page
3-7 Health effects to members of the general population
in the vicinity of a model conversion facility using
the wet solvent extraction process 88
3-8 Health effects to members of the general population
in the vicinity of a model conversion facility using
the hydrof luor process 89
3-9 Airborne waste treatment effectiveness and costs for the
hydrof luor process 91
3-10 Radiological impact of airborne effluents vs controls for
uranium conversion facilities 93
3-11 Radiological impact of waterborne effluents vs controls
for uranium conversion facilities 94
Section 4
4-1 Distances to gaseous diffusion plants from nearby popula-
tion centers and UF^ production plants. 98
4-2 Discharges of radionuclides to the environment from a
model enrichment facility 104
4-3 Radiation doses to individuals in the general population
in the vicinity of a model enrichment plant, through
inhalation 106
4-4 Radiation doses to individuals in the general population
in the vicinity of a model enrichment plant, through
drinking water 107
4-5 Aggregate dose to the general population in the vicinity
of a model enrichment facility 108
4-6 Health effects to members of the general population in
the vicinity of a model enrichment facility 109
4-7 Radiological impact of airborne effluents vs controls for
uranium enrichment facilities 112
4-8 Radiological impact of waterborne effluents vs controls
for uranium enrichment facilities 113
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Page
Section 5
5-1 LWR fuel fabrication plants 116
5-2 Fuel fabrication plants - site size and demography 117
5-3 Discharge of radionuclides to the environment from a model
uranium oxide fuel fabrication facility 121
5-4 Reported and estimated effluent uranium quantities 123
5-5 Radiation dose to members of the general population from
a model uranium oxide fuel fabrication facility through
inhalation 125
5-6 Radiation doses to individual in the general population
in the vicinity of a model uranium fuel fabrication plant
through drinking water 126
5-7 Aggregate dose to the general population in the vicinity
of a model uranium fuel fabrication plant 127
5-8 Health effects to members of the general population in the
vicinity of a model fuel fabrication plant 128
5-9 Uranium fuel fabrication and scrap recovery gaseous waste
treatment effectiveness and costs 130
5-10 Uranium fuel fabrication and scrap recovery liquid waste
treatment effectiveness and cost 133
5-11 Radiological impact of airborne effluents vs controls for
a model uranium fuel fabrication plant 134
5-12 Radiological impact of waterborne effluents vs controls
for a model uranium fuel fabrication plant 135
Section 6
6-1 Summary of transportation parameters for the LWR nuclear
power industry 138
6-2 Transportation requirements for a 900 LWR nuclear power
program (low enriched uranium oxide fuel) 139
xi
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Page
6-3 Individual dose at a given distance from the apparent
centerline of the shipping route for the passage of one
shipment 143
6-4 Health effects to members of the general population from
transportation of radioactive materials associated with
nuclear fuel cycle 146
Appendix A
A-l Average (x/Q) values vs distance for h = 10 meters A-3
A-2 Dispersion factors for airborne and waterborne pathways... A-6
A-3 Population models for air and water pathways A-7
A-4 Dose conversion factors for the airborne pathway A-9
A-5 Dose conversion factors for airborne insoluble particu-
lates A-10
A-6 Dose conversion factors for airborne soluble particulates. A-12
A-7 Dose conversion factors for the water pathway A-13
A-8 Dose conversion factors for uranium for the water pathway. A-15
A-9 Dose conversion factors for radium-226 for the water path-
way A-16
A-10 Dose conversion factors for the water pathway for soluble
radionuclides calculated from the recommendations of the
ICRP A-17
A-ll Health effects conversion factors A-18
A-12 Airborne pathway dose calculations for model facilities -
current best technology A-20
A-13 Waterborne pathway dose calculations for model facilities -
current best technology A-21
A-14 Health effect calculations for model facilities - current
best technology A-22
A-15 Radiation dose and health effect calculations - flowsheet. A-24
xii
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A-16 Doses to critical organs for naturally occurring
uranium and an air release of one curie of uranium
from a model plant A-27
A-17 Health effects resulting from the 100-year dose commitment
from uranium - 30 years of facility operations A-30
xiii
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PART I. FUEL SUPPLY
1.0 Overview of the Uranium Fuel Industry
1.1 Introduction
Because of the rapid increase in the use of light-water-
cooled nuclear reactors to generate electricity, there is par-
allel growth in the basic industry that provides enriched urani-
um fuel for these operations. This industry includes various
operations broadly classified as: (1) milling, (2) conversion
of uranium oxide (U30g) to uranium hexafluoride (UFg), (3) en-
richment, (4) fuel fabrication, and (5) radioactive material
transportation between these facilities. Radioactive waste pro-
ducts are associated with each of the above activities. This
report examines the predominant facilities and operations within
these five categories which have the highest potential for envi-
ronmental impact. Fuel reprocessing also relates to fuel sup-
ply; however, this activity has been analyzed separately in part
3 of the environmental analysis of the uranium fuel cycle.
Natural uranium (0.71% uranium-235) ore is rained and milled
to a concentrate containing about 85% U.,0-. The conversion step
purifies and converts U000 to UF£, the chemical form in which
Jo 0
uranium is fed to the enrichment plants. This is followed by
isotopic enrichment where the uranium-235 concentration of the
uranium feed is increased to the design specification (usually 2
to 4% uranium-235) of the power reactor by a gaseous diffusion
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process. The greatest portion of uranium becomes a plant tail
impoverished in uranium-235 and is stored in cylinders as UF,.
o
The enriched UF, portion is processed into UCL pellets, loaded
into alloy tubing, and finally fabricated into individual fuel
element bundles. The tube bundles fuel the reactor. These pro-
cesses are shown in simplified form in figure 1-1 as they relate
to a 1,000 megawatt electrical (1000 MW(e)) power reactor. This
figure also includes the basic parameters for assumed model plants
for each of these operations. Such models are important to a
consistent analysis of the environmental impact of the various
operations.
The "model" facilities described herein represent the better
features of current practice; as such, they are not exemplary
facilities and health hazards from their operation are not nec-
essarily acceptable. The model plant sizes are generally simi-
lar to those which have been described by Pigford (1) and the
Atomic Energy Commission (2). Expressing the model operations
in terms of 1,000 MW(e) equivalents gives a common base for the
comparison of environmental dose and risk commitments as a func-
tion of radioactive waste control technology and cost."Table
1-1 indicates the relation between the various components in
terms of 1,000 MW(e) equivalents.
Each step in the fuel supply generates radioactive wastes.
Most of this material is controlled becoming solid waste of vari-
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CONVERSION:
MODEL PLANT
6,000 MT U3 O8 is converted to
7,600 MT UF0 each year to supply
the average requirements of 28 LWR
power plants.
MODEL MILL
600,000 MT ore is milled to
produce I, 140 MT 1)3 Og in
ore concentrate each year, to
supply the average requirements
of 5.3 LWR power plants.
MODEL MINE
600,000 MT ore is mined each
year to supply the average
requirements of 5.3 LWR power
plants.
ENRICHMENT:
MODEL PUNT
4,700 MT enriched UF6 to
supply the average requirements
of 90 LWR power plants is
separated from 29, 000 MT UFO
each year. 24,300 MT UF6
depleted to 0.25% 235u are
added to the stockpile each year.
FUEL REPROCESSING:
MODEL PLANT
Used fuel containing I,500 MTU
is reprocessed each yeor to meet
the average requirements of 43
LWR power plants.
FUEL FABRIC AT I ON:
MODEL PLANT
I, 350 MT enriched UF6 is
converted to 1,040 MT
which is sealed inside fuel
assemblies to meet the average
requirements of 26 LWR power
plants.
THE MODEL UNIT
LWR POWER PLANT
A plant of 1000 MWe capacity
requires a lifetime annual average
of 40 MT UO2 (35 MTU enriched
to 3.2% 235U). Used fuel
removed from the reactor is assumed
to average 0.84% 235|j anc| to
have produced 33,000 MWD/MTU.
WASTE STORAGE AND DISPOSAL
Transportation represented by
the connecting arrows—»-
MT = metric ton
= 2,205 pounds
MTU = metric ton of uranium
Figure 1-1. Model facility relationships in the uranium fuel cycle for LWR power plants
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Table 1-1
Number of LWR's supported by model fuel supply facilities
Type of model fuel Number of on-line LWR's
supply facility Supported by model facility
Mill 5.3
Conversion 28
Enrichment 90
Fabrication 26
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ous kinds; a small amount is released under controlled conditions
to air and water in most steps in the cycle. The limited number
of radionuclides involved in these releases are naturally occur-
ring radionuclides which make up part of the radiation background
to which all people are exposed. Mills and conversion facilities
will release uranium-238 and its daughters including uranium
-234, thorium-230, radium-226, and radon-222. By the time the
uranium leaves the conversion facility, it is purified to the
point where only uranium-238 uranium-235, and uranium-234 are
present.
A projection of the growth of fuel supply facilities from
1980 up to the year 2000 is shown in figure 1-2. The number of
such facilities up to about 1980 will be about the same as cur-
rent industry capacity, which is discussed below. The actual
total number of LWR's installed can be estimated by dividing the
number of on-line plants by the fraction of time that they are
on line (load factor). For the purpose of this report, no elec-
trical power generation based on plutonium fuel was assumed.
1.2 Industry Operations
1.2.1 Mills
The purpose of milling is to obtain U.O- in such a form that
it can be converted, enriched, and eventually fabricated into
reactor fuel. Milling of uranium ore must be done to separate
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800
700
600
500
400
300
200
100
7
LWRs
AND POWER- —
MILLS
CONVERSION-^//
FABRICATION -
NRICHMENT
40
30
20
10
1970
1980
1990
2000
Figure 1-2. Projected nuclear fuel cycle needs
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uranium from extraneous rock. This process is accomplished by
mechanical crushing of the ore so that it can be dissolved by an
acid leach, solvent extraction process. Uranium is purified and
concentrated in solvent extraction steps, separated by thickening
and centrifuging, and finally calcined and pulverized for pack-
aging in 55 gallon drums for shipment. An alternate process for
uranium milling is to use a carbonate leach process, however,
there are no particular environmental advantages over the acid
leach process.
1.2.2 Conversion Facilities
A typical conversion facility converts uranium to uranium
hexafluoride (UF&) prior to enrichment in sufficient quantities
to support 28 1,000 MW(e) reactor equivalents. The two facili-
ties in operation at the present time have capacity considerably
in excess of that needed. It is expected that about 30 such facil-
ities will be operational by the year 2000.
Plans are underway at the present time to start the recycling
of uranium recovered from the fuel reprocessing step. It appears
that each of the fuel reprocessing facilities may have a UF- pro-
6
duction capability for recycle of recovered uranium. This approach
would result in UFfi production facilities being located at fuel
reprocessing sites. A total of 3 or 4 such facilities could be
operational by about 1980.
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At this time, no significant technological advances can be
projected that will affect environmental considerations from
these plants. Accordingly, it is projected that UF, production
plants in 1980 will resemble in most respects the plants of to-
day, but, may be about double their size.
1.2.3 Enrichment Facilities
Enrichment services for the nuclear industry are supplied by
the three AEC-owned and contractor-operated gaseous diffusion
plants. Essentially any one of the existing plants operated
independently would be adequate to meet the current demand for
nuclear power plant fuel of about 5,000 metric tons of separa-
tive work units (SWU). However, since all three installations
are operated in a combined fashion, it is difficult to isolate
the environmental consideration of an individual plant.
It is anticipated that the capability of the existing three-
plant complex will be increased from about 10,000 MT SWU to
22-27,000 MT SWU by 1980. This increased capability, enough to
support about 230 on-line reactors, will be achieved without the
construction of major new gaseous diffusion facilities and will
be accomplished primarily by improving and upgrading present
units. Thus, no new processing plants are required to meet the
projected 1980 industry demands. The power demand in the year
2000 will require about 4 such upgraded enrichment plant complexes,
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1.2.4 Fuel Fabrication Facilities
One model fuel fabrication plant is sufficient to support 26
on-line 1,000 MW(e) reactors. There are currently 10 commercial
plants which are capable of performing all or part of the current
fuel fabrication process. According to projections (3), a sub-
stantial expansion of production capacity is anticipated in the
next 5 years. Existing plants with some shutdowns and some new
additions will produce fuel assemblies from an enriched UF, feed.
6
The industry is expected to comprise some 20 to 30 plants by
the year 2000. Most of the chemical effluents will be eliminated
by future process improvements.
1.2.5 Transportation
Transportation of fuel and waste products will increase sub-
stantially over the next several years as the number of reactors
and supporting facilities increases. Both rail and truck ship-
ments will be made. The primary mode of exposure to the popu-
lation will be from direct radiation resulting from the passage
of such shipments along shipments routes.
1.3 Environmental Impact of Fuel Supply
With the exception of transportation, the various components
of the uranium fuel industry introduce naturally-occurring radio-
active materials into the environment through discharges to both
air and water. The results are long-term radiation exposure to
the skeleton and other organs of the body, especially the luag.
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10
The doses and resultant effects were calculated using the tech-
niques described in appendix A.
The impacts of the various representative facilities are shown
in table 1-2 in terms of discharges to the general environment,
the resultant individual and population doses committed, and poten-
tial health effects. The doses were assumed to be delivered in
two principal ways: (1) directly from the effluent as it dispersed
in the surrounding air or water environment, and (2) through en-
trainment in environmental pathways such as ingestion through
food chains. Dose commitments through environmental pathways
after the original deposition appear to be small in comparision
to immediate exposure from the effluents directly, therefore,
the environmental considerations for these radionuclides were
based on the plant operations themselves; the dose commitments
and resultant effects from environmental buildup were found to
be mininal by comparison.
The data in table 1-2 indicate that, with the exception of
radon, discharges of naturally-occurring materials for all model
facilities controlled to current levels of good practice are on
the order of 4 Ci/yr or less and that most individual organ doses
are grouped below 70 mrem/yr. With the exception of milling,
bone doses range from 0.6 to 3 mreni/yr; lung doses from 1 to 70
mrem/yr. Bone doses from milling were calculated to be as high
as 13 mrem/yr and lung doses as high as 450 mrem/yr in a case
with a short site boundary distance, a situation that occurs
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Table 1-2
Radiological impact of model fuel supply facilities - current effluent control procedures
Urani urn and
daughters
Facility discharged
Air Water
(Ci/yr) (Cl/yr)
Millb 0.2 4
Conversion
wet solvent 0.02 2
hydro fluor 0.06 0.8
Enrichment 0.05 0.6
Fabrication 0.005 0.5
Transportation
Maximum radiation
dose at boundary i^diate health 100 yr health
(mren/yr) effects committed effects committed
Air
450
(lung)
30
(lung)
70
(lung)
1
(bone)
10
(lung)
Water (effects/facility-30 yr) (effects/facili ty-30 yr)
13 0.10 0.006
(bone)
2 0.4 0.002
(bone)
1 0.2 0.005
(bone)
0.7 0.1 0.004
(bone)
0.6 0.1 0.0005
(bone)
0.02C 0
Total health >
effects commi ttecr
(effects/facility/yr)
A
0.1
0.4
0.2
O.I
0.1
0.02
Calculated assuming the dose is that which follows long term chronic exposure - but only to that fraction of intake resulting from
30 years of plant operations; uranium effluents only.
bRadon-222 is not included.
cTotal health effects associated with 30 years of operation of 1 GW(e) cower reactor.
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12
infrequently because of relatively remote siting of mills. The
resultant health effects committed for 30 years operation of such
facilities range as high as 0.4 effects taking into account 100-
year exposures from environmentally deposited uranium and daugh-
ters (radon excluded).
The data in table 1-3 indicates that, with the exception of
radon, discharges of naturally-occurring materials from model
fuel facilities using optimum effluent control technologies are
held to less than 1 curie/yr and that individual organ doses are
below 15 mrem/yr. Additional effluent controls were required on
uranium mills and conversion facilities.
1.4 Control Technology Effectiveness
A number of technologies are available to influence the en-
vironmental impact of discharges of uranium and daughter products.
Several of these were analyzed in other sections of this report
for an optimum control level based on current practice and, wher-
ever possible, a cost-effectiveness of alternative control options.
Table 1-4 shows the results of this analysis in terms of the total
fuel supply industry. The data were determined for present worth
costs of averting annual health effects fron operations of repre-
sentative facilities.
The greatest effectiveness of reducing health effects occurs
from control options applied to atmospheric discharges from con-
version and fabrication facilities. Holding pond treatment for
liquid discharges at conversion facilities and enrichment opera-
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Table 1-3
Radiological impact of model fuel supply facilities - optiura effluent control procedures
Facility
Mill
Conversion
wet solvent
Hydrofluor
Enrichment
Fabrication
Tr«ancnnir*1'a'f"i nn
Optimum level of
controls
Additional bag filters and
HEPA filters; catch basin
and pumps, to present con-
trols
Additional bag filters and
water treatment to present
controls
Additional bag filters to
present controls
Present controls
Present controls
Prp«;pnt- rontrn"l<;
Uranium and daughters
discharged
air water
(Ci/yr)
0.004
0.01
0.02
0.05
0.05
n
(Ci/yr}
0
0.2
0.8
0.06
0.5
n
Maximum radiation
dose at boundary
air water
(mrem/yr) (mrem/yr)
11 0
(lung)
3 0.3
(lung) (bone)
7 1
(lung) (bone)
1 0.7
(bone) (bone)
10 0.6
(bone) (bone)
total health effects
commi tted
(effects/facility-30 yr)
0.002
0.03
0.2
0.1
0.1
n.n?
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Table 1-4
Coit effectiveness of control technology for the fuel supply
Operation
Milling
Conversion -
Wet solvent
Conversion -
Hydrofluor
Enrichment
Fabrication
a - Effects not
Health effects
before control
System control option (effects/facility-30 yr)
Wet dust collector (yellow cake drying)
Wet dust collector (ore crushing)
HEPA System (yellow cake drying)
Bag filter (ore crushing)
Clay core dam
Seepage return added to clay core dam
Bag filters
2nd bag filters in series
Settling ponds
Additional chemical treatment
Bag filters
2nd bag filters in series
Settling ponds
Additional chemical treatment
(Airborne releases)9
Holding pond
Scrubber and prefilter
1 HEPA
2nd HEPA in series
3rd HEPA in series
Settling tank
Precipitation and flocculation
critical compared to the water pathway
>1
8 x ID'2
1 x TO'2
3 x ID'3
>9
0.09
<1.5
0.015
>4
0.4
>4
0.04
>1
0.15
---
>10
>500
0.5
0.005
>0.9
0.09
Health effects Present
after control worth
(effects/facility-30 yr) (1970 $)
8 x TO'2
1 x ID'2
3 x 10~3
2 x 10'4
0.09
0
0.015
0.001
0.4
0.04
0.04
0.004
0.15
0.01
0.13
>500
0.5
0.005
0.0002
0.09
0.002
52,000
280,000
10,000
280,000
4,000,000
600,000
1 1 ,000
11,000
240,000
14,000,000
190,000
190,000
240,000
14,000,000
240,000
530,000
530,000
530,000
1,000,000
4,300,000
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15
tions are next in order of effectiveness. The control options
chosen are already operational in most facilities; therefore,
even though the effectiveness of reducing health effects for
other options is low, industry practice has already achieved
these levels of control. In many cases such controls have pro-
bably resulted from attempts to recover economically valuable
uranium as part of the process involved.
1.4.1 Airborne Discharge Control
Mills. Gaseous and particulate effuents are controlled at
mills in three primary places: the ore crushing area, the fine
ore bins, and the yellow cake packaging and drying area. Wet
dust control systems are generally used in the ore crushing area
and the fine ore storage bins; wet scrubbers and bag filters in
various combinations are used in the drying and packaging area.
Conversion Plants. The major treatment of gaseous effluents
for conversion facilities is a wet scrubber system combined with
HF recovery and a HL burner. Bag filter systems are used to con-
trol uranium dusts in both processes (4>J5). The wet solvent ex-
traction system also uges absorption towers for scrubbing out
oxides of nitrogen (2). Gaseous effluents escaping from the
scrubber and bag filter,systems are released through stacks with-
out further treatment.
The scrubber system.configuration consists of a venturi sec-
tion, a liquid-gas separator (demister), fans and associated
motors and controls. The venturi section is a vertical conver-
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16
gent-divergent section connected to the separator system by a
horizontal elbow. Waste gas from the process systems enters
from the top into the converging system. A concurrent flow of
scrubber-liquid is introduced into the converging venturi where
liquid-gas mixing takes place. The gas and liquid flow to the
separator with gaseous flow upward and the liquid exiting down-
ward through a port. The fan draft moves the gas through a
final demister after which it is exhausted (6).
Bag filter systems or fabric collectors use woven or felted
fabric bags. Dust deposits in the porous fabric and the pres-
sure increases until the dust is removed by manual or automatic
means. Rapping, shaking gear, or automatic flow reversal mech-
anisms are used to remove and collect the dust (7).
Enrichment Facilities. Effluent treatment is part of the
recovery of uranium from gaseous streams because of dollar value
(8). Gaseous and airborne materials are removed from streams
with cold traps and aluminum traps. The efficiency of uranium
removal is unknown (8). Descriptions of the trap systems used
are not presently available.
Fabrication Plants. The system for conversion of UF^ to UO
6 2
is equipped with a scrubber-demister and one high efficiency
particulate air (HEPA) filter for dust removal (2_). Scrap recov-
ery operations exhaust chemical systems through a scrubber-
demister. Each system is equipped with a HEPA filter for urani-
um dust removal. The process systems handling U0_ are assumed
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17
to use two HEPA filters independent of those in the conversion
and scrap recovery systems.
The effectiveness of high efficiency particulate air filters
for uranium was assumed to be that for measurements made on pluto-
nium entering and leaving two banks of HEPA filters (9). The
measured fractional removal of plutonium from air passing through
two HEPA filters was 0.99999 (9). Assuming uranium aerosols have
the same characteristics as plutonium and the first filter is
the most effective, the fractional removal was apportioned as
0.999 for the first filter and 0.99 for the second filter. A
third HEPA filter in series was 94% efficient for removal of
particulate plutonium passing through the first and second fil-
ters (9). The usual testing procedures for HEPA filters state
removal efficiencies of 99.97% for 0.3 pm dioctyl phthalate
(DOP) as a raonodispersed aerosol (10). This measurement is a
quality assurance test rather than an in-place filter perfor-
mance test.
Transportation. There are no planned releases of radioactive
materials from transportation activities supporting the uranium
fuel cycle.
1.4.2 Liquid Discharge Control
Liquid waste streams are treated both to recover uranium and
to reduce concentrations of entrained pollutants.
A basic scheme of liquid waste treatment is to mechanically
separate solid waste particles from the water of the waste
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18
stream; if necessary, the waste is changed chemically so that it
will become particles which can be separated. Residual chemicals
in the waste stream must then be neutralized, so that the efflu-
ent is neither too acidic nor basic.
Processes used are high efficiency centrifuges, chemical treat-
ments for flocculation, precipitation and neutralization, filtra-
tion and settling.
Waste matter which is dissolved is usually in the form of
ions which may be separated from the waste stream by passing it
through beds of small spheres of chemically treated resins on
which the ions are absorbed. Chemicals may be added to the waste
stream to change its characteristics (e.g., neutralization) to
cause the dissolved waste to become particles which may settle
out or be separated from the water by filters or centrifuges.
Flocculation is the addition of chemicals which either causes
particles to be formed which are large enough to settle rapidly
or causes the formation of large, quick-settling particles to
which small, slow-settling particles become attached.
Settling basins or ponds are a preferred treatment for
liquid waste streams because, once constructed, their use re-
quires little power and maintenance. They are usually an exca-
vation about 4 feet deep and may be several acres in size.
Losses of the waste liquid into the ground are usually consi-
dered undesirable; if the soil is such that much seepage into
the ground is likely, the ponds may be lined with special soils
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19
or artificial liners of plastic or chemically-sealed fabric.
The effectiveness of individual processes varies widely,
depending upon the nature of the waste. At one conversion
facility, neutralization and settling of liquid waste reduced
its uranium content by a factor of 500, its radium content by a
factor of 4, and its thorium content by a factor of 180 (11).
Waste concentrations are commonly reduced further before release
by simple dilution.
1.5 Summary and Conclusions
This analysis of the potential environmental impact of the
uranium fuel supply industry, and of the feasibility of mini-
mizing it in the various operations comprising the industry, has
involved consideration of:
1. projections of the growth of milling, conversion, enrich-
ment, fabrication and transportation operations through the
year 2000,
2. the technology for influencing discharges of natural-
ly-occurring uranium and daughters, including estimates of
relative costs based on initial cost and operating expense,
3. the distribution throughout the environment of the rad-
ionuclides released during normal operation,
4. estimates of radiation doses to the affected organs of
individuals and populations for sites with assumed atmospheric
parameters and typical aquatic environments, and
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20
5. health effects expected to be associated with the esti-
mated population doses.
The major conclusions derived from these considerations are
as follows.
1. The various operations in the fuel supply industry are,
in combination, an integral part of the entire nuclear power
industry. Because most of the radionuclides involved in all
parts of the industry are naturally-occurring uranium and
daughters, the industry can be treated as a combined opera-
tion for purposes of evaluating its contribution to the over-
all discharge of radionuclides to the environment and their
effects.
2. The quantities of uranium and daughter products dis-
charged, with the exception of radon and radon daughters,
can be maintained to less than 1 curie per year for repre-
sentative plants with currently available and commonly used
control technologies. Consequently, if projections of nuc-
lear power are substantially correct, the overall discharges
of such materials would not be expected to be large through
the year 2000 since the various facilities support several
reactor requirements.
3. Control technology exists to avert health effects from
discharges of uranium and daughter products although many
such technologies are a part of uranium recovery processes.
The most cost-effective system to apply for risk reduction
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21
appears to be HEPA filters at fabrication plants followed by
bag filters at conversion facilities and holding ponds for
liquid effluent treatment at various facilities.
4. The consequences of the environmental buildup of uran-
ium and daughter products, although many have long half-
lives, do not appear large. The major route of exposure is
direct interaction of populations with the effluents immedi-
ately following discharge. Doses to individuals and organs
(with the exception of doses from radon-222) have been esti-
mated to be less than about 70 mrem/yr, with the exception
of lung dose from airborne discharges from a uranium mill,
which for short site boundary distances can go up to about
450 mrem/yr. These doses can be held to less than 15 mrem/yr
by additional commonly used control technologies.
2.0 Uranium Milling
2.1 General Description of the Milling Process
A uranium mill extracts uranium from ore. The product is a
semirefined uranium compound (U,0g) called "yellowcake" which is
the feed material for the production of uranium hexafloride (UF,).
6
About 20 mills are currently operating in the United States with
average outputs ranging from 400 to 7,000 MT/yr. These mills
are characteristically located in arid, isolated regions of the
west.
Eighty percent of the yellowcake is produced by a sulfuric
acid leach process; the remainder by a sodium carbonate, alka-
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22
line leach process. The principal steps in the acid-leach, sol-
vent-extraction process diagrammed in figure 2-1, are:
a. Ore is blended and crushed to pass through a 2.5 cm (1
inch) screen. The crushed ore is then wet ground in a
rod or ball mill and is transferred as a slurry to leach-
ing tanks.
b. The ore is contacted with sulfuric acid solution and an
oxidizing reagent to leach uranium from the ore. The
product liquor is pumped to the solvent-extraction cir-
cuit while the washed residues (tailings) are sent to
the tailings pond or pile.
c. Solvent extraction is used to purify and concentrate
the uranium.
d. The uranium is precipitated with ammonia and transfer-
red as a slurry.
e. Thickening and centifuging are used to separate the
uranium concentrate from residual liquids.
f. The concentrate is calcined and pulverized.
g. The concentrate or yellowcake is packaged in 208 liter
(55 gallon) drums for shipment.
Large amounts of solid waste tailings remain following the
removal of the uranium from the ore. A mill may generate 1,800
metric tons per day of tailings solids slurried in 2,500 metric
tons of waste milling solutions. Over the life time of the mill,
about 100 hectares (250 acres) may permanently be committed to
-------
H2S04
Solvent
Ammonia
Ore.
Crushing
Screening
Grinding
(a)
1
*-
Concentrate ^ —
Product
f
Acid Solvent
Leaching • •• ^- Extraction
(b) (c)
Tailings to >^ Raffinate
Retention Pond >v
Scrub
s J
ber
> ^w
Packaging Calcining _^
Pulverizing
(g> (f)
A
i
Precip
• ^^ nn
Wash
(d
Stack to
Atmosphere
t
Underflow
X
11
Cent
r
itation
d
ing
)
-
ilckening
and
:rifuging
(e)
Heat
T
Waste Liquor
to Retention Pond
N5
OJ
Figure 2-1 Uranium mill process diagram
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24
store this material. These "tailings piles" will have a rad-
iological impact on the environment through the air pathway by
continuous discharge of radon-222 gas (a daughter of radium-
226), through gamma rays given off by radon-222 and daughters
which undergo radioactive decay, and finally through air and
water pathways as radium-226 and thorium-230 are blown off the
pile by wind and leached from the pile into surface waters.
The radiological impact of these piles requires special
considerations; therefore, they will be treated separately in
section 2.7.
2.2 The Model Mill
A system of model plants has been assumed for each segment
of the nuclear fuel cycle in order to achieve a common base for
the comparison of radiation doses, committed health effects, and
radioactive effluent control technology.
The model mill is defined in terms of contribution to the
nuclear fuel cycle that is consistent with current and projected
commercial industry practice (1). Characteristics of the model
mill are assumed to be:
1. 600,000 MT ore milled per year,
2. 1,140 MT U-0- as yellowcake produced per year,
3. use of the acid leach process,
4. tailings retention pond system which uses a clay core
earth dam and local topographic features of the area to
form the impoundment,
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25
5. seepage through the dam into a river with a flow rate
of 14 m /s,
6. production to supply the requirements of 5.3 LWR power
plants on line, i.e., 5.3 GW(e) power/yr, and
7. location in a western state in an arid, low-populated
density region.
Radiation dose rates and health effects that might result
from the discharges of radioactive effluents from the model mill
were calculated using standard (x/Q) values, dose conversion
factors, model pathways, and health effect conversion factors
that are similar to those for other facilities in this discus-
sion of the fuel supply cycle. The factors and assumptions are
discussed in appendix A.
The operating lifetime of a uranium mill is commonly froia 12
to 15 years, depending upon the local ore supply and the demand
for uranium. In a few instances, the operating lifetime may be
longer and allowances are sometimes made for that possibility if
it appears feasible. For this model mill, an operating lifetime
of 20 years has been selected. Discussion of health effects from
30 years of plant operation is only for convenience in comparison
with other operations in the fuel supply system.
2.3 Releases of Radioactive Effluent from Uranium Mills
The radioactivity associated with uranium mill effluents
comes from the natural uranium and its daughter products present
in the ore. During the milling process, the bulk of the natural
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26
uranium is separated and concentrated while most of the radio-
active daughter products of uranium remain in the uranium-
depleted residues that are pumped to the tailings retention
system. Liquid and solid wastes from the milling operation will
contain low level concentrations of these radioactive materials.
Airborne radioactive releases include radon gas and particles of
the ore and the product uranium oxide. External radiation levels
associated with uranium milling activities are low, rarely exceed-
ing a few mrem/h even at surfaces of process vessels.
2.3.1 Airborne Releases
The radiological releases from uranium milling operations
include airborne particulates and vapor. Dusts containing urani-
um and uranium daughter products (thorium-230 and radiun-226)
are released from ore piles, the tailing retention system, and
the ore crushing and grinding ventilation system. Natural ura-
nium is released from the yellowcake drying and packaging oper-
ations as entrained solids.
Radon gas is released from the leach tank vents, ore piles,
tailings retention system and the ore crushing and grinding ven-
tilation system. There is no practical method presently avail-
able to prevent the release of radon gas from uranium mills.
Table 2-1 gives the estimated maximum release rates and
conservative estimates of site boundary concentrations consid-
ering all potential sources of airborne dust fumes and mists as
predicted for the Highland Uranium Mill in Wyoming (.2,3). The
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Table 2-1
Predicted airborne releases of radioactive materials from the Highland Uranium Mill
Release rate Site boundary A a Site boundary B
Radionuclide (Ci/yr) Air concentration Air concentration
(pCi/m3) (pCi/rrr1)
Uranium-natural
Thorium-230
0.1
.06
0.003
.001
0.0004
.0001
(insoluble)
Radium-226
(insoluble) .06 .001 .0001
Distance to site boundary A assumed to be 800 m (2,600 ft) west of mill.
Distance to site boundary B assumed to be 5,200 m (12,700 ft) east of mill.
ro
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28
capacity of the Highland Mill is about 1,200 MT/yr of yellow-
cake.
2.3.2 Waterborne Releases
The following discussion refers to the best of current pro-
cedures of handling mill liquid wastes where these wastes plus
tailings are stored in a tailings retention pond system which
uses an impervious clay cored earth dam and local topographic
features of the area to form the impoundment.
The liquid effluent from a mill (acid-leach process) con-
sists of waste solutions, from the leaching, grinding, extraction,
and washing circuits of the mill. The solutions, which have an
initial pH of 1.5 to 2, contain the unreacted portion of the sul-
fur ic acid used as the leaching agent in the mill process, and
sulfates and some silica as the primary dissolved solids with
trace quantities of soluble metals and organic solvents. This
liquid is discharged with the solids into the tailings pond.
Concentrations of radioactive materials predicted in the
2,500 MT/day of waste liquor from the Highland milling plant are
shown in table 2-2. Radioactive products of radon decay may also
be present in small concentrations. Since the concentrations of
radium-226 and thorium-230 are about an order of magnitude above
the specified limits in 10 CFR 20, considerable effort must be
exerted to prevent any releases of this material from the site.
The waste liquor is, therefore, stored in the tailings retention
pond which is constructed to prevent discharge into the surface
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29
Table 2-2
Concentrations of radioactive effluents in
waste liquor from a uranium mill
Concentration
Radionuclide (pCi/1)
Uranium-natural 800a
Radium-226 350
Thorium-230 22,000
aAbout 0.001 g/ml.
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30
water system and to minimize percolation into the ground. This
is a continuing potential problem requiring monitoring programs
to insure that there is no significant movement of contaminated
liquids into the environment.
If an earth-fill, clay-cored dam retention system serves as
a collection and storage system for the liquid and solid process
wastes generated in the mill, it will permit the evaporation of
most of the contained waste liquids and serve as a permanent
receptacle for the residual solid tailings. However, after the
initial construction of the retention system, it is to be expec-
ted that there will be some seepage of radionuclides through and
around the dam (2_,_3_). It has been estiroated that this seepage
will diminish over a period of about 2 years because of the
sealing effect from accumulation of finer particles between the
sandstone grains. On the other hand, sealing may not occur.
Examples of the total quantities of radionuclides that are esti-
mated to be released under such conditions are shown in table
2-3. Radium-226 is a radionuclide of concern in this case.
Radium-226 levels as high as 32 pCi/1 (4_) have been found in
seepage from current operating mills. Assuming a seepage rate
of 300 liters per minute (80 gpm), the concentration of radium-
226 seeping into a stream of 140 liters per second (5 cubic feet
per second) is approximately 1 pCi/1 which is 33% of the current
drinking water standards. In the applicants environmental report
for the Highland Uranium Mill, a concentration of 350 pCi/1 was
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31
Table 2-3
Estimates of quantities of radionuclides seeping through the
impoundment dam of a uranium mill initially and at 2 1/4 years
Radionuclide
Uranium
Thorium-230
Radium-226
Initial seepage
per day
350 pCi
9,600 pCi
150 yCi
Seepage per day
after 21/4 years
35 yCi to 3.5 yd'
960 yCi to 96 pCi
15 yCi to 1.5 yCi
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32
assumed bringing the concentration of radium-226 in such a
stream up to 12 pCi/1. The Highland Uranium Mill is also
estimated to release to the tailings pond 22,000 pCi/1 thori-
um-230 and trace quantities of short-lived radon daughter
products.
As an additional example, the analysis of plant tailings
effluents for the Humeca Uranium Mill is given in table 2-4 ( j>).
The radiological significance of seepage from tailings ponds
will depend on the location of the pond. In arid regions, the
seepage may evaporate before leaving the site, leaving the radio-
activity entrained and absorbed on soil. Should the tailings
pond be located near a river, minor leakage might be diluted suf-
ficiently by the additional river water to meet relevant drink-
ing water standards. Discharge of pond seepage into streams pro-
viding insufficient dilution and not under the control of the
licensee would not be acceptable. In such a case, a secondary
dam may be built below the primary dam to catch the seepage
which may then be pumped back into the tailings ponds.
2.3.3 Radioactive Effluents from a Model Uranium Mill
Because regulations have not required uranium mills to report
the total amounts of each radionuclide discharged per year, the
source terms chosen for the model mill are based on limited infor-
mation (,2_»J3,J5). Source terms listed in table 2-5 for the model
mill are believed to be reasonably accurate estimates of the quan-
tities of radioactive materials discharged to air and water path-
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33
Table 2-4
Analysis of plant tailings effluents
from the Humeca Uranium Mill
(alkaline leach process)
Radionuclide pCi/1
Radium-226 10 to 2,000
Thorium-230 0.1
Uranium-238 4,000
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34
Table 2-5
Discharge of radionucTides to the environment from a model
uranium mill (acid leach process)
Radionuclide
Uranium
Radium-226
Thorium-230
Uranium
Thorium-230
Radium-226
Pathway
Air
Air
Air
Water
Water
Water
Possible
chemical
states
oxides
a
a
uo2++
Th++
Ra++
Source term
(Ci/yr)
0.1
0.06
0.06
0.1
3.5
0.06
Not known.
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35
ways by real facilities but experimental confirmation of these
values is not yet available. Because uranium is discharged to
the air pathway as ore dust and as calcinated yellowcake, it will
be considered as an insoluble aerosol. Radium-226 and thorium-
230 discharged as ore dust will also be considered insoluble
aerosols. Seepage of process liquors from tailings ponds will
be assumed to be discharged directly into a river (appendix A).
2.4 Radiological Impact of a Model Mill
Estimates of the radiation doses received by individuals in
the vicinity of a model mill from routine effluents are given in
tables 2-6 and 2-7, for doses through the air and water pathways,
respectively. The estimated aggregate doses to the population
in the vicinity of a mill are given in table 2-8. The models
for the dispersion and dose calculations are discussed in appen-
dix A.
2.5 Health Effects Impact of a Model Mill
The expected cost in health effects to members of the general
population in the vicinity of a model mill are presented in table
2-9. The models used for the calculation of health effects are
given in appendix A.
2.6 Control Technology
2.6.1 Airborne Effluent Control Technology
Hazardous airborne gaseous and particulate wastes are genera-
ted in the milling operation from a number of different sources.
The major areas of the milling operations in which gaseous and
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36
Table 2-6
Radiation doses to individuals in the general population
in the vicinity of a model mill, from inhalation
Source term
(Ci/yr)
0.1
(uranium)
0.06
(radium-226)
0.06
(thorium-230)
Critical
organ
Lung
Lung
Lung
Maximum dose to
Individual at plant
boundary
(mrem/yr per facility-yr)
190
130
130
Total 450
critical organ3*
Individuals within
(mrem/yr per facili
0.04
0.03
0.03
0.10
80 ki
ty-y
Each facility supports 5.3 on-line 1 GW(e) power plants.
3Listed mrem/yr radiation dose will result from each year of facility
operations.
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37
Table 2-7
Radiation doses to individuals in the general population
in the vicinity of a model mill, from drinking water
Source term
(Ci/yr)
Critical
organ
Maximum dose to critical organ**
Individual at plant
boundary
(mrem/yr per facility-yr)
Individual within 300 km
(mrem/yr per facility-yr)
0.1 Bone
(uranium) $oft t1ssuec
0.06
(radium-226)
3.5
(thorium-230)
Bone
Soft tissue0
Bone
2
0.2
2
0.06
9
Total dose — bone 13
Total dose -soft tissue0 0.3
0.2
0.02
0.2
0.006
0.9
1.3
0.03
aEach facility suooorts 5.3 on-line 1 GW(e) power plants.
Listed mrem/yr radiation dose will result from each year of facility
operations.
cAverage radiation dose to all organs except bone.
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38
Table 2-8
Aggregate dose to the general population
in the vicinity of a model mill
Source
(Ci/yr)
0.1
(uranium)
0.06
(radium-226)
0.06
(thorium- 230)
0.1
(uranium)
0.06
(radium-226)
3.5
(thorium- 230)
Pathway
Air
Air
Air
Total dose
Water
Water
Water
Total dose
Tribal Hnco
Critical
organ
Lung
Lung
Lung
Bone
Soft tissue
Bone
Soft tissue
Bone
Soft tissue
Aggregate dose to population
(dose to critical organs)
(rem/yr per facility-yr)
2
2
2
c.
10
1.0
8.1
0.3
39
K7
— 1
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39
Table 2-9
Committed health effects to the general population
in the vicinity of a model mill
(acid leach process)
Pathway
Air
Water
Critical Mortal ityb
organ (H.E./facility-yr)
Lung 0.0003
Bone 0.0015
Soft tissue 0.0002
Nonfatal effects'5
(H.E./facility-yr)
0
0.0009
0.0002
Genetic effects
(H.E./facility-yr)
0
0
0.0002
Totals 0.002
0.001
0.0002
Total health effects for 30 years of plant operations is 0.1 effects/facility-30 yr.
aEach plant will support 5.3 on-line 1 GW(e) power reactors.
^Listed health effects will result from each year of facility operations.
H.E. - health effect
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40
particulate effluents must be controlled are the ore crushing
area, the fine ore bins, and the yellowcake packaging and drying
area. Current practice involves the use of wet dust control
systems for the ore crushing area and fine ore storage bins and
wet scrubbers with bag filters for the yellowcake packaging and
drying areas. A 95% efficient wet dust control system for the
crusher costs approximately $14,000. Including the cost of instal-
lation, the total cost could range from $30,000 to $50,000. The
fine ore bin can be controlled with a similar device except that
it is smaller because the air flow rate from the bins is less.
The purchase price of such a wet collecting device is $3,000 to
$7,000. The total cost of the unit including installation would
range from $6,000 to $11,000.
Control of effluents produced in the yellowcake packaging
and drying system is currently achieved by using wet scrubbers
or wet scrubber-bag filter units. A wet scrubber-bag filter
unit currently in use in a modern mill is 99% efficient and can
handle a flow rate of 1 m3/s (2,250 cfm) of effluent gases (2).
The purchase price of such a unit is approximately $4,000 to
which an installation cost of from $8,000 to $12,000 is added.
The total cost is about $15,000.
Older mills use only wet scrubbers for control of off-gases
from the yellowcake drying furnace and a bag filter at the yel-
lowcake drumming station. The scrubbers used on furnace off-
gases cost from $8,000 to $12,000 including installation and
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41
have an efficiency of 98.5%. Bag filters currently in use range
from $5,000 up, including installation. A bag filter effectively
controls 99.8% of the fugitive yellowcake. The total cost for
scrubbing and filtering of the yellowcake system ranges from
$13,000 to $17,000 and up and gives an effective system control
of about 99%. Dust collection and control efficiencies can vary
depending upon the type of equipment and the power input. By
increasing the power applied to a given control device, the pres-
sure drop across the collector increases resulting in an increased
efficiency of particulate removal. The increase in power use
results in an increased cost of operation and may also result in
increased maintenance costs. Assuming that a given unit is de-
signed for such efficiencies, the capital cost for a control
unit would be similar whether the unit is to be operated at 90%
efficiency or 99%, but operating costs would vary depending upon
the cost of power, the desired efficiency and the frequency of
maintenance. Additional control efficiencies and costs can be
calculated for the model mill and these are found in table 2-10.
Other sources of gas and dust which can be controlled are
the open pit mine haul roads and the ore storage and blending
piles. In some instances the liquid content of the ore as mined
has been said to be sufficiently high to eliminate most dust
formation in the ore storage and blending area; due to insuf-
ficient information, this case will not be considered at present
beyond stating that the problem appears significant, and it can
-------
42
Table 2-10
Cost and reduction of effluents for control technology for mills
Capital
cost
Control Method (1970 dollars)
Annual
operating
cost
(1970 dollars)
Effluent
percent
reduction
A. Liquids and solids
1
2.
3.
4.
5.
6.
7.
8.
Diked solids retention
(total liquid discharge)
( 1.) + neutralization
( 2.) + Bad 2 treatment
Clay core dam retention
system
4.)
5.
a.
b.
(4.)
+ neutralization
+ Bad 2 treatment
+ seepage return
wel1s & pump
basin & pump
+ pond lining
Gaseous
(Crusher-fine ore bins)
1 . 95% control
2. 99% control
Gaseous
(Yellowcake packaging
& drying)
1. 98.5% control
(drying only)
2. 99.8% control
(packaging)
3. 99.30% control
Gaseous
(Tailing stabilization)13
1. No stabilization
2. 0.6 meters (2 feet) of cover
3. 6 meters (20 feet) of cover
800,000
1,800,000
9,400
8,000
3,200,000
50,000
50,000
10,000
5,000
48,000
0
250,000C
2,400,000°
400,000
510,000
90,000
490,000
600,000
3,500
3,900
350
280
1,000
94.30
99.50
99.90
99.20
99.92
99.99
100.00
100.00
100.00
95.00
99.00
99.00
0.00
5.0-10.0
90.00
-------
43
Table 2-10 (continued)
Cost and reduction of effluents for control technology for mills
This figure is based on a total amount of radium discharged per day that is
available for dissolution to the water. This assumes 3% of initial radium
in the ore will be dissolved in milling operation. The dissolved 3% gives
a starting figure from which reductions can be made.
Stabilization of the tailings pile is being considered from the standpoint
of radon reduction. A major advantage of stabilization which is not being
considered is the elimination of wind erosion and the decreasing of water
erosion from the tailings pile.
GThe figures listed are for stabilization of a depression-fill tailings pile
and would have to be increased for a surface pile because of additional costs
for contouring for side slope reduction.
-------
44
be controlled in principle by sprinkling. Dust generation on
the ore haul road can also be controlled by sprinkling. The
model mill is assumed to utilize two sprinkling trucks at a cost
of approximately $15,000 per item or a cost of $1,500 per year
plus maintenance.
2.6.2 Waterborne Effluent Control Technology and Solid Waste Control
Technology
New mills in the Rocky Mountain area are using impoundment
technology to try to reach zero liquid discharge levels. Recent
practice for treatment of solid and liquid wastes is to select a
natural ravine which has three basic qualifications for waste
storage: (1) limited runoff, (2) dammable downstream openings,
and (3) an underlying impermeable geologic formation. Diversion
systems (dams and canals) are used to limit the runoff area
emptying into the storage basin to prevent flooding of the ravine
during a 50-100 year maximum rainfall occurrence. The tailings
dam, which should be clay cored, is keyed into the underlying
impermeable formation, which, in one example, is a low porosity
shale. Tailings solids slurried in waste process liquids are
pumped to the impoundment reservoir for storage and liquid reduc-
tion. Liquid reduction is accomplished primarily by evaporation,
but also by seepage through the dam, the reservoir walls and
floor. By filling a dammed natural depression with tailings a
relatively flat, stable contour is achieved.
Impoundment of solids is being accomplished in older mills
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45
merely by construction of a dike with natural materials and filling
the diked area with slurried tailings. When full, the height of
the dike is increased with dried tailings to accommodate even
more waste material. Process liquids which overflow the tailings
dike or seep through the dike are sometimes routed through a treat-
ment system and discharged to the environment. The diking proce-
dure, which is less costly initially, creates an above-ground
pile of tailings which is difficult and costly to stabilize.
Stabilization of a tailings pile involves grading the tailings
area to lessen side slopes, establishing drainage diversion, cover-
ing with nonradioactive material, and revegetating the area. In
semiarid regions it may be necessary to irrigate the pile to achieve
initial vegetation growth. Other types of stabilization may also
be feasible. One method involves the covering of the tailings
with large aggregate gravel from a river bottom. Silt fines
which accompany the river gravel will blow away in a short time
leaving what is effectively a wind proof rip rap, thus signifi-
cantly reducing or eliminating migration of the tailings outside
the controlled area. Assuming costs of 38c per cubic meter (29
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46
of difficulties faced in grading, covering, and revegetating the
potentially steep side slopes (3). The maintenance associated
with perpetual care of a stabilized dike system would also be
higher than that for the depression fill system if there is col-
lapse of side slopes and possibly inadequate drainage of preci-
pition from the pile. A rough calculation, performed by the Oak
Ridge National Laboratory, indicates that covering a tailings
pile with 6 meters (20 feet) of dirt will reduce the radon emana-
tions by about 90%. A 6 meter cover and seeding would cost approx-
imately $23,500 per hectare ($9,500 per acre) or $2,375,000 for
a 100 hectare pile. Two waste control options are thus available
for liquid and solid waste control from a mill. Either the li-
quids are treated to remove all but insignificant amounts of rad-
ioactivity before being discharged into a river or the liquids
can be totally impounded in ponds so that further treatment is
not necessary. Solids must be impounded in both procedures.
A mill operator can route the tailings, which are slurried
in process liquids, into a settling pond. Following settling of
the solids, the process liquids will overflow into a collection
basin below the settling pond. Waste liquids then can be treated
with lime to neutralize the acid picked up in the leaching opera-
tion. (Process liquids from an alkaline leach mill can be neu-
tralized with FeSO,-H.O). Assuming an average dosage of 136 kg
(300 pounds) of sulfuric acid to leach each metric ton of ore
milled, and a lime cost of 0.44$ per kilogram of sulfuric acid
-------
47
applied, the cost of neutralization chemicals is calculated to
be $360,000 per year for the model mill (.T.,.8). (A dosage of 18
kg of sulfuric acid/metric ton of ore would have -an associate
neutralization cost of $48,000 per year). The total costs of
neutralization include purchasing, hauling, and applying the lime
to the process liquids is estimated to cost around $400,000 per
year. Neutralization will effectively decrease the thorium con-
centration in the process liquids by 100% and the radium by 90%
since both are insoluble in neutral or alkaline carbonate solutions.
If additional reduction of dissolved radium levels in the
process liquids from the mill is desired the waste can be further
treated with barium chloride (BaClp. This treatment will cause
the radium to be precipitated as barium-radium sulfate. Assuming
an average liquid waste production of 1,250 liters per metric
ton of ore processed, a barium chloride dosage of 300 mg/1 of
waste, and a cost of $23 per 50 kg, the yearly cost of barium
chloride would be $104,000. The total cost of barium treatment
would be around $110,000 or effectively equal to the purchase
price of the barium chloride since hauling and application equip-
ment cost is estimated to be only a few thousand dollars. This
treatment removes 90% of the dissolved radium in the liquids sb
that when coupled with neutralization, a 99% radium reduction is
achieved. If the concentrations of radioactivity are small enough,
the process liquids are then discharged to a nearby river.
A second option involves the construction of a complete tail-
-------
48
ings retention system. This system with its clay core dam and
impervious underlying geological formation is considered to be
effectively a 100% liquid holdup system, although seepage can be
expected. The retention system, however, alleviates the need
for neutralization and barium chloride treatment except to the
extent that the radioactivity concentrations of the seepage would
be lessened if this treatment were carried out. Assuming that
the clay-cored retention dam was 460 meters long, 30 meters high,
and 6 meters at the crest, has face slopes of 2 to 1 and 2.5 to
1, and a cost of about $2.00 per cubic meter to construct, the
total construction cost would be approximately $1,750,000. This
control would eliminate all effluents and therefore all exposures
to uncontrolled areas with the exception of those exposures due
to radon-222 and its daughter products.
Two methods for seepage collection and return are being con-
sidered for new mills. Seepage has been estimated to occur from
a clay-core retention dam at a rate of 300 liters per minute.
In that situation where an impermeable geological formation
underlies the retention system, seepage can be collected in a
catch basin located at the foot of the dam. The collected seepage
can be pumped back into the retention pond thus eliminating re-
lease to the offsite environment. Assuming $0.38 per cubic meter
of earth moved, the collection basin would have an approximate
Personal communication, Richard Bock, U.S. Bureau of Recla-
mation, Department of the Interior, Washington, B.C. (April 1973).
-------
49
cost of $800 (2). This cost would vary depending upon whether
it was necessary to line the basin. If necessary, lining could
be either clay or synthetic in nature. Synthetic lining, at an
approximate cost of $3.20 per square meter including instalation
would be $32,000 per hectare.(about $13,000 per acre). Total
cost with lining for a 460 mx6mxl.5m catch basin would be
about $6,000. In that situation where either an underlying
impermeable geological formation is not existent or is not con-
tinuous, vertical seepage may occur to the underlying ground
water formation. Wells may be drilled downstream of the reten-
tion system into the subsurface formations where seepage would
collect and this water is pumped back to the retention system.
Such a system requires specific favorable subsurface conditions.
If four wells were drilled to an average depth of 15 m (50 feet),
the cost per well would be about $500 or $2,000 total for
drilling.
Pumps for the collection systems would be deployed in the
basin and in each well. Two pumps of about 200 liter per minute
capacity would be utilized in the basin and one pump of about
100 liter per minute capacity could be utilized in each well.
Pump costs are approximately $1,700 for a 200 liter per minute
(50 gpm) unit and $1,500 for 100 liter per minute (25 gpm) units.
It Is necessary that the pump be able to withstand acid corrosion
and, therefore, must be made of stainless steel.
The total cost of seepage collection would then be $9,400
-------
50
for the basin system or $8,000 for the well system or $470 a year
and $400 a year, respectively.
Another method to be considered to eliminate all liquid efflu-
ents would involve the complete lining of the tailings retention
system. For the model mill, the area to be utilized for the tail-
ings retention system is 100 hectares (250 acres). Assuming the
lining of the total surface area at a cost of $3.20 per square
meter ($0.30 per square foot) including installation, the total
expenditure for lining would be $3.2 million (JJ). Exercise of
this option would completely alleviate the need for neutraliza-
tion, barium chloride treatment or any other liquid waste treat-
ment, and would result in total containment of the liquid with
the exception of evaporation. The dose attributable to process
liquids received by a person in the uncontrolled area would be
zero.
2.6.3 Effluent Controls for the Model Mill
Current effluent control systems for the model mill were
assumed to be:
1. ore crushing area and fine ore storage - wet dust
control system
(95% control)
2. yellow cake packaging and drying areas - wet scrubbers
and bag filters
(99.3% control)
3. liquid and solid wastes - clay core
dam retention
system
(99.2% control)
-------
51
The additional add-on controls for airborne releases were
assumed to be bag filters rated at 95% effectiveness and HEPA
filters rated at 99% effectiveness. Liquid releases seeping
through the dam were assumed collected in a basin and returned
to the tailings pond by pumps.
The radiological impact of radioactive effluents versus
controls for the model mill is shown in tables 2-11 and 2-12.
2.7 Uranium Mill Tailings Piles
2.7.1 Introduction
Large scale milling of uranium ore began in the United
States in the late 1940's and will continue indefinitely. The
average time of operation of a uranium mill is about 12 to 15
years. As a result there are currently as many inactive mills
as active mills. When the uranium is extracted from the ore,
more than 99 percent of the ore material becomes the mill wastes
or tailings, a slurry of sandlike material in waste solutions.
The tailings are pumped to a nearby location where the solids
settle out and soon accumulate to form a tailings pile. Each
location where a mill is operating or has operated has an accumu-
lation of tailings. As of 1970, there were more than 80 million
metric tons of tailings occupying more than 850 hectares (2,100
acres) of land.
More than 97 percent of the radioactive decay products of
uranium and about 4 percent of the uranium from the ore remain
in these tailings. The concentration of radium-226 in the
-------
Table 2-11
Radiological impact of airborne effluents vs controls for a model uranium mill
Controls
None
Wet dust collector on
yellowcake packaging
and drying
Wet dust collector on
crusher and ore binsa
Additional HEPA System
on yellowcake packaging
and drying area
Additional bag filter
on crusher and ore bins
Source
term
(Ci/yr)
>15
1.6
0.2
0.07
0.004
Max. dose to the
individual
(mrem/yr, lung)
>30,000
3,600
450
150
10
Total health effects
30 yr plant operations
(H.E./facility-30 yr)
>1
8 x 10'2
f\
1 x 10"^
3 x 10*3
2 x 10~4
Capital
cost
(1970 $/facJ
0
15,000
100,000
1,400
100,000
Annual
operating
cost
(1970 $/fac)
0
630
7,400
600
7,400
Present worth
(1970 $/fac.)
0
52,000
280,000
10,000
280,000
Current levels of controls.
H.E. - health effects
-------
Table 2-12
Radiological impact of waterborne effluents vs controls for a model uranium mill
Controls
None
Clay core dama
retention system
Seepage return sys-
tem with clay core
damb
Lined clay core dam
retention system
Source
term
(Ci/yr)
>400
4
0
0
Max. dose to the
individual
(mretn/yr, bone)
>1,300
13
0
0
Total health effects
30 plant operations
(H.E./facilitv-30 vr)
"9
0.09
0
0
Capital
cost
(1970 $/fac.)
0
1,800,000
1,809,000
5,100,000
Annual
operating
cost
(1970 $/fac.)
0
90,000
90,000
90,000
Present worth
(1970 $/fac.)
0
4,000,000
4,600,000
11,000,000
aCurrent level of control if seepage is to a river.
Current level of control if seepage would be to offsite stream.
H.E. - health effects
Ul
UJ
-------
54
tailings averages about 700 pCi/g, indicating an inventory of
about 56,000 Ci of radium-226 in these piles. The radon-222
decay product of radium-226 emanates from these piles at an
2
average rate of about 600 pCi/m -s, representing a total release
of radon gas of more than 150,000 Ci/yr. The tailings piles
release radioactive material to the air as radon gas, as air-
borne particulates, and as waterborne radionuclides leached out
by precipitation, surface runoff, and the waste solutions. Suf-
ficient radioactivity is in the tailings to create a weak field
of gamma radiation in the immediate vicinity of the tailings.
Because of the presence in the tailings of thorium-230, which by
its decay maintains the radium inventory, the radioactivity in
the tailings will remain almost constant for thousands of years.
2.7.2 Source Term
Uranium is widely distributed in nature, but generally in
concentrations too small for economic recovery. Although ura-
nium is found in a number of different types of rocks, the ores
most commonly mined in the United States are ground water depo-
sited uranium minerals in sandstone. The uranium is frequently
accompanied by vanadium in these ores. The only thorium commonly
found in these ores is that resulting from the radioactive decay
of uranium.
Most of the uranium ores being milled in this country are
sandstones consisting of silica grains poorly cemented together
with materials such as calcium carbonate and containing some
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55
clay minerals. In the milling process, the sandstone is broken
down and after the uranium values (and sometimes vanadium) are
extracted, essentially the whole original mass remains as tails.
The tails consist of the original sand grains plus slimes made
up of the clay minerals and other materials from between the sand-
stone grains. The slimes may constitute 15 to 60 percent by
weight of the tailings solids; commonly about one quarter. The
slimes may have concentrations of radioactivity 3 to 20 times
those in the sands, and commonly have about three quarters of
the total radioactivity in the tailings.
The radioactivity in uranium ores is not uniform from ore to
ore. It depends, of course, on the amount of uranium in the ore.
It also depends upon the length of time the uranium has been in
the ore; for example, if the uranium has been in place ten mil-
lion years, secular equilibrium may have been reached. This
would be so if no processes of removal or addition, other than
radioactive decay, were changing the quantities of radionuclides
present. Secular equilibrium means, in this case, that one curie
of uranium-238 is accompanied by one curie of each decay product
in its decay chain, i.e., one curie of thorium-230, one curie of
radium-226, etc. If the uranium has been in place only 100,000
years, secular equilibrium will not have been reached. Examples
of removal processes which may upset equilibrium are leaching by
ground water, and diffusion of the radon gas away from the ura-
nium,for instance, in highly permeable sandstones. In the
-------
56
majority of the uranium ores being mined in the United States,
the distribution of radioactivity approximates the condition of
secular equilibrium. The principal decay chain is that begin-
ning with uranium-238 and ending with lead-206. The other decay
chain, beginning with uranium-235 and ending with lead-207, con-
tributes about 5 percent of the total radioactivity.
Most of the ores being mined are 0.1% or more uranium oxide.
The ores are usually stockpiled by the mill and blended to pro-
vide a uniform mill feed of about 0.2% uranium oxide. The radio-
activity concentrations are those associated with the 0.2% ura-
nium. Twenty to eighty percent of the radon gas in the ore and
small fractions of the other radionuclides are released to the
environment during milling. Small fractions of some radionuclides
are also shipped out in the yellowcake and vanadium pentoxide
mill products. The radium-226 in the yellowcake varies from
0.001% to 0.2% of that passing through the mill for the acid-
leach process, and from 1.5% to 2.2% for the Alkaline-leach pro-
cess. The radium in the vanadium pentoxide may be 1% of that
passing through the mill. It is likely that similar fractions
of the other radionuclides are also in the mill product. The
problem of radium contamination and the radon it produces is, of
course, present in other industries, e.g., the vanadium processing
industry. In any case, almost all the radioactive decay products
of the uranium in the ore find their way to the tailings, mostly
in the tailings solids, but with small percentages in solution.
-------
,v<
57
Ore that is 0.2% by weight of uranium oxide (U00_) contains
J o
0.00056 grams radium-226 per metric ton of ore if secular equili-
brium of the radionuclides is assumed. Assuming that essentially
all the radium remains with the tailings there will be 560 pCi
radium-226 per gram of tailings. Reported values range from 100
to 1,000 pCi/g with values from 700 to 800 common. The higher
values probably represent tailings from richer ores. Culot, et
al. (10) have measured the sand fraction of tailings to deter-
mine that portion of radium which releases radon to the atmosphere.
The tailings sand they used contained 125 pCi radium per gram.
They determined that 23% of the radium in the tailings sand could
release radon to the atmosphere. This will be assumed to hold
for the slimes also, although Pearson (11) indicates that the
smaller particles may release a greater fraction.
The net density of dry tailings is about 1.6 g/ml, implying
a possible release rate of 430 pCi/m3-s of tailings. However,
these piles may be over 10 m thick and average over 5 m. Not
all radon produced deep in the pile will escape to the surface
before undergoing radioactive decay. A measure of the effect of
the depth is the "relaxation length," the depth from which the
fraction (1-1/e) or 63% of the radon escapes. For an average
relaxation length of 1.15 m for sand (10), the release rate of
radon is calculated to be about 500 PCi/m2-s from the surface of
226
a pile containing 560 pCi Ra/g. Approximately 3 meters depth
of tailings are required to produce close to the maximum radon
-------
58
flux.
While mills are in operation, most of the tailings may be
saturated with water or under water in tailings ponds. The
water will tend to prevent the escape of the radon gas. There
do not seem to be any measurements of radon releases from satur-
ated tailings or tailings ponds, but theoretical calculations
(3) indicate that saturating a dry tailings pile with water will
reduce its radon emissions by a factor of about 25. This indi-
2
cates a release rate of about 20 pCi/m -s for a radium concen-
tration of 560 pCi/g. For comparison, the natural background
2
release of radon from the ground is about 1 pCi/m -s in most of
the country; it may be 100 times greater over uranium deposits
(11).
c o
Applying these values to 5 x 10 m of tailings (about 125
acres) containing 560 pCi 226Ra/g gives for a years release:
1. 7,900 Ci/yr - 5 x 105 m2 - Dry tailings
2. 320 Ci/yr - 5 x 105 m2 - Tailings pond
5 2
3. 16 Ci/yr - 5 x 10 m - Natural
background
At operating mills where the tailings pond and pile are still
receiving tailings, most of the tailings solids will be quite
wet; perhaps two-thirds of the area will be saturated to the
surface or under water and the remaining third may have a depth
of only about a meter that is relatively dry. When the milling
operation is discontinued, the sands will dry quickly. However,
if the slimes have been segregated and not mixed with the sands,
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59
they may retain moisture longer.
The radon release rate at any one location is known to vary
over a factor of 10 due to effects of weather, i.e., wind speed,
barometric pressure, atmospheric stability, rain fall and snox?
cover.
The gamma radiation from tailings is low intensity and the
majority of the photons are low energy. Net average exposure
rates have been measured with thermoluminscent dosimeters on the
face of four piles and ranged from 0.2 to 1.1 mrem/hour (12).
These measurements also indicated that the external gamma radia-
tion dose 50 meters from the tailings piles is at normal back-
ground level.
2.7.3 Radiation Dose and Health Effects to Members of the General
Population in the Vicinity of a Uranium Mill Tailings Pile
A plume from a tailings pile 1,000 meters square (250 acres)
is more diffuse at 1 km from the pile edge than if the same
amount of curies were discharged from a point source at the
center of the pile. The (x/Q) for an area source is therefore
cl
smaller by a factor of about 4 compared to the value for a point
source located at the center of the pile. For a model pile
— —7 3
1,000 meters square, a (x/Q) of 4.5 x 10 s/m was assumed to
ci
calculate air concentrations of radon-222 1 km from the pile
edge. Calculations and pathway assumptions are otherwise the
same as for the model mill; the dose conversion factor for
radon-222 was assumed to be 4 mrem/yr per pCi/m of radon-222
-------
60
(section 2.7.5).
For a model dry pile 1,000 meters square containing 560 pCi
radium-226 per gram of tailings, the maximum exposed individual
living downwind 1 km from the pile edge receives a dose of 900
mrem/yr. Persons living within 80 kri receive 2.2 mrem/yr giving
an aggregate dose of 120 rem/yr to members of the general popula-
_3
tion. The aggregate dose predicts 6 x 10 mortality events/pile-
yr and 0.2 mortality events/pile-30 yr.
Table 2-13 gives the radiological impact of a model uranium
mill tailings pile.
2.7.4 Experimental Measurement of Radon around Tailings Piles
Under a joint agreement between the U.S. Public Health Service
and the U.S. Atomic Energy Commission and in cooperation with
the Colorado Department of Public Health and the Utah State Divi-
sion of Health, a project was begun in 1967 to evaluate the pub-
lic health aspects of radon-222 in the vicinity of certain tail-
ings piles. Data from the study can be used to confirm radiation
exposure predictions from radon emmanating from tailings piles.
Sampling stations for the radon were operated to provide an
estimation of yearly average radon concentrations. These values
include both the natural radon-222 background and the radon-222
from the pile. Stations not directly downwind were assumed to
receive a portion of the radon from the pile.
Table 2-14 gives the results of the experimental measurements
as well as predicted values for the two piles for which the
-------
Table 2-13
Radiological impact of airborne effluents vs controls for a model uranium mill tailings pile
(250 acres) (/orrL*Q pe^-^-^^ }
Control s
None
Tailings ponda
,(2 ft water)
2 ft coverb
20 ft cover
TOO ft cover0
Source
term
(Ci/yr)
20,000
800
15,000
2,000
0
Max. dose to the
individual
(mrem/yr, lung)
1,300
50
980
130
0
Total health effects per
30 yr plant operations
(H.E,/facility-30 yr)
0,25
0.01
0.19
0.025
0
Present worth
(1970 dollars/facility)
0
0
150,000
1,400,000
Unknown
aPresent level of control while operational for recent mills.
^Present level of control for recent mills - post operational.
cTails used as back fill in strip mine; original layer of overburden is replaced. Ground-water effects must
be considered in this option.
H.E. - health effect
-------
Table 2-14
Experimental and predicted values of radon-222 emanating from tailings piles
Predicted
Amount of Radium-226 concentration of
tailings Area concentration radon, 1 km from pile
Pile (tons) (m3) (pCi/g) (pCi/m3)
Grand Junc-a 2 x 10*6 2.2 x 10+5 900 80
tion
Salt Lakeb 1 x 10»6 4.3 x 10*5 1,100 200
City
Measured con- Radiation dose cal-
centration of radon culated from measured
1 km froqj pile concentration
(pCi/rn^) (mrem/yr)
600 3,200
300 1,100
aNatural radiation background from radon-222 - 2,400 mrem/yr.
bNatural radiation background from radon-222 - 600 mrem/yr.
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63
greatest amounts of experimental data were available. For the
Grand Junction pile the difference between predicted and measured
values could be partly explained by the fact that the wind is
blowing lengthwise down a long, narrow pile. The source term is
behaving more as a point source than as a. diffuse source and the
concentration of radon-222 downwind would, therefore, tend to be
higher. Both piles have since been stabilized by the addition
of an earth cover. While such a cover would tend to eliminate
the loss of radium and thorium from the piles by wind erosion
and water leaching, it would not be expected to reduce the amount
of radon-222 emanating by more than 25%.
2.7.5 Radon-222 Dosimetry
Radon-222 dosimetry is a complex and difficult subject.
Radon-222 (and daughters) is one of the few radionuclides known
to have caused significant numbers of fatalities in occupationally
exposed workers (12). It is, therefore, unfortunately true that
a certain limited amount of clinical evidence is available con-
cerning the consequences to humans from exposure to radon. A
biological effect of excess exposure to radon is a form of lung
cancer, considered to be nearly 100% fatal. One concerned with
radon exposure is aware of past failures to provide adequate pro-
tection and has reason to be conservative in any value judge-
ments and assumptions he is forced to make concerning radon
dosimetry. Some of the major elements of lung dosimetry for
radon are described below as they are related to the problem of
-------
64
radon emanations from tailings piles. A general review is given
in reference 14. An estimate of the dose conversion factor for
radon will then be made following a brief discussion of the more
relevant opinions concerning this problem.
Table 2-15 lists the decay scheme and daughter products of
radon-222. Examination of this table indicates that pure radon-
222 of itself is not as hazardous as is radon and its daughters
together. Being a noble gas, radon does not remain in the lungs
and in addition, radon is not in intimate contact with the
tissue. In the event of a disintegration of radon, the alpha
energy is likely to be expended into a noncritical area of the
lung. Radon daughters, in particular polonium-218 and polonium-
214 which are not noble gas elements, have chemical and physical
properties that cause them to be deposited on the mucus layer
covering the bronchial epithelium of the lung leading to high
dose rates to a region of the lung where tumors are most likely
to arise. Radiation dose caused by radon is therefore a function
of the state of equilibrium between radon and its daughters at
the time of exposure, and therefore it is not enough simply to
know the radon-222 concentration alone.
Given the case of radon emanating from tailings piles when a
5 m/s wind is blowing, it is likely that the dose near the pile
will be quite low because emanating radon is free of daughters.
By the time the wind has carried the radon 1 or 2 km (3 to 6
min.) the polonium-218 daughter has grown in and the dose
-------
65
Table 2-15
Radon-222 decay scheme
Radionuclide
222Rn
218PO
214Pb
214Bi
214po
210pb
210B1
210Po
Half life
3 day
3.05 min
26.8 min
19.7 min
1.6 x l(Hs
22 yr
5 day
138 day
Emission
alpha
alpha
beta
beta
beta
beta
beta
alpha
Energy
5.5 MeV
6.00 MeV
.0.6 MeVa
>1 MeVa
7.68 MeV
Low3
1.2 MeVa
5.3 MeV
aMaximum energy of most intense beta
-------
65
delivered to individuals exposed to this air increases sharply.
The same effect Is achieved if the radon-loaded air is held up
or delayed until the radon daughters can grow in. A house may
have as few as two turnovers of air per hour and can delay the
radon in its passage more than 30 minutes. For this reason,
dose measurements made inside a house in the plume of a tailing
pile should show higher exposure levels than similar measure-
ments made outside the house.
When an atom of radon-222 decays in air, it becomes an atom
of polonium-218. This polonium atom exists as an ion, i.e., it
has an electrical charge. This charge makes it highly suscepti-
ble to absorption into an airborne aerosol particle; if inhaled,
the ion is more likely to be captured by the lung than is the
particle to which it might have become attached. The radiation
dose delivered by radon daughters is assumed to be very much a
function of the fraction of these daughters remaining as ions
rather than absorbed into particles. The cleaner the air from
particulate matter, the higher the dose rate will be from radon
daughters because more daughters will be present as ions rather
than absorbed into particles. Therefore, in western states
where tailings piles are located, and where air is presumably
lowest in aerosol concentrations, radiation dose rates delivered
by exposure to a given concentration of radon-222 should be
higher than in more industrial areas. Air conditioning may also
increase doses by removing significant numbers of particles from
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67
the air.
The critical biological target in the lung is assumed to be
the nuclei of the basal cells of the bronchial epithelium. Alpha
particles have a limited range in tissue. To reach the basal
cells they must penetrate first a mucus layer, then parts of cer-
tain other cells. The bronchial epithelium shows extreme vari-
ations in thickness because the wall of the bronchus tends to
fold upon itself. In addition, there is to be expected natural
variation in thickness of both the mucus layer and the inter-
vening cells if a population of all age groups is considered.
It is, therefore, a matter of some controversy exactly to what
depth the alpha particle must penetrate to deliver a critical
exposure.
Given the above difficulties, which are by no means a com-
plete review of the problem, it should be clear why radon dosi-
metry does not give a simple correlation between the radon-222
concentration in air and the radiation dose delivered to the
lung. The several approaches considered in this analysis are
described below.
ICRP Publication No. 2 states that for occupational exposures:
"... Recent studies have indicated that when radon and
its daughters are present in ordinary air the free ions
of RaA constitute only about 10 percent of the total
number of RaA atoms that would be present at equilibrium
and these unattached atoms deliver all but a small frac-
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68
tion of the dose to the bronchi. Based on these measured
dose rates the (MFC) for exposure to radon and daughter
SL
products is found to be 3 x 10 /(I 4- lOOOf) where f is
the fraction of the equilibrium amount of RaA ions which
are unattached to nuclei."
If f is taken to be 0.1 for "ordinary" air and assumed to
range to 0.5 for "clean" air, the concentration of required to
_O -5 _O
give 15 rem to the lung is 3 x 10 yCi/cm to 0.6 x 10 \id/
3
cm ,respectively, for a 40 hour week. These figures must be
divided by 2.92 for continuous exposure (168 h week). The result
o
is that 1 pCi/m radon gives a dose rate of 1.5 to 7.5 mrem/yr
depending on the assumptions concerning the number of free ions
of polonium-218 present. It is notable with this model that in
clean air where the uptake of ions by particles may be slow, that
it would require only 3 minutes time for 50% of equilibrium of
RaA ions to occur thereby yielding the higher dose rate to the
lung.
A recent UNSCEAR report (IS) has reviewed the literature on
radon dosimetry and has recommended using a more complex method
of estimating dose rate (table 2-16) from radon-222 concentra-
tions. Calculations based on values presented in table 2-15 and
using QF of 10 to convert from rads to rems yields for various
penetration depths to cells at risk:
1. For adequate ventilation, 60 urn depth
3
1 pCi/m radon-222 « 3.6 nrem/yr
-------
69
Table 2-16
222
Calculated alpha dose rates in mrad/yr from inhalation of short-lived Rn
daughter products to the basal cell nuclei of segmental bronchi (14)
Depth (urn)
Living accommodation:
ventilation3
Living accommodation:
ventilation15
Industrial premises0
Air-conditioned sites*1
adequate
inadequate
a 222Rn, Z18Po, 214Pb and 21
0.057 pCi/1, respectively.
ppp 91Q 91A 91
b *"Rn, ^'°Po, "*Pb and *'
15
550
,1,490
840
280
30
280
790
445
140
4
Bi concentrations:
Annual dose (6,000
Bi concentrations:
45
100
330
190
50
0.164, 0.
h) in mil
0.37, 0.3
60
40
120
75
15
148, 0.083
lirads.
5, 0.26 anc
70
1.5
5
3
0.6
and
i O.?l
. . .
pCi/1, respectively. Annual dose (6,000 h) in millirads.
c 222Rn, 218Po, 214Pb and 214Bi concentrations: 0.32, 0.31, 0.27, 0.25
pCi/1, respectively. Annual dose (2,000 h) in millirads.
ppp 71ft 91A 214.
d "''Rn, iloPo, iMfPb and tlHBi concentrations: 0.17, 0.15, 0.074, 0.060
pCi/1, respectively. Annual dose (2,000 h) in millirads.
-------
70
2. For inadequate ventilation, 60 inn depth
o
1 pCi/m radon-222 = 11 mrem/yr
3. For inadequate ventilation, 30 \im depth
3
1 pCi/m radon-222»71 mrem/yr
Experimental measurements of radon daughter abundance levels
in living quarters have been made by Yeats, et al. (16) and con-
verted to dose rates through the format of:
1 working-level-month = 7 rads to the bronchial epithelium.
They yielded an overall conversion factor for radon exposure as
experimentally measured in dwellings of:
3
1 pCi/m radon-222 = 35 mrem/yr
The 10 CFR 20 (MFC) occupational limit of 1 x 10~7 yCi/ml,
which is assumed to be equivalent to 15 rem/yr (40 hour week),
yields if calculated for a 168 hour week:
3
1 pCi/m =0.5 mrem/yr
3
A value of 1 pCi/m radon-222 = 4 mrem/yr has been used to
calculate dose rates resulting from exposure to radon-222 and
daughters from tailings piles. This represents an acceptance of
the UNSCEAR recommendations and assumes adequate ventilation and
60 ym penetration depth. It does not represent the worst con-
ceivable case; but because of the uncertainties, it is considered
an acceptable conversion factor to estimate the average dose to
large numbers of members of the general population and for the
setting of generally applicable environmental standards.
"
<*/<> "
-------
71
2.7.6 100-year Dose Commitment from Radon-222 Emanations from Tailings
Piles
The 100-year dose committment is an attempt to estimate the
long term radiation dose and health effects that will result from
radon emanating from the tailings piles of uranium mills. These
health effects are in addition to the effects that occur from
the immediate exposure of individuals within 80 km of the plant
during the 30 year operating lifetime of the plant. Long term
exposure occurs because the tailings from the mill contain long
half-life radium-226, the parent of radon, so that radon will be
continually produced and will emanate from the pile indefinitely.
After release, it was assumed that the radon will distribute
over the eastern United States and into the northern hemisphere
causing health effects. While the dose to any individual is
extremely small, the number of people exposed is large so that
because of the linear, nonthreshold health effects model, the
number of predicted health effects is significant.
The health effects committed as the result,of each year's
plant operations (effluents) exposing members of the general
population to radiation for the following 100 years are calcu-
lated. The committed health effects for each year of plant
operations are then summed over the lifetime of the plant to
give the health effects resulting from the 100-year dose com-
mittment for 30 years of plant operations. The calculations are
-------
72
similar in concept to those given in appendix A, section 7.
Results of the 100-year dose committment calculated for a
uranium mill tailings pile are given in table 2-17.
2.8 Summary and Conclusion
Both theoretical predictions and experimental evidence indi-
cate that individuals in the general population may be receiving
very high levels of radiation exposure to the lung caused by the
release of radon from uranium mill tailings piles.
As examples: For the Grand Junction pile, the value is 3
rem/yr to certain residential locations downwind of the pile,
and for the Salt Lake City pile, the corresponding value is 1
rem/yr.
For mills of recent design where most of the tailings are
expected to be under water for the operating life time of the
plant, radon release rates from the wet tailings are expected to
be about 4% of those from dry tailings. However, when a. plant
ceases operations, currently the pile is allowed to dry. A 20-
foot covering of earth would then be required to reduce radon
emissions by 90% compared to an uncovered dry pile.
The highest radiation dose from the model mills at current
control levels is to the lung (450 mrem/yr) of individuals that
might live within 1 km of the plant. Additional filtration of
the air streams can reduce this value to less than 10 mrem per
year.
-------
73
Table 2-17
Health effects resulting from the 100-year dose commitment
from radon-222 emanation from a uranium mill tailings pile
(250 acres)
Health effects committed
_
Exposed population during plant operation following plant operatior
Eastern United States 1 120
Northern Hemisphere 0.2 80
Health effects from 100-year dose commitment - 200 effects/facility-30 yr
-------
74
The model mill currently discharges 4 curies of activity
(mostly thorium-230) to the water pathway. This discharge, can
be eliminated by a catch basin and pumps.
Immediate health effects committed under current control
levels are predicted to be 0.7 and 0.01 health effects/facili-
ty-30 yr for the mill and the mill tailings pile, respectively,
excluding radon. As many as 200 health effects may result from
the 100-year dose committment due to radon emanations from large
uranium mill tailings piles.
3.0 Conversion Facilities
3.1 General Description of the Uranium Conversion Process(1,2)
Uranium concentrate milled from ore must be converted to the
volatile compound uranium hexafluoride (UF,) in order to be en-
riched by the gaseous diffusion process. Two different indus-
trial processes are used for uranium hexafluoride production.
The "hydrofluor process" consists of reduction, hydrofluorina-
tion and fluorination of the ore concentrates to produce crude
uranium hexafluoride followed by fractional distillation to ob-
tain a pure product. The wet solvent extraction process employs
a wet chemical solvent extraction step at the head end of the
process to prepare high purity uranium feed prior to reduction,
hydrofluorination, and fluorination steps. Each method is used
to produce roughly equal quantities of uranium hexafluoride feed
for the enrichment plants. Illustrative flow sheets are given
in figures 3-1 and 3-2.
-------
VOLATILE IMPURITIES
75
CRACKED AMMONIA
Nil
A . n A
ANHYDROUS HF
ORE J f |
CONCENTRATIONS PRE-PROCESS REDUCTION HYDROFLUOR-
^ nAHULmu »- »- (NATION
LfL I") id
VOLATILE u /
| 'MPURITIES /IMPURITIES
LIQUID WASTES BURN[R /
OFF-GASES H n /
| i2 /
1
HF SCRUBBER
RECOVERY "*
x
r. CRUDE , _ _ X
REFINED UF, FRACTIONAL UF. ho
6 ^ .. „"",..,"„„ --6 FlimRINATIDN Z F
PRODUCT LOADOUT DISTILLATION ^ ^*
I I
WASTES SOLID WASTES V
UF4
LUORINE
PLANT
1
VASTES
Figure 3-1. UF production-hydrofluorprocess block diagram
-------
HNO 3
PRE-PROCESS DISSOLUTION
CONCENTRATIONS "^ 'NG UI(itSIIUN
[a] (b)
DILUTE
FOR REC\
F , HEAT
2'
REFINED UFC
V
HF
fCLE ^
b U^6 FLUORI- UF4
PRODUCT LOADQUT ^ nniiun ^
(h)
F2
TBP&HEXANE
*
(c)
vt
^ RE-EXTR
(d
1 RAFFINATE TO
0 STACK f HOLDING PONDS
A
» !
HE
t "2°
BUR
VOLATILES 1
HF, HEAT *
HYDROFLUOR- uu ;
(NATION "^
(el
t
ANHYDROUS
HF
AT CALH
(<
KER
I
\
ACTION
)
INING ,
*}*
uo3
I REDUCTION
(f)
1
CRACKED
AMMONIA
HEAT
Figure 3-2. UF production-wet solvent extraction-fluorination block diagram
6
-------
77
The two commercial plants (3.-5_) currently in operation
process approximately 10,000 metric tons of uranium into uranium
hexafluoride per year; 180 metric tons of uranium converted to
270 metric tons of uranium hexafluoride are required to support
a GW(e)-yr of electricity generated by light water reactors.
The uranium concentrate feed to a conversion plant contains
the equivalent of about 75% to 85% uranium oxide. The conver-
sion process removes essentially all of the remaining impurities
and produces a highly purified uranium hexafluoride product.
The dry hydrofluor process separates impurities either as vola-
tile compounds or as solid constituents of ash. The wet solvent
extraction method separates impurities by extracting the uranium
into organic solvent leaving the impurities behind dissolved in
an aqueous solution. Therefore, the nature of the radioactive
effluents from the two processes differ substantially; the hydro-
fluor method releases radioactivity primarily in the gaseous and
solid state, while the solvent extraction method releases more
of its radioactive wastes dissolved in liquid effluents.
Both plant designs stipulate virtually complete recovery of
uranium, total utilization of fluorine, and high utilization of
the other main reactants, (hydrogen, hydrogen fluoride, ammonia,
and nitric acid). The plants are located in relatively sparsely
populated areas. The range of population density in the vicinity
of the two existing production facilities is 10-15 people per
square kilometer (25-40 people per square mile); the region sur-
-------
78
rounding the plant using the dry hydrofluor process is the more
densely populated.
3.2 The Model Conversion Facilities
A system of model plants has been assumed for each segment
of the nuclear fuel cycle in order to achieve a common base for
comparison of radiation doses, committed health effects, and also
radioactive effluent control technology.
The model plant is defined in terms of a contribution to the
nuclear fuel cycle that is consistent with current and projected
commercial nuclear industry practice. However, because the ura-
nium hexafluoride used by the commercial reactor industry is pro-
duced in approximately equal quantities by two different methods,
each with its own characteristic amounts of radioactive effluents,
two types of model conversion facilities will be assumed. One
is based on the wet solvent extraction procedure, the other on
the hydrofluor process.
Each type of model plant is assumed to process 5,000 MTU per
year, enough to support 28 GW(e)-years of electricity generated
by light water reactors. Additional resources required by the
model plant are listed in table 3-1. In this table, the require-
ments of each plant type are appropriately averaged and considered
as a single model plant of 5,000 MTU/yr capacity.
Radiation dose rates and health effects that might result
from the discharge of radioactive effluents from these model
facilities were calculated using standard (x/Q) values, dose
-------
79
Table 3-1
Model uranium conversion plant8
Produce ion (annual)
LWR/EQAb
Land-industries
Permanent commitment--land
Water use-LWR/EQA
Air discharge
Surface water discharge
Electrical energy consumption (LWR/EQA)
Natural gas for LWR/EQA
5,000 MT uranium as UFC
0
28
570 hectares
4 hectares
170 million liters
14 million liters
160 million liters
2.1 million kilowatt hours
0.9 million cubic meters
Assumes equal production by each of two current processes.
bLWR/EQA = average annual requirements for model 1 GW(e) light water
reactor
-------
80
conversion factors, model pathways, and health effect conversion
factors that are common to all facilities considered in this exami-
nation of the fuel supply cycle. These assumptions are discussed
in appendix A.
3.3 Release of Radioactive Effluents from Conversion Facilities
Because no irradiated material is handled by conversion facili-
ties, all radionuclides present also occur in nature. They are
radium, thorium, uranium, and their respective decay products.
Some of the decay products are delivered to the facility as impur-
ities in the mill concentrate and others reoccur there by the
continuing radioactive decay of the uranium. Uranium is present
in the majority of the plant processes, appears in liquid efflu-
ents and is essentially the only source of radioactivity in the
gaseous effluents. The radium, thorium, and decay products are
separated from the uranium in the conversion process and thus
appear in the liquid effluents or solid waste associated with
specific purifying procedures.
Uranium may appear in the gaseous effluents in several chem-
ical forms. Possible chemical species are U000, U0_, UF., UF,.
jo / 4 0
(NH,)_U_0_, and UCLF-. In the conversion process employing sol-
vent extraction, uranium is present as uranyl nitrate which may
also appear in gaseous effluents. Thus, the uranium may be re-
leased as both soluble and insoluble aerosols. Measurements at
one facility (5) indicate that about two thirds of the airborne
uranium is in an insoluble form, and about one third is in a
-------
81
soluble form. Because uranium has a low specific activity (0.7
Ci/ MTU), the insoluble uranium is amenable to filtration. The
discharge to the environment is through low stacks and vents.
Liquid effluents are associated with the various solvent ex-
traction and scrubber systems so that the radionuclides in these
effluents are considered to be in solution. About 1.7 metric
tons of uranium, 0.03% of the material processed, appears in the
liquid effluent streams.
Conversion of 10,000 MTU/yr to uranium hexafluoride produces
3
an estimated 1,000 metric tons of solid waste (about 450 m }
that must be shipped to a commercial waste disposal burial site
(2). Materials in the solid wastes are filter fines, sediments,
pond muds, bed materials, and miscellaneous materials. The wastes
are shipped in 208 liter (55 gallon) drums. The activity disposed
of per year in the solid waste is estinated as (51&):
Natural uranium 0.5 Ci
Natural thorium 4.0 Ci
Uranium-234 0.5 Ci
Thorium-234 0.6 Ci
Protactinium-234 0.6 Ci
Radium-226 0.5 Ci
6.7 Ci
3.3.1 Radioactive Effluents^from Model Conversion Facilities
Because regulations have not required the reporting by con-
version facilities of the total amounts of each radionuclide dis-
charged per year, the source terms chosen for the two types of
model facilities are based on information assembled from a variety
of sources that, in many cases, cannot be adequately documented,
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82
although much reliance has been placed on reference 2_ and sup-
porting material referred to therein. Source terms listed in
tables 3-2 and 3-3 are believed to be reasonably accurate esti-
mates of the quantities of radioactive materials discharged to
the air and water pathways by operating facilities but are nor-
malized to a production rate of 5,000 MTU per year.
3.4 Radiological Impact of Conversion Facilities
Estimates of the radiation doses received by individuals in
the vicinity of the two types of conversion facilities from their
routine effluents are presented in tables 3-4 and 3-5, for doses
through the air pathway and the water pathway, respectively.
The estimated aggregate doses to the population in the vicinity
of conversion facilities are given in table 3-6. The models for
the dispersion and radiation dose calculations are discussed in
appendix A.
3.5 Health Effects Impact of a Model Conversion Facility
The expected cost in health effects to members of the general
population in the vicinity of a model conversion facility are
presented in tables 3-7 and 3-8 for facilities using the solvent-
extraction process and the hydrofluor process respectively. The
models used for the calculation of health effects are given in
appendix A.
3.6 Control Technology
3.6.1 Airborne Effluent Control Technology
The major airborne waste control systems for conversion of
-------
Table 3-2
Discharges of radionuclides to the environment from a model conversion facility9
using the wet solvent extraction process
Radionuclide
Uranium
Uranium
Radium-226
Thorium-230
Pathway
Air
Water
Water
Water
Possible chemical
states
U3°8> U02
UFg, U02F2
uo2+
Ra++
Th++
Source term
(Ci/yr)
0.02 (insoluble)
0.008 (soluble)
2 (soluble)
0.006 (soluble)
0.0006 (soluble)
oo
U)
'Each facility supports twenty-eight 1 6W(e) power plants
-------
Table 3-3
Discharges of radionuclides to the environment from a model conversion facility9
using the hydrofluor process
Radionuclide
Uranium
Uranium
Radi um-226
Thorium-230
Pathway Possible chemical
y states
Air U30g, U02
UF6, U02F2
Water uot*
Water Ra++
Water Th++
Source term
(Ci/yr)
0.04 (insoluble)
0.02 (soluble)
0.8 (soluble)
b
b
Each facility supports twenty-eight 1 GW(e) power plants
Information not available
oo
-------
Table 3-4
Radiation doses to individuals in the general population in the vicinity of a
model conversion plant, through inhalation
Source term
(Ci/yr)
Wet solvent
Critical
organ
extraction process
0.03 Lung
(uranium) Bone
Hydrofluor process
0.06
(uranium)
Lung
Bone
Maximum dose to
Individual at
plant boundary u
(mrem per yr/facility-yr) '
29
0.2
72
0.5
critical organ
Individual within
80 km b
(mrem per yr/facility-yr) '
7 x 10"!
5 x 10"15
2 x 1Q-_1
1 x 10 4
JEach facility supports twenty-eight 1 GW(e) power plants
'Listed mrem per yr radiation dose will result from each year of facility operation
oo
Ln
-------
Table 3-5
Radiation doses to individuals in the general population in the vicinity of a
model conversion plant, through drinking water
Source term
(Ci/yr)
Maximum dose to
Cr1t1cal Individual at
organ plant boundary ,
(mrem per yr/facility-yr) '
critical organ
Individual within
300 km .
(mrem per yr/facility-yr) '
Wet solvent extraction process
2
(uranium)
6 x 10"3
(radium- 226)
6 x 10"4
(thorium-230)
Hydrofluor process
0.8
(uranium)
Bone
Soft tissue
Bone
Soft tissue
Bone
Soft tissue
Bone
Soft tissue
2
0.2
9 x ™~_l
3 x 10 *
8 x 10"5
1
0.1
0.2 9
2 x 10 £
9 x 10~J
3 x 10"5
8 x 10"6
0.1
0.01
aEach facility supports twenty-eight 1 GW(e) power plants
bListed mrem per yr radiation dose will result from each year of facility operations
00
0\
-------
Table 3-6
Aggregate dose to the general population in the vicinity of a model conversion facility
Source term
(Ci/yr)
Pathway
Critical
orqan
Aggregate dose to population
(dose to critical organs)
(rem per yr/facility-yr)a>b
Wet solvent extraction process
0.02
(uranium)
2
(uranium)
0.006
(radium-226)
0.0006
(thorium-230)
Hydrofluor process
0.06
(uranium)
0.8
(uranium)
Air
Water
Water
Water
Air
Water
Lung
Bone
Bone
Soft tissue
Bone
Soft tissue
Bone
Lung
Bone
Bone
Snft tissue
10
0.08
140
14
0.6
0.02
0.005
25
0.2
55
6
facility supports twenty-eight 1 GW(e) power plant?.
Yisted organ rem will result from each year of facility operations.
00
-------
Table 3-7
Health effects to members of the general population in the vicinity of a model conversion
facility using the wet solvent extraction process
Pathway
Air
Water
Total
Critical
organ
Lung
Bone
Bone
Soft tissue
Health effects per facility-year3
Mortality
0.0005
a
0.004
0.002
0.006
Nonfatal effects
0
a
0.002
0.002
0.004
Genetic effects
0
a
0
0.002
0.002
Total health effects for 30 years of plant operations are 0.4.
aListed health effects will result from each year of facility operations.
00
oo
-------
Table 3-8
Health effects to members of the general population in the vicinity of a model conversion
facility using the hydrofluor process
Pathway
Airborne
Waterborne
Total
Critical
organ
Lung
Bone
Bone
Soft tissue
Health effects
Mortality
0.001
b
0.002
0.0008
0.004
jDer facility-year3
Nonfatal cancers
0
b
0.0009
0.0008
0.002
Genetic effects
0
b
0
0.0008
0.0008
Total health effects for 30 years of plant operations are 0.2.
aListed health effects will result from each year of facility operations.
Not significant compared to bone dose from water pathway.
00
-------
90
uranium ore concentrates to uranium hexafluoride combine product
recovery and waste control. Table 3-9 lists the systems used
within the different product handling areas of the hydrofluor
facility (5). Costs are estimated for the bag filter systems
using the air flow rates as listed. However, the number of bags
per unit differs and thus costs based on air flow rates alone
may contain errors.
Several major problems are encountered when adequacy of pre-
sent waste treatment systems is addressed. Uranium conversion
facilities use waste treatment systems which are not discussed
in detail in the open literature, and there is lack of discus-
sion on the effectiveness of the waste treatment systems compared
to the corresponding source terms for these facilities. For air-
borne waste treatment, the systems are not specifically aimed at
control of radiological hazards, but rather are combined with
control of chemically toxic effluents (3). The gaseous wastes
contain fluorides, nitrates, and other chemicals which must be
removed before being exhausted to the atmosphere.
3.6.2 Waterborne Effluent Control Technology
In conversion facilities the control technology for removal
of radioactive materials from liquid effluents is combined with
control technology for chemical wastes (5). Of the two existing
conversion facilities, one recovers uranium from wastes by a wet
chemistry recovery process and the other combines the uranium-
bearing wastes with fluoride liquid wastes to be impounded in
-------
Table 3-9
Airborne waste treatment effectiveness and costs for the hydrofluor process
Process
U03
Fluorinator off-gas
Genera] dust collection
Air cleaning device3 Air
2 bag filters in parallel
(50 bags)
1 bag filter in series
with above (30 bags)
1 bag filter (30 bags)
1 bag filter (20 bags)
in series with above
2 bag filters in series
(128 + 128 bags)
2 bag filters in series
(128 + 96 bags)
2 bag filters in series
2 bag filters in series
(25 + 25 bags)
Total
flow treated (cfm)c
1,400
1,400
2,300
6,000
6,000
3,000
2,100
Capital costb
$ 3,000
3,000
6,000
20,000
20,000
14,000
10,000
$82,000
Annual operating cost
(@ 0.05/cfm)c
$ 168
84
138
720
720
360
252
$2,580
aBased on Allied Chemical Corporation system.
Capital cost based on automatic cleaning bag filters (6)
C1.0 cfm equals 1.7 cubic meters per hour.
-------
92
limestone-lined ponds. Descriptions of those ponds are avail-
able (4), but estimates of costs for the waste lagoons at the
one plant could be quite inaccurate when applied at other loca-
tions because of varying costs of construction. The combining
of radiological waste treatment with chemical waste treatment
prevents accurate estimation of the fraction of the costs in-
curred for radioactive waste treatment.
Improvement in control of radioactive materials can be ef-
fected by the application of more stringent liquid waste treat-
ment. The improvements in liquid waste treatment which may be
brought about by application of the 1972 amendment to the Federal
Water Pollution Control Act to control chemical wastes should
reduce the impact of radioactive liquid effluents (_7). The costs
and effectiveness of the technology which would be applied to
conversion facilities are presently unknown.
3.6.3 Solid Wastes
Solid wastes from conversion facilities are not expected to
result in health effects commitments from wastes onsite. Ship-
ments to commercial burial sites are discussed in section 6.
3.7 Environmental Controls
The effect of environmental control systems on total curies
discharged, maximum radiation dose to an individual, total health
effects for 30 years of plant operations, capital and annual oper-
ating costs of environmental control systems are given in tables
3-10 and 3-11 for both types of model plants. Details of the
-------
Table 3-10
Radiological impact of airborne effluents vs controls for uranium conversion facilities
Controls
Source
term
(Ci/yr)
Met solvent extraction process
None
Bag filters*
Additional bag .
filter 1r» series0
Hydrqfluor process
None
Bag filters8
Additional bag .
filter 1n series5
>2
0.02
0.01
>6
0.06
0.02
Max. dose to the
individual
(mrem/yr, lunq)
>3,000
30
3
>7,000
70
7
Total health effects
30 yr plant operations
(H.E./facility-30 yr)
>15
0.015
0.001
>4
0.04
0.004
Capital
cost
(1970 $/fac.)
0
3,000
3,000
0
82,000
82,000
Annual
operating
cost
(1970 $/fac.)
0
100
100
0
2,600
2,600
Present worth
(1970 $/fac. )
0
11,000
1 1 ,000
0
190.000
190,000
Current level of control
DAdd-on controls remove Insoluble aerosols only
H.E. - health effect
-------
Table 3-11
Radiological impact of waterborne effluents vs controls for uranium conversion facilities
Controls
Source
term
(Ci/yr)
Max. dose to the
individual
(mrem/yr, bone)
Total health effects
30 yr plant operations
(H.E./facility-30 yr)
Wet solvent extraction process
None
Settling pondsa
Additional
treatment
Hydrofluor process
None
Settling pondsa
Additional .
treatment
>20
2
0.2
>8
0.8
0.08
>25
2
0.2
>10
1
0.1
>4
0.4
0.04
>1
0.15
0.01
Capital
cost
(1970 $/fac.)
0
-
1,000,000
0
-
1,000,000
Annual
operating
cost
(1970 $/fac.)
0
20 ,000b
1,000,000
0
20,000C
1,000,000
Present worth
(1970 $/fac.)
0
240,000
14,000,000
0
240,000
14,000,000
Current levels of control
^Estimated as one man-year of effort for radioactive materials control
'Decontamination factor of 10
H.E. - health effect
-------
95
economic models are given in appendix B.
3.7.1 Environmental Controls on Airborne Waste Streams
It was assumed that removal of all controls would increase
the amounts of radioactive effluent discharged by a factor of
>100. The add-on control was assumed to be bag filters that
would reduce the amount of insoluble aerosols discharged by a
factor of 10. The numbers of health effects as a function of
controls were adjusted to conform to these factors.
3.7.2 Environmental Controls on Water Waste Streams
It was not known what the effect of removal of the various
waste treatment systems would be, but it was assumed that the
quantities discharged would increase by a factor of >10. The
add-on control was assumed to be flocculation followed by a
settling pond that would reduce the amount of uranium discharged
by a factor of 10. The numbers of health effects as a function
of controls were adjusted to conform to these factors.
3.8 Summary
The highest radiation doses from these facilities are to the
lung (70 mrem/yr; 30 mrem/yr) of individuals living within 1 km
of the plants and are caused by insoluble uranium aerosols. It
is believed that additional filtration of air streams can reduce
this dose rate by at least a factor of 10.
The largest amount of radioactive material discharged is that
of 2 curies/yr to the water pathways from a solvent extraction
process plant. It is believed that this amount of discharge can
-------
96
be reduced by at least a factor of 10 by addition of waste
treatment systems.
An average of 0.3 health effects are to be expected from 30
years of plant operations under current levels of effluent controls.
4.0 Uranium Enrichment Facilities
4.1 Description of the Uranium Enrichment Industry(l)
Natural uranium contains about 0.7% of fissionable uranium-
235. Light-water nuclear power reactors, however, utilize ura-
nium that is enriched in uranium-235 to the range of 2-4%. Gas-
eous diffusion is the technology that has been developed in this
country for performing the enrichment operation. Uranium is en-
riched by pumping the volatile uranium hexafluoride through a
system of numerous porous barriers. These barriers discriminate
against the passage of the heavier isotope of uranium by a theo-
retical maximum enrichment factor of 1.0043. Existing plants
would require about 1,700 barrier stages to produce a material
of 4Z uranium-235. The uranium hexafluoride gas is driven through
the barriers by compressors driven by electric motors. It is
the compression of the gas that generates process heat which in
turn requires cooling water that is ultimately discharged into
the environment as thermal effluent. The electric motors re-
quire very large quantities of electricity, the generation of
which causes additional effluents to be discharged into the
environment from the electric power generating plants. The
-------
97
gaseous diffusion plants also produce uranium hexafluoride
depleted in uranium-235 (0.25%) which is stored as a solid in
cylinders at the plants.
There are three government-owned gaseous diffusion plants in
the United States. They were built between 1943 and 1955 and
are located at Oak Ridge, Tennessee; Paducah, Kentucky; and
Portsmouth, Ohio, on sites chosen for their remote location and
low surrounding population densities. Distances from the gas-
eous diffusion plants to population centers are given in table
4-1. The plants average about 800 meters to their nearest site
boundaries.
Figure 4-1 gives the mode of operation for the existing gas-
eous diffusion plants. The current complex of plants has a pro-
duction capacity of about 10,000 metric tons of separative work
2
units (SWU) per year, enough to support 90 GW(e)-years of elec-
tricity generated by light water reactors.
It is planned to increase the capacity of the existing three-
plant complex by a factor of 2.5 by 1980. This will be accom-
plished by improving and upgrading the present units and will be
2
A separative work unit (SWU) is a measure of the effort
expended to separate a quantity of uranium of a given assay
into two components, one having a higher percentage of ura-
nium-235. Separative work is expressed in kg units to give
it the same dimensions as material quantities.
-------
98
Table 4-1
Distances to gaseous diffusion plants
from nearby population centers and UFg production plants (2j
Gaseous
Diffusion
Plant
Oak Ridge
Paducah
Portsmouth
Nearby
City
Oak Ridge,
Term.
Knoxvil le,
Tenn.
Paducah,
Ky
Waverly,
Ohio
Portsmouth
Ohio
population
Miles
13
30
16
12
20
centers
Population
28,000
170,000
31 ,000
5,000
28,000
UFg Production Plants
Allied Chemical
Metropolis, 111.
200
20
400
(miles)
Kerr-McGee
Gore, Okla.
600
500
900
-------
99
SHIPMENTS
TO INDUSTRY
AND GOVERNMENT
FEED
(VARIOUS ASSAYS)
NATURAL fEED
0.711%
INHRPLANT SHIPMENTS
0.3 TO 0.55%
SHIPMENTS
TO INDUSTHY
AND GOVERNMENT
FEED
(VARIOUS ASSAYS)
NATURAL FEED
0.711%
INTEBPLAMT SHIPMENTS
0.3 TO 0.55%
Figure 4-1
MODE OF OPERATION FOR GASEOUS DIFFUSION PLANTS
(% Values Ate Weight % U-235)
-------
100
enough to meet the projected 1980 industry demands. At the
present time the existing production capability of the plants is
only partially used for comnercial production of LWR fuel. Cap-
acity for this purpose can be increased to 10,000 MT SWU only by
diverting capacity now used for other government needs.
4.2 The Model Facility
A system of model plants has been assumed for each segment
of the nuclear fuel cycle to achieve a common base for compari-
son of radiation doses, committed health effects and also of
radioactive effluent control technology. The model plant is
defined in terms of a contribution to the nuclear fuel cycle
that is consistent with current and projected commercial indus-
try practice.
The characteristics of the model enrichment plant shown in
figure 4-2 are assumed to be those of the current enrichment
plant complex described in section 4.1 and are identical to the
model plant described in reference (1). The production capacity
is 10,000 MT SWU per year and will support the requirements of
ninety 1 GW(e) light water reactors.
Radiation dose rates and health effects that might result
from the model facility were calculated using standard proce-
dures that are common to all facilities considered in this
examination of the fuel supply. These assumptions are discussed
in appendix A.
4.3 Release of Radioactive Effluents from Enrichment Facilities
Gaseous diffusion plants are large complexes, processing
-------
ENRICHMENT COMPLEX
URANIUM-2350.7-*-4%
OTHER LWR
PLANTS
90 TOTAL
UF,
CONVERSION
MILLING
FABRICATION
FUEL
NOTES: CAPACITY 1972 • 10,000 SWU
1980 • 28,000 SWU
PLANT • 1,500 ACRES
s x 10"GAL. WATER/YEAR
28.5 x 10 6 MW - HR./YR.
ELECTRICAL ENERGY
X
1 GWIe)
MODEL
LWR
LWR-EQA
116 SWU MT/YR.
52 MT/YR. UFg ENR
35 MT/YR. FUEL
Figure 4-2. Model plant characteristics
-------
102
large quantities of materials, and having many effluent streams.
The effluent streams bearing radioactive materials are limited
to a few types, resulting in releases of uranium to the air and
river water, most probably as UO F . Reported release quantities
are very small compared to permissible discharge limits and con-
sist solely of uranium. However, the large quantities of ura-
nium necessarily produce radioactive daughter products during
storage and processing which must be handled in some manner.
Those which decay to uranium-234, a comparatively worthless
isotope, can eventually, via the uranium recovery facilities, be
shipped out as uranium-234 in the product uranium. Other radio-
nuclides, daughters of uranium-234 and minor contaminants of re-
cycle fuel, must be handled through some waste disposal system.
Since they are not reported as being present in effluents, it is
assumed for the present that 100 percent of them go into solid
waste. The effluent uranium is reported as natural uranium,
which is selected as a representative isotopic mix because in
producing low-enrichment LWR fuel by gaseous diffusion, the por-
tion of plant capacity which is processing depleted uranium is
comparable to the portion which is processing enriched uranium.
Effluent data for uranium enrichment plants are minimal.
The Atomic Energy Commission (1) has provided source data on
quantities of uranium discharged. These data are given as kilo-
grams of uranium in gaseous and liquid effluents and are "based
on releases which occurred in 1971, the annual releases attri-
-------
103
butable to the support of a 1000 MW(e) LWR...." (2). Because
the plants furnish enrichment services for several other pur-
poses, e.g., naval reactor fuel, these data are uncertain; and,
the task of determining what portion of the measured effluents
that can be attributed to supporting the fuel requirements for
LWR's is complicated. While the releases indicated represent
roughly 0.007 percent of the material processed, they are remark-
ably small in view of the amount of processing performed.
Solid wastes consist of sludge from onsite holding ponds.
The sludge is collected and buried onsite. The AEC estimates
the sludge amounts to less than 1 metric ton of uranium per
annual LWR fuel requirement requiring less than 0.01 acre per
year (2).
No data other than that provided by the AEC in the "Environ-
mental Survey of the Nuclear Fuel Cycle" has been made avail-
able. However, the treatment used by the AEC of converting kg
quantities to curies of uranium enriched to 4 percent uranium-
235 was considered unsuitable because this is about the upper
limit of enrichment for LWR fuel. Because similar amounts of
enrichment plant capacity are processing enriched uranium as
well as depleted uranium, the uranium-235 enrichment value of
natural uranium was selected as the representative enrichment
for this calculation. The kilogram releases were accordingly
converted to curies of natural uranium (0.7% uranium-235).
Table 4-2 gives the amounts of radioactive material assumed
-------
Table 4-2
Discharges of radionuclides to the environment from a model enrichment facility3
Radionuclide Pathway
Possible
chemical
states
Source term
(Ci/yr)
Uranium
Air
Water
U02F2
0.05
0.6
Each facility supports ninety 1 GW(e) power plants.
-------
105
to be discharged from a model enrichment facility.
4.4 Radiological Impact of Enrichment Facilities
Small quantities of uranium are released under controlled
conditions to the environment from enrichment facilities. This
material is transported through the environment by atmospheric
and liquid exposure pathways and results in a dose to man.
4.4.1 Atmospheric Pathways
The atmospheric pathway considered most significant for this
analysis was inhalation of and the subsequent dose from airborne
uranium. Airborne pathway dose computations are summarized in
appendix A and the results are listed in table 4-3.
4.4.2 Liquid Pathways
The most significant liquid exposure pathway was considered
to be ingestion of uranium bearing drinking water. Dose compu-
tations via this pathway are also discussed in appendix A and
the results are listed in table 4-4.
The aggregate dose to individuals listed in table 4-5 were
computed by multiplying the respective per capita organ dose by
the number of persons receiving the organ dose.
4.5 Health Effects Impact of a Model Enrichment Facility
The health effects impact of a model enrichment facility are
discussed in appendix A together with the health effects inpact
of other fuel supply components. The projected health effects
from operation of a model enrichment facility are listed in
table 4-6.
-------
Table 4-3
Radiation doses to individuals in the general population
in the vicinity of model enrichment_plant, through inhalation3
Source
term
(Ci/yr)
0.05
(uranium)
Maximum dose to
Critical , ,. . , ,
orqan Individuals
9 at plant boundary
(mrem/yr per facility-yr)
Bone 1
critical organ
Individuals
within 80 km
(mrem/yr per facili
3 x 10"4
ty-yr)
Each facility supports ninety 1 GW(e) power plants.
b,,
Listed mrem per yr radiation dose will result from each year of facility operations.
-------
Table 4-4
Radiation doses to individuals in the general population
in the vicinity of a model enrichment plant, through drinking water
Source
term
(Ci/yr)
0.6
(uranium)
Critical
organ
Bone
Soft tissue
Maximum dose
to critical organs
Individuals Individuals
at plant boundary within 300 km
(mrem/yr per facility-yr) (mrem/yr per facility-yr)
0.7
0.07
0.07
0.007
-------
108
Table 4-5
Aggregate dose to the general population in the vicinity
of a model enrichment facility
Sourcp
tprm Pathway Critical Aggregate dose to individuals
(Ci/yr) organ (rem/yr per facility-yr)a'b
0.05 Airborne Bone 0.4
(uranium)
0.6 Waterborne Bone 40
(uranium)
Soft tissue 4
Q
Each facility supports ninety 1 GW(e) power plants
Listed aggregate dose will result from each year of facility
operations
-------
Table 4-6
Health effects to members of the general population in the vicinity of a model enrichment facility
Pathway
Air
Water
Critical
organ
Bone (bone cancer
& leukemia)
Bone (bone cancer)
Bone (leukemia)
Soft tissue
Mortality
c
1 x 10 b
7 x lO'4
4 x 10"4
6 x 10"4
. 2 x in
Health effects/facility-yra
Nonfatal effects
_6
7 x 10 b
7 x 10"4
0
6 x 10"4
13 x 10"4
Genetic effects
0
0
0
6 x 10"4
6 x 10'4
Total health effects for 30 years of plant operations—0.1
aListed health effects will result from each year of facility operations.
-------
110
4.6 Control Technology
4.6.1 Airborne
Present information indicates that the treatment of gaseous
wastes in uranium enrichment facilities is part of the product
recovery system (_3). The costs and efficiencies of this system
for uranium removal are unknown (3).
Because uranium enrichment technology is classified for
national security reasons, the possible changes in waste manage-
ment systems for control of uranium isotopes released to the
environment are unknown.
4.6.2 Liquid Effluents
The liquid effluents from uranium enrichment facilities pass
through process cleanup for recovery of uranium and chemicals
such as fluorides and nitrates. The treated water is released
to settling ponds which discharge into a nearby stream (1.) . The
costs and effectiveness of existing waste control technology are
not presently available.
Possible improvements in liquid effluent control may result
from application of the Federal Water Pollution Control Act Amend-
ments of 1972 (5). The costs and effectiveness of the possible
improvement are unknown.
4.6.3 Solid Wastes
Miscellaneous solid radioactive wastes are incinerated and
the ashes are processed through the uranium recovery plant (1).
The tails from the enrichment process are collected and stored
-------
Ill
(1). There are no anticipated health effects from solid wastes
produced by the enrichment plant.
4.7 Environmental Controls - Enrichment Facilities
Specific information on liquid and gaseous waste control
systems is not available. Therefore, a detailed evaluation of
the reduction in the radiological impact on the environment from
the use of additional controls cannot be made. A first order
approximation of the value, in terms of environmental effects,
of the present control technology versus no controls is shown in
tables 4-7 and 4-8. In these tables, it is assumed that radio-
activity releases with no waste control systems in effect would
be approximately two and one order of magnitude higher for the
air and water waste control system, respectively.
4.8 Summary
The highest radiation dose from the model enrichment facility
is expected to be less than 2 mrem per year (bone) to the maxi-
mum exposed individual delivered in about equal amounts through
inhalation and drinking water.
Approximately 0.1 health effects are to be expected from 30
years of plant operations under current levels of effluent con-
trols.
Less than 1 curie per year of uranium is discharged.
5.0 Fuel Fabrication and Scrap Recovery
5.1 Description of the Fuel Fabrication Process
Fuel for the light water power reactor is fabricated from
-------
Table 4-7
Radiological impact of airborne effluents vs controls for uranium enrichment facilities
Control s
None
Cold traps3
Source
term
(Ci/yr)
>5
0.05
Max. dose to the
individual
(mrem/yr)
>0.01 (Bone)
1 (Bone)
Total health effects
30 yr plant operation
(H.E./facility-30 yr)
>3 x 10"2
3 x 10"4
Capital
cost
(1970 $/fac.)
0
Unknown
Annual
operating
cost
(1970 $/fac.)
0
Unknown
Present worth
(1970 $/fac.)
0
Unknown
and aluminium
traps
Current levels of controls.
H.E. - Health effects
-------
Table 4-8
Radiological impact of waterborne effluents vs controls for uranium enrichment facilities
Source
term
Controls (Ci/yr)
None > 6
Chemical3 0.6
treatment and
settling ponds
Max. dose to the
individual
(mrem/yr)
>8
0.7
Total health effects
30 yr plant operation
(H.E./facility-30 yr)
>1
0.13
Capital
cost
(1970 $/fac.)
0
-
Annual
operating
cost
(1970 $/fac.)
0
20,000b
Present worth
(1970 $/fac.)
0
240,000
aCurrent levels of controls
Estimated as one man year effort for radioactive materials control
H.E. - Health effects
-------
114
uranium hexafluoride enriched to 2-4% in the uranium-235 iso-
tope. The slightly enriched uranium hexafluoride is shipped from
the uranium enrichment facility to a fuel fabrication facility
(in sealed 2,300 kg cylinders) where it is hydrolyzed to uranyl
fluoride, converted to ammonium diuranate, and calcined to the
dioxide. The dioxide powder is pelletized, sintered, and loaded
into stainless steel or Zircaloy tubing which is then capped and
welded. A process flowsheet is presented in figure 5-1 (1).
The fuel rods, each about 3.7 meters (12 feet) long and slightly
less than 13 mm (1/2 inch) in diameter, are assembled in arrays
to be handled as fuel assemblies.
A list of fuel fabrication plants, their products, and site
data are given in tables 5-1 and 5-2 (1). Scrap material from
fuel fabrication is dissolved in nitric acid, purified by sol-
vent extraction, calcined and reduced to the dioxide which may
be cycled back to the fabrication process.
The most significant chemical effluents are fluorine, fluo-
rine compounds and nitrogen compounds. The bulk of the fluorine
released from the UFfi appears ultimately as solid CaF resulting
from lime neutralization.
5.2 The Model Facility
The model fuel fabrication plant is assumed to be identical
to the one described in reference JL. The model site for the
plant that is assumed for the purposes of dose calculations has
the general characteristics discussed in appendix A of this report.
-------
BF6IM
2300KE CYLINDERS
TO ATMOSPHERE
CERAMIC U02POWDER
HEAT
VOLATIL-
IZATION
(aj
HEPA*
FILTER
CALCINATION
(f)
f f
HEAT H2,N2
\
WATER HH4OH
1 I
HYDROLYSIS
SCRUBBER
'
DRYING
(e)
PREC
III
it
X
CONDI
(d
IPITA-
)N
b 1
ENTRA-
IN
)
f f 1
N, HEAT WASTE LIQUOR T
2 TREATMENT
*HIGH EFFICIENCY
PARTICULATE
AIR FILTER
Figure 5-1 Fuel fabrication-chemical processing (ADD) block diagram
-------
Table 5-1
LWR fuel fabrication plants (3)
Licensee
Plant
location
Plant feed
material
Plant
product
Babcock & Wilcox
Combustion Engineering
General Electric
Gulf United Nuclear
Gulf United Nuclear
EXXON
Kerr-McGeea
Nuclear Fuel Services3
NUMEC
Westinghouse
Lynchburg, Va.
Windsor, Conn.
Wilmington, N.C.
Hematite, Mo.
New Haven, Conn.
Richland, Wash.
Crescent, Okla.
Erwin, Tenn.
Apollo, Pa.
Columbia, S.C.
U02 pellets
U02 powder
UF6
UF6
U02 pellets
UF6
UF
UF
UF
Fuel assemblies
Fuel assemblies
Fuel assemblies
UCL powder or pellets
Fuel assemblies
Fuel assemblies
UCL powder or pellets
UCL powder or pellets
U02 powder or pellets
Fuel assemblies
JKerr-McGee and Nuclear Fuel Services data are from USAEC Regulatory files,
-------
Table 5-2
Fuel fabrication plants - site size and demography (4_)
Nearby population centers
Plant location
Babcock & Wilcox
Lynchburg, Va.
Combustion Eng.
Windsor, Conn.
General Electric
Wilmington, N.C.
Gulf United Nuclear
Hematite, Mo.
Gulf United Nuclear
New Haven, Conn.
EXXON
Richland, Wash.
Kerr-McGee
Crescent, Okla.
Nuclear Fuel Services
Erwin, Tenn.
NUMEC
Apollo, Pa.
Westinghouse
Columbia, S.C.
Site size
(hectares)
205
215
668
61
31
a
65
405
24
2
461
Population density
(people/km^)
16
240
20
120
240
8
40
40
160
50
City
Lynchburg
East Granby
Windsor
Castle Hayne
Wilmington
Hematite
St. Louis
Hartford
New Haven
Richland
Crescent
Oklahoma City
Erwin
Johnson City
Apollo
Pittsburgh
Columbia
Population
54,000
35 ,000
22,500
700
46,000
< 2,500
622,000
158,000
138,000
26 ,000
1,600
363,000
4,700
33,800
< 2,500
520,000
113,500
Distance
(km)
6
5
8
3
13
1
53
15
0
5
8
48
2
21
0
40
13
aShared by manufacturing and research divisions of Olin Corporation and naval reactor fuel
operations of United Nuclear Corp.
-------
118
In addition, the plant has particular characteristics re-
lated to its function in the fuel cycle. These are indicated in
figure 5-2.
Other requirements for the model plant include:
1. 40.5 hectare (100 acre) site with buffer zone of at
least 500 meters to the nearest site boundary.
2. 1.7 million liters (450,000 gallons) of water per day
or 20 million liters (5.2 million gallons) per LWR annual fuel
'requirement.
3. 100,000 cubic meters (3.6 x 10 scf) natural gas for
process heat in fabrication of annual LWR fuel requirements.
For the purposes of this analysis, a model fuel fabrication
plant has a capacity of 3 MTU per day and operates 300 days per
year. Assuming a lifetime for the plant of 30 years, 780 fuel
requirements for the model LWR are available.
5.3 Radionuclide Effluents from Fuel Fabrication Facilities
The majority of the present fuel fabrication plants perform
all the post enrichment operations necessary to produce finished
fuel assemblies, including converting UFfi to U09, making the UO^
pellets, putting the pellets into cladding tubes, and putting
the tubes together to form fuel assemblies. Due to leakage, spil-
lage and breakage, some of the enriched uranium is released to
the waste streams of the plant, and small quantities escape, most
probably as UO F or UO .
-------
119
^
MODEL
LWR
IGW(e)
40 MTU02
35 MTFUEL
FABRICATION UNIT
UO
POWDER
PELLETS
2~ UF6
ZIRCALOY
OR
STAINLESS
ASSEMBLIES
. FROM
6 ENRICHMENT
FACILITIES
OTHER
LWR'S
FUEL
AREA - 40.5 HECTARES
WATER - 1.7 MILLION LITERS/DAY
POWER • 6 MW(e)
-• 1,700 MW-HR. (ANNUAL)
OR 620 MT COAL
OUTPUT - 3 MTU/DAY FOR 300 DAYS/YR
LIFE - 30 YEARS
Figure 5-2 Model fuel fabrication plant
-------
120
Estimated radioactive effluent quantities are given in table
5-3. The airborne radionuclides are conservatively assumed to
be in the form of insoluble aerosols; the radionuclides in liquid
effluents are probably in solution. Upon consideration of the
available data and the intent to base the estimates upon good
current practice, the values used by the AEC in their "Environ-
mental Survey of the Nuclear Fuel Cycle" (3) were used for these
estimates. The decay of the uranium isotopes -234, -235, and
-238 in natural uranium provides 0.7 Ci per MTU compared to ura-
nium enriched to 3.2 percent uranium-235 which provides about
1.8 Ci per MTU.
The economic incentive to minimize losses of uranium in
effluents is only weakly related to the radioactivity of the
effluents. Releases at a constant rate in kilograms per year
become more hazardous as the radioactivity per kilogram increases.
This may be expected to occur. Most of the radioactivity in the
uranium released is due to uranium-234 which is relatively worth-
less. Uranium coming from enrichment facilities is expected to
contain higher percentages of uranium-234 in the future as more
recycled uranium fuel is used. The recycled fuel will also bring
radioactive impurities with it, although at extremely small con-
centrations. Because some of this material will find its way
into effluents, careful watch over the amount and nature of
radioactivity in effluents is necessary.
Uranium scrap recovery operations are performed both in the
-------
121
Table 5-3
Discharge of radionucTides to the environment from a model
uranium oxide fuel fabrication facility9
Radionuclide
Uraniumb
Uraniumb
Thorium-234
Pathway
Air
Water
Water
Probable
chemical
state
U02F2;U02
U02++
Th++
Source
term
(Ci/yr)
0.005
0.5
1
aEach facility supports twenty-six 1 GW(e) power plants
bEnriched to 3.2 wt % uranium-235.
-------
122
fuel fabrication plant and in separate facilities. The scrap
recovery operations tend to release a larger fraction of the
material processed to effluent streams than do the fuel fabri-
cation lines. This happens in part because the scrap recovery
operations are relatively small and it is difficult to find
economic justification for the expense of high efficiency waste
treatment systems for them. Available data from one separate
facility indicate that quantities on the order of one percent of
the material processed are released in effluents for that oper-
ation. Its releases of radioactivity to air and water are com-
parable to those of fuel fabrication facilities. However, when
the scrap recovery is contained within the fabrication plant,
scrap recovery effluents become a contribution to the total
plant effluents and indistinguishable from them in available
data. Effluents from uranium scrap recovery operations are
expected to be a minor fraction of those from all LWR fuel
fabrication operations.
Some reported and estimated radioactive effluents for speci-
fic fuel fabrication plants are given in table 5-4. The data
serve to indicate order-of-magnitude values, but many reports do
not relate the effluent data to the quantities processed.
Estimates of solid wastes from fuel fabrication facilities
involve both onsite burial and shipment offsite to commercial
burial sites. Onsite burial of calcium fluoride removed from
liquid wastes streams is estimated as 680 MT per year for a 900
-------
Table 5-4
Reported and estimated effluent uranium quantities
Plant
Gaseous
Exxon (2)
Kerr-McGee (3)
NUMEC (!)
General Electric (jf
Westinghousec
GUNF
0.00015 Ci/yr
0.04 Ci/yr
0.0028 Ci/yra
0.002 Ci/yr
0.2 Ci/yra
0.03 Ci/yr
Liquid
0.08 Ci/yr
0.24 Ci (7.3 mo.)
1.75 Ci (6 mo.)
1.36 Ci/yra
0.3 Ci/yra
0.1 Ci/yr
K)
U)
aEstimated on 900 MT/year basis.
bGeneral Electric Company, private communication with AEC (December 1, 1972)
cWestinghouse Electric Company, private communication with AEC (December 1, 1972)
-------
124
MTU/yr model plant (1). The volume of waste buried per year is
about 220 cubic meters (290 cubic yards or 7,800 cubic feet).
The quantity buried per annual LWR fuel requirement is estimated
as 8.4 cubic meters (296 cubic feet) containing 0.06 Ci of
uranium.
The volume of solidified waste shipped to commercial waste
burial firms is not estimated by the AEG. The radioactivity is
estimated to be 0.0025 Ci per annual LWR fuel requirement or
0.07 Ci per year for the 900 MTU model fuel fabrication plant.
5.4 Radiological Impact of a Model Fuel Fabrication Facility
Estimates of the radiation doses received by individuals in
the vicinity of a model fuel fabrication facility from their
routine effluents are presented in tables 5-5 and 5-6, for doses
through the air pathway and water pathway, respectively. The
estimated aggregate doses to the population in the vicinity of
conversion facilities are given in table 5-7. The models for
the dispersion and dose calculations are discussed in appendix
A.
5-5 Health Effects Impact of a Model Fuel Fabrication Facility
The expected cost in health effects to members of the gen-
eral population in the vicinity of a model fuel fabrication
facility are presented in table 5-8. The models used for the
calculation of health effects are given in appendix A.
5.6 Control Technology
5.6.1 Airborne
-------
Table 5-5
Radiation dose to members of the general population from a model
uranium oxide fuel fabrication facility through inhalation
Source term
(Ci/yr)
Maximum dose to critical organ
Critical organ
Individual at plant boundary
(mrem/yr per facility-yr)a>"
Individual within 80 km
(mrem/yr per facility-yr)
0.005
(uranium)
Lung
10
0.002
aEach facility supports twenty-six 1 GW(e) power plants.
^Listed mrem/yr radiation dose will result from each year of facility operations.
-------
Table 5-6
Radiation doses to individual in the general population in the vicinity
of a model uranium fuel fabrication plant through drinking water
Source term
(Ci/yr)
Critical organ
Maximum dose to critical organs
Individuals at plantAverage individual
boundary dose within 300 km
(mrem per yr/facility-yr)a'D (mrem per yr/facility-yr)
N3
ON
0.5
(uranium)
Bone
Soft tissue
0.6
0.06
0.06
0.006
Each facility supports twenty-six 1 GW(e) power plants.
Listed mrem/y radiation dose will result from each year of facility operations.
-------
Table 5-7
Aggregate dose to the general population in the vicinity
of a model uranium fuel fabrication plant
Source
term
(Ci/yr)
0.005
(uranium)
0.5
(uranium)
Pathway
Air
Water
Water
Critical organ
Lung
Bone
Soft tissue
Aggregate dose to
(rem per yr/facil
3
34
3
populatior
ity-yr)a>b
Each facility supports twenty-six 1 GW(e) power plants
Lfsted aggregate dose will result from each year'of facility operations
-------
Table 5-8
Health effects to members of the general population in the
vicinity of a model fuel fabrication plant
Pathway
Air
Water
Water
Critical organ
Lung
Bone (bone cancer)
Bone (leukemia)
Soft tissue
Total
Mortality
0.0002
0.0006
0.0004
0.0005
0.0016
Predicted health effects/faci
Nonfatal effects
0
0.0006
0
0.0005
0.0011
lity-yra»b
Genetic effects
0
0
0
0.0005
0.0005
^Listed health effects will result from each year of facility operations.
"Total health effects for 30 years of plant operations are 0.1.
to
CD
-------
129
The model uranium fabrication plant has three major gaseous
waste treatment systems. The system for conversion of UF, to
6
U02 is equipped with a scrubber-demister and one high-efficiency
particulate air (HEPA) filter (1). The processes handling UCL pow-
der, pellets, and fuel tube loading are exhausted through the
HEPA filters (1). Scrap recovery chemical systems exhaust
through a scrubber-demister and through one HEPA filter. The
solid wastes incinerator also exhausts through one HEPA filter
(1). For the purposes of this discussion, the conversion (UF,
o
to U02) and scrap recovery are assumed to use a common scrubber-
demister. Each system is equipped with a HEPA filter. The pro-
cess systems handling U02 are assumed to use two HEPA filters in
tandem independent of those on the conversion and scrap recovery
systems. The major fraction of airborne particulates is assumed
to be from the process systems.
Table 5-9 lists the associated cost parameters for the gas-
eous waste systems. The AEC-model 900 MTU capacity fuel fabri-
cation plant was used for describing the air cleaning system.
Costs of the systems are reported as a range of costs. For the
purposes of a later summary table, the midpoint of the range was
used. The capital and operating costs listed in table 5-9 were
reported as annualized costs in the reference quoted. Selection
of the midpoint of the range of costs and inflation may have lead
to somewhat of an underestimate of costs for application to
specific facilities in 1973. However, the ratio of operating
-------
Table 5-9
Uraniun fuel fabrication and scrap recovery gaseous waste treatment effectiveness and costs
Type of
equipment
Scrubber and
demister
Pref liter
1st HEPA
2nd HEPA
3rd HEPA
Fraction
of uranium
removal
0.9(1)
-
0.999a
0.99a
0.94(8)
Estimated13
cfm
treated
105
105
105
105
105
Approximate0
capital cost
($/cfm)
1.50 - 3.00[2)
0.02 - 0.06
0.42 - 0.60(2)
Annual
operating
cost Capital
($/cfm) cost
0.37 - 0.75(2) 150,000 - 300,000
(1964 $)
0.015 - 0.025(2) 2,000 - 6,000
(1960 $)
0.10 - 0.45(2) 42,000 - 60,000
(1960 $)
Annual
operating
cost
37,000 - 75,000
(1964 $)
15,000 - 25,000
(1960 $)
10,000 - 45,000
(1960 $)
Capital
cost
190,000 - 380,000
(1970 $;d
2,500 - 7,500
(1970 $)d
52,000 - 75,000
(1970 $)d
Annual
operating
46,000 - 94,000
(1970 $)d
19,000 - 57,000
(1970 $)d
12,500 - 56,000
(1970 $)d
aFirst and second banks together reported to remove 0.99999(_3) fraction removal apportioned between the filters as indicated.
1.0 cfm =1.7 cubic meters per hour.
cCosts converted to capital cost from annualized costs using a annual fixed charge rate of 16.65-..
dCosts converted to 1970 $ using the Marshall and Stevens Equipment Cost Index.(9J
-------
131
costs to total annualized costs is 75% which is in agreement
with an estimated range of 70-85% in reference 12.
Several major problems are encountered when the question of
the adequacy of present waste treatment systems is addressed.
Uranium fuel fabrication facilities use waste treatment systems
which are not described in detail in the open literature. An
additional problem is the minimal amount of discussion of the
effectiveness of the waste treatment systems and the corres-
ponding source terms for these facilities. The gaseous wastes
also contain fluorides, nitrates, and other chemicals which must
be removed before exhausting to the atmosphere.
Added HEPA filters are a possible way of adding waste treat-
ment. A current practice being introduced in uranium fuel fabri-
cation plants is the use of glove boxes. The primary purpose is
product containment within the plant, but the quantities re-
leased to the environment may also be reduced. Costs of glove
boxes or hoods are not included in the estimates of costs of
control.
5.6.2 Liquid Effluents
The liquid wastes from scrubber-demisters, drains, and
laboratories are collected in settling tanks or ponds. Treat-
ment consists of adding flocculating agents and chemicals for
removal of fluorides and nitrates which are discharged from the
systems for conversion of UF, to UO. and scrap recovery
-------
132
The review of available data on uranium fuel fabrication
plants conducted by Battelle was used for estimating the costs
of liquid waste treatment (10). For 1970, the total industry
had a production of 950 MTU. Table 5-10 lists the data for
costs of control. Annualized costs were calculated using the
factors and assumptions in appendix B.
Possible improvements in control of radioactive materials
could result from the application of more stringent liquid waste
treatment. The improvements in liquid waste treatment which may
be brought about by application of the 1972 amendment to the
Federal Water Pollution Control Act (11) to chemical wastes
could reduce the impact of radioactive materials by also reducing
their discharge quantities. The costs and effectiveness of the
technology which would be applied to fuel fabrication facilties
is not known.
5.6.3 Solid Wastes
Solid wastes from fuel fabrication facilities are not expected
to result in health effects commitments from wastes onsite. Ship-
ments to commercial burial sites are discussed in another section.
5.7 Environmental Controls - Fuel Fabrication
The waste systems, estimated costs and estimated health ef-
fects for airborne releases from fuel fabrication facilties are
summarized in table 5-11. The changes in costs and health effects
for the addition of another HEPA filter in series are also listed.
Table 5-12 lists the present control technology, costs and health
-------
Table 5-10
Uranium fuel fabrication and scrap recovery
liquid waste treatment effectiveness and cost (10)
Tvnp nf Fraction of Capital cost3 . , . .. . T . ,
lype or uranium - Annual Annualized . Total
equipment removal Equipment Facility °Pei"ating cost capital cost annualized cost
— — . ,—
Settling
tanks 0.90(6_) $280,000(6) $56,000(6.) $28,000(6.) $56,000 $84,000
aFor 950 MTU throughout 1970.
Annual fixed charge rate on depreciable capital - 16.6%.
-------
Table 5-11
Radiological impact of airborne effluents vs controls for a model uranium fuel fabrication plant
Controls
Scrubber and
pref ilter base
Plus 1 HEPA
2nd HEPA in
series3
3rd HEPA in
series
Source
term
(Ci/yr)
> 500
0.5
0.005
0.0003
Max. dose to the
individual
(mrem/yr, lung)
> 1,000,000
1,000
10
0.5
Total health effects
30 yr plant operations
(H.E. /facility- 30 yr)
> 500
0.5
0.005
0.0002
Capital
cost
(1970 $/fac.)
0
64,000
64,000
64,000
Annual
operating
cost
(1970 $/fac.)
0
34,000
34,000
34,000
Present worth
(1970 $/fac.)
0
530,000
530,000
530,000
Current level of control
H.E. - health effect
-------
Table 5-12
Radiological impact of waterborne effluents vs controls for a model uranium fuel fabrication plant
Controls
None
Settling tanksa
Precipitation and
flocculation
Source
term
(Ci/yr)
> 5
0.5
0.02
Max. dose to the
individual
(mrem/yr, bone)
> 6
0.6
0.01
Total health effects
30 yr plant operations
(H.E./facility-30 yr)
> 0.9
0.09
0.002
Capital
cost
(1970 $/fac.)
0
340,000
1,000,000
Annual
operating
cost
(1970 $/fac.)
0
28,000
200,000
Present worth
(1970 $/fac.)
0
1,000,000
4,300,000
Current level of control
H.E. - health effect
LO
Ul
-------
136
effects estimates for radioactive materials released to the
water pathway. Costs for liquid waste treatment were calculated
using the factors and assumptions in appendix B. For additional
liquid waste treatment, it was assumed a two-stage precipita-
tion-flocculation system reported in reference 13 could attain
an additional decontamination factor of 50.
5.8 Summary
The highest radiation dose from the model fabrication faci-
lity is 10 mrem per year (lung) to the maximum exposed indivi-
dual living within 1 km of the plant.
Approximately 0.1 health effects are to be expected from 30
years of plant operations under current levels of effluent
controls.
Less than 1 curie per year of uranium is discharged.
6.0 Transportation
6.1 Description and Growth Patterns
An overview of the transportation requirements for the LWR
nuclear power industry is depicted in figure 6-1. Radioactive
materials are currently shipped by rail and truck. However,
barge transportation is projected in the near future. A summary
of the parameters for each transportation pathway is presented
in table 6-1. Predictions of growth patterns are summarized in
table 6-2 for a 900 unit LWR program at 1,000 MW(e) each which
can reasonably be expected to occur by the year 2000. Plutonium
-------
137
U FUEL
FABRICATION
A
K
L»- A *-
U
ENRICHMENT
A
J
U
CONVERSION
i
I\LHUIUK ^"
' A
5 A
, » G
^ "" H CHEMICAL C *•
PROCESSING -^ *~
E D~
Pu V
STORAGE
i
Pu -*^
^^ FUEL — t — ^.
FABRICATION
A
I
MINING
MILLING
LOLEVEL
WASTE
LOLEVEL
WASTE W/Pu
HIGH LEVEL
WASTE W/Pu
LOW LEVEL
WASTE W/Pu
Figure 6-1 Simplified schematic of transportation requirements for the LWR
nuclear power industry
-------
138
Table 6-1
Summary of transportation parameters for the LWR nuclear power industry
Path
(See figure 6-1 )
A
B
C
D
E
F
G
H
I
J
K
L
Material
Miscellaneous
Spent fuel
Miscellaneous
w/transuranics
High level
wastes
w/transuranics
Plutonium
oxide
Miscellaneous
w/transuranics
Recycle fuel
Pu02 + U02
Uranyl nitrate
Uranium oxide
UP.
6
UFC
6
Fresh fuel
Form
Packaged
solids
Spent fuel
assemblies
Packaged
solids and
cladding
Solid
Non-dis-
persible
solid
Packaged
solids
Fuel
assemblies
Liquid
Powder
(yellow
cake)
Powder
(natural U)
Powder
(enriched U)
Fuel
assemblies
Mode
Truck
Rail
Truck
Rail
(Barge)
(a)
Rail
Truck
Truck
Truck
Truck
Truck
Rail
Truck
Truck
Truck
Annual quantity
shipped
to facility
3,000 drums
per reactor
1 ,500 MT
..
500 casks
(141.5 m3)
15,000 kgb
per chemical
plant
(c)
(d)
500 MT
15,000 MT
10,000 MT
750 MT
30 MT
Quantity
per
shipment
50 drums
150 drums
0.5 MT
3.0 MT
--
10 casks
(2.88 m3)
30 kgb
..
_.
5 MT
15 MT
38 MT
11 MT
(11 MT/cask)
11 MT
(2.2 MT/cask)
3 MT
Miscellaneous low and intermediate level wastes are currently stored onsite at the re-
processing plant. Appendix F of 10 CFR Part 50 requires decontamination of reprocessing
sites upon decommissioning which will probably result in shipment of these wastes to a
repository.
Very limited quantities of plutom'um have been shipped to the mixed oxide fabrication
plants. In addition, the quantity of plutonium permitted in shipping containers varies
widely. Thus, it is not possible to estimate the quantities in this path with any accu-
racy at this time. Shipment of plutonium would range from 900 kg/yr for a 30 MT mixed
oxide plant to 4,500 kg/yr for a 150 MT plant assuming a plutonium content of 3% in the
fabricated assemblies.
Insufficient information exists at this time to estimate these quantities. However, it
is suspected volumes will be large with plant startups and then decrease gradually.
This path will probably be similar to path L.
-------
139
Table 6-2
Transportation requirements for a 900 LWR nuclear power program
(low enriched uranium oxide fuel)
Path
(See figure 6-1)
A
B
D
E
H
I
J
K
L
Number of facilities
shipped to/from
from 900 reactors
from 900 reactors
from 18 chemical
plants
from 18 chemical
plants
to 9 enrichment
pi ants
to 6 conversion
plants
to 9 enrichment
plants
to 36 fabrication
plants
to 900 reactors
Total
annual
shipments
54,000 (T)
18,000 (R)
54,000 (T)
9,000 (R)
900 (R)
9,000 (T)
5,400 (T)
6,000 (T)
7,500 (T)
2,600 (T)
9,000 (T)
Average
kilometers
per
shipment
805
1,610
4,025
<800b
1,208
1,610
805
1,208
1,610
Annual
population
dose
(person-rem/yr)
150a
176a
40
Not applicable
0.24C
Negligible
Negligible
0.12C
0.54C
a90% rail and 10% truck weighted values.
In many cases it is suspected the mixed oxide (Pu02 - U02) fabrication plant will
be located close to or adjacent to the chemical reprocessing plant. However, it is
difficult to estimate the average shipping distances at this time.
/»
A dose rate of 0.1 mrem per hour at 3 meters from the apparent centerline of the
shipping route was used for these shipping paths. Thus the population dose per km is
0.01 that listed in table 6-3.
-------
140
recycle is shown In the figure and is summarized in table 6-1,
since current trends indicate that such recycle will occur in
the future,
6.2 Shipping Containers
A variety of containers are required to ship the diverse
radioactive materials listed in table 6-1. Rather than describe
the individual containers, many of which are under development
and/or in design, a list of the major design requirements ap-
pears more appropriate:
a. radiation exposure limits external to the container
b. criticality control
c. heat dissipation
d. survival and maintenance of integrity under accident
conditions, and
e. weight limitations.
6.3 Exposure Levels
Exposures from the normal transportation of radioactive mater-
ials for the nuclear power industry are estimated by assuming
that direct radiation is the only source and using the dose rate
limit imposed by Department of Transportation (DOT) regulations
(10 mrem per hour at 6 feet from the surface of the shipping
vehicle)(1). DOT regulations also inpose requirements on the
dose and temperature at the surface of the vehicle or container.
However, these are of little consequence in estimating popula-
tion exposure. In a recent analysis (2), the AEC used a value
-------
141
of 10 mrem per hour at 10 feet from the apparent centerline of
the shipping route instead of the DOT requirement. This AEG
value was used here (10 mrera/hour at 3 meters) for purposes of
consistency and comparison.
6.4 Radiological Impact
The dose to the population can be calculated by assuming a
uniform population density along a 2 kilometer corridor of the
road or track (1 kilometer on either side of the road or track
centerline). Since the exposure at any distance from the road
is dependent on the given distance and on the speed of the vehi-
cle, it is necessary to integrate the dose to the individual at
the given point as the shipment passes. An equation was set up
and the solution is:
D - individual dose at given distance (mrem)
B - -*— tan'1 [t&1/2I
Cab)1'2
where:
K = 0.025 mrem-m2/s
a = constant = (distance in meters from centerline of path)
b = f(t) = (velocity of vehicle in m/s)2
*. . 2.000 .
-in seconds
velocity of vehicle
The results using this solution and vehicle speeds of 320
km/day (representative of rail transport) and of 966 km/day
(representative of truck transport) are listed in table 6-3.
-------
142
The dose to an individual standing 100 meters from the center-
line of the path is 1.04 x 10 mrem for the passage of one
shipment at 320 km/day.
The AEC presented a similar analysis in a recent report (_2)
wherein corrections were made for a buildup factor and air at-
tenuation. With these corrections numerical integration was
required to solve the equation. An analysis was performed using
the simplified method presented here for the same distance para-
meters used by the AEC. This analysis indicated that the inte-
grated dose for both methods agreed to within 10% of each other.
However, there were significant differences in the distribution
of the dose with distance from the centerline. The method used
here produced lower results close to the shipping path and higher
results at the further distances (0.5 to 1.0 km). It is con-
cluded that both methods are acceptable since a uniform popula-
tion density is considered.
The population dose per kilometer (table 6-3) was obtained
by lumping the population in groups at 25 meter intervals from
25 to 150 meters from the road or track centerline and at 100
meter intervals from 200 to 1,000 meters from the centerline,
and then multiplying by the individual dose at the respective
distance on both sides of the path. These results were then
summed for the particular vehicle speed. The population density
used was 127.4 persons per square kilometer. These calculations
were performed for both rail (320 km/day) and truck (966 km/day)
-------
143
Table 6-3
Individual dose at a given distance from the apparent centerline
of the shipping route for the passage of one shipmenta
Distance
(meters)
25
50
75
100
125
150
200
300
400
500
500
700
800
900
1,000
Population dose
(person-rem per
km)
Dose at .
320 km/ day0
(mrem)
4.24 x 10"4
2.11 x 10"4
1.49 x 10"4
1.04 x 10"4
8.22 x 10"5
6.81 x 10"5
4.99 x 10"5
3.22 x 10"5
2.34 x 10"5
1.82 x 10"5
1.45 x 10"5
1.2 x 10"5
1.02 x 10"5
8.6 x 10~6
7.49 x 10"6
1.1 x 10'5
Dose at .
966 km/ day0
(mrem)
1.4 x 10"4
6.97 x 10"5
4.59 x 10"5
3.41 x 10"5
2.7 x 10"5
2.25 x 10"5
1.65 x 10"5
1.06 x 10"5
7.76 x 10"6
5.99 x 10"6
4.79 x 10"6
3.97 x 10"6
3.36 x 10"6
2.88 x lO"6
2.5 x 10'6
k "
3.7 x 10"6
Based on 10 mrem per hour at 10 feet (^3 m) from the apparent
centerline of the shipping route.
Accurate to one significant figure.
-------
144
transport.
The population doses were then multiplied by the total kilo-
meters traveled for each path, as presented in table 6-2. Summa-
tion of the population doses for each pathway results in the total
dose of 367 person-rem/yr for the 900 unit reactor program. This
yields a value of 0.41 person-red per reactor per year. If a
load factor as low as 0.7 is used, this planned Impact is not
more than 0.6 person-rem per GW(e)-yr.
An additional planned exposure must be considered for radio-
active material shipments to account for the stops made enroute.
Trains stop at rail stations and trucks at truck stops and, in
most cases, it is reasonable to expect that the population densi-
ty will be much greater at the stop than the 127.4 persons per
square kilometer used above. However, for the most part, the
stops will be of short duration. Therefore, this planned ex-
posure should lead to an impact of about the same magnitude as
that from the moving vehicle and it is recommended that the
planned impacts listed above be doubled to account for this
additional exposure (2). The totals are:
Planned impact
from normal = 0.82 person-rem = 1.2 person-rem
transportation reactor-year GW(e)-yr
These values do not include occupational exposure. It is ex-
pected that transport workers will be considered radiation work-
ers with the exception of rail workers (brakemen, conductors,
engineers, etc.).
-------
145
The maximum dose to an individual is expected to occur to
railroad brakenen. In the AEG analysis it was estimated that
brakemen would spend from 1 to 10 minutes in the vicinity of the
cask car during a trip for an average exposure of about 0.5 mrem
per shipment. Assuming a highly unusual set of circumstances in
which one brakeman receives such an exposure from all incoming
shipments of spent fuel to a chemical plant (450 annual shipments),
the maximum individual exposure rate would be 225 mrem per year
to this particular brakeman. However, it appears difficult to
envision such circumstances and, if encountered in practice, steps
could be taken to monitor and/or reduce this exposure.
6.5 Health Effect Impact
The conversion factor used to obtain health effects tor the
radiological impact as estimated in section 6.4 are:
Total body irradiation:
200 deaths per year per 10 annual person-rem;
200 nonlethal cancers per year per 10 annual person-rem.
Gonadal irradiation:
300 serious effects per year per 10 annual person-rem.
Health effects are defined as the sum of lethal, nonlethal and
genetic effects.
For the transportation impact in the postulated 900 unit LWR
nuclear power program the health effects due to radiation exposure
are listed in table 6-4.
6*6 Cost Effectiveness of Reducing Transportation Exposure
-------
Table 6-4
Health effects to members of the general population
from transportation of radioactive materials associated with nuclear fuel cycle
Pathway
Critical
organ
Health effects /reactor-yra
Mnv*»-m-u Nonfatal Genetic
Mortality effects effects
Direct gamma Whole body 0.0002 0.0002 0.0003
dose
Total health effects for radioactive transportation associated with 30 years of
operation of one reactor facility = 0.02
aListed health effects will result from transportation associated with one
year of operation of one reactor.
-------
147
The transportation of radioactive materials in a reactor
program must be viewed in perspective when attempting to perform
a cost-effectiveness analysis for reducing the radiological im-
pact. The quantity of radioactive material which must be shipped
in a reactor program (or per year per reactor) is fixed. The
impact resulting from the shipment of this material is dependent
on the dose rate at a given distance from the shipping container
(or cask). If the dose-rate limit is reduced by a factor of 2,
implying the use of additional shielding, the impact will be re-
duced by a factor of 2. However, from a practical standpoint,
it is much more likely that this lower dose rate could best be
met by reducing the quantity of material in each shipment since
the truck casks are currently at legal weight limits for most
states and rail cask weight must be held to some limit to pre-
vent special routing requirements. Crane capacities must also
be considered. This situation would lead to a lower impact per
shipment but a larger number of shipments, resulting in about
the sane total Impact. Therefore, it is likely that additional
restraints (i.e., other than a dose rate limit external to the
cask) must be considered to insure that the total transportation
impact is reduced if the analysis indicates this is a cost-
effective requirement.
The fabrication costs for the General Electric IF 300 rail
cask for spent fuel are approximately one million dollars per
cask.. Current estimates for the larger G.E. IF 400 cask range
-------
148
between 1.5 x 10 and 1.75 x 10 dollars for fabrication. The
capacities of these casks are about 3 MT and about 6 MT, respec-
tively. Design and licensing costs must be written off for some
number of casks. For purposes of this analysis, the costs will
be increased by about 20% to account for these factors. Using
these figures, a straight linear relationship can be developed
on a per-shipment basis for cost vs dose reductions for spent
fuel shipments. However, the usefulness of this relationship is
doubtful as discussed above. The same type of analysis applies
to the shipment of high-level solidified waste from the chemical
plant to the disposal site.
The shipment of low- and intermediate-level radioactive wastes
from reactors to waste disposal sites contributes about 40% of
the total impact from transportation (see table 6-2). It appears
this exposure can be reduced through the addition of shielding
to the transportation vehicle or by segregation of this inter-
mediate-level waste from the low-level waste and the addition of
shielding to the intermediate-level shipping container and/or
vehicle. About 80% of the radioactivity in these wastes is in
about 3% of the volume (2). Thus, segregation with selective
shielding appears particularly attractive.
The cost of providing shielding for a truck trailer to re-
duce the exposure rate by a factor of two is estimated as fol-
lows: About 1.26 cm of lead would be placed on the sides, ends,
and bottom of the trailer. The added weight of the lead would
-------
149
be about 18,000 kg, and the cost would be about $7,000. The total
cost, including installation and protective steel sheet, would
probably be about $20,000. However, this amount of lead would
reduce the payload by a factor of at least two and thus the re-
duction in exposure achieved through use of shielding and the
smaller quantity of material per shipment would be offset by
more numerous shipments. If segregation of the wastes was re-
quired and one section of the truck shielded, it is estimated
that about 5,500 kg of lead would be required at a cost of about
$2,000, with a total cost of about $7,000. Such shielding would
also reduce the payload, but would provide a dose reduction of
about 2 for 80% of the waste for a total dose reduction factor
of about 1.7. However, there would be additional costs involved
in segregating the wastes. It is estimated that this cost would
not exceed one man-year per reactor per year resulting in an an-
nual cost of about $20,000. The results of these estimates are
shown in figure 6-2.
Route control appears more attractive for the reduction of
exposure than the addition of shielding. Figure 6-3 depicts the
dose reduction which can be achieved through selection of routes
with lower population densities. The population dose due to
stops along the route would probably be lower for a lower popu-
lation density. However, since this factor is not well known in
any case, the same value was used as in, section 6.4 as a lower
limit of exposure.
-------
150
0.4
0.2
10
DOLLARS
3x10
LOW LEVEL WASTE
FRACTION-40%
1
1
1
1
1
1 1 1
HIGH LEVEL
& SPENT
FRACTION-
WA!
FUE
-60
1
"/
fa
5x10
Figure 6-2 Cost effectiveness for segregation of low level wastes
with an additional shielding compartment on the truck
(1 truck per reactor year)
0.8
0.6
0.4
0.2
50 100 127.4 150
POPULATION DENSITY (PEOPLE/SQUARE KILOMETER)
Figure 6-3 Impact reduction available with route control through various areas
of population density (impact from stops remains constant)
-------
151
The costs of route control would be highly variable because
of its administrative nature. However, it is estimated that one-
half man-year per reactor per year would suffice in accounting
for this cost. Additional costs would probably involve the total
distance traveled by the vehicles and would be highly site depen-
dent. For purposes of this review a cost of $20,000 per reactor
per year would appear adequate for route control requirements.
-------
A-l
Appendix A
Exposure Pathway, Radiation Dose and Health Effect Paraneters
1.0 Introduction
This appendix explains the assumptions and parameters used
in (1) calculating how radioactive effluents discharged from
fuel supply operations are transported through the environment,
(2) calculating the doses resulting fron the interaction of rad-
iation with man, and (3) converting radiation doses to health
effects.
The radioactive effluents include uranium and associated
daughter products both as aerosols and soluble liquid effluents.
Source terms were derived from operations considered reasonable
for a "model facility." Generalized "model exposure pathways"
were used to relate radioactive airborne and waterborne effluent
concentrations to radiation doses to humans. These pathways do
not apply to any specific site but were chcsen to be representa-
tive of what may take place around a "model facility." Dose con-
version factors were derived primarily from "ICRP Publication 2"
(1) and "Ionizing Radiation: Levels and Effects" (a report of
the United Nations' Scientific Committee on the Effects of Atomic
Radiation)(2). Conversion factors to express radiation doses as
resultant health effects were derived from the BE1R Report (_3).
The term "uraniun" as used in the Fuel Supply Technical Re-
port is defined as the sum of uranium-234, uranium-235, and ura-
nium-238. It does not include any other radionuclides nor does
-------
A-2
it define any particular isotopic ratio. The terms "uranium-
238," "uranium-235," "uranium-234," and "thorium-230" are defined
as referring to the named isotope only. Daughter products and
the effects of daughter products are not included. In contrast,
the term "radium-226" may include varying amounts of short-lived
daughters depending upon the exact situation.
2.0 Model Exposure Pathways
2.1 Air Pathway
The following assumptions apply to the respective terms used
in the atmospheric pathway model:
(x/Q)max is the meteorological dispersion factor that relates
the plant source term to the dispersed radionuclide concentrations
in air at a given downwind distance, x is the yearly average
-^
* concentration of activity to which a human at the specified down-
> 3
wind distance is exposed (pCi/m ). Q is the rate at which an
> airborne radionuclide leaves the site (pCi/s).
>* ^ It was assumed that the plant boundary lies between 0.5 and
£ 1.5 kilometers from the point of release. x/Q was then estimated
^
(4^ assuming a ten meter stack for three types of sites at 0.5,
1.0, 1.5 kilometers (table A-l). (x/Q) was taken to be the
max
average of these values multiplied by 2 to obtain the maximum
<>r concentration. The standard deviation was roughly esti-
mated aX^/6/the range of these values.f
This value for (x/Q)
niflx
6 ± 4 x 10 was used for all fuel supply facilities.
C(a) is the ratio of the average individual radiation dose
1
-------
A-3
Table A-l
Average (x/Q) values vs distance for h = 10 meters9 (4)
Distance (km)
Site
River
Lakeshore
Seashore
Average
9 x
4 x
6 x
6 x
0
10
10
10
.5
-6
-6
-6
io-6
3 x
1 x
2 x
2 x
1.0
10-6
lO'6
1(T6
ID'6
2
5
1
1
1.
x
x
X
X
5
io-6
ID'7
io-6
io-6
7
2
5
5
3.
x
x
X
X
0
ID'7
io-7
10'7
ID"7
For the maximum sector concentration, multiply these average values by 2.
ah = height of release of gaseous effluent in meters,
-------
A-4
within 80 kilometers of the plant to the maximum individual dose
c
at the plant boundary, fCalculating C(a) for 50 reactor sites
gives a value of 2.3 x 10~ . This value was accepted for fuel
supply facilities.
The radial distance of 80 kilometers was chosen to define
the members of the general population exposed to a specific
source term because the distance is large enough to include
nearby large population centers yet small enough so that the
area effected can be considered a local area. Beyond this dis-
tance, the diffusion equations that characterise the source term
plume are not considered reliable./ No correction for parti-
culate depletion from the plume was made.
»•—^—•—'
2.2 Water Pathway
The standard river model does not represent a specific site
but was constructed from parameters that are believed to be rea-
sonable and credible. Because most fuel supply facilities are
located on rivers, no attempt was made to construct sea coast or
lake water exposure pathways. The river model assumes:
1. A river flowing past the outfall at a rate of 280 cubic
meters per second.
2. The maximum exposed individuals are those assumed to
drink water continuously from the river following complete di-
lution of effluents by the river water.
3. The average exposed individuals are represented by X
people per kilometer of river for 300 kilometers downstream from
-------
A-5
the outfall that drink water from the river.
(x/Q)rlver is the dispersion factor that relates the plant
source term to the dispersed radionuclide concentration in water.
X is the yearly average concentration of activity (pCi/1) to
which a human is exposed by drinking water from the river
following complete dilution of effluents by the river water. Q
is the rate that a waterborne radionuclide leaves the site (pCi/s),
C(w), an additional dilution factor of 0.1, is assumed for
members of the general population exposed to the effluents to
accommodate additional river dilutions from tributaries, loss of
activity to surfaces of the river, multiple uses of water, etc.
The dispersion factors discussed above are summarized in
table A-2.
3.0 The "Model" Population at Risk
Table A-3 summarizes the various "model" population groups
used for computing somatic and genetic dose effects.
3.1 Through the Air Pathway '
The total population within 80 km of 50 reactor sites, taken
primarily from environmental impact statements, indicates that
there is an average population of 1.5 x 106 people as projected
to 1980. The doubling time for this population is assumed to be
40 years based upon an annual increase in population of about
2.5%. Fuel supply facilities are assumed to be similar except
for mills which are located in sparsely populated western
states. For mills it is assumed that 5.5 x 104 persons are
-------
A-6
Table A-2
Dispersion factors for airborne and waterborne pathways
^x'^'maximum
Pathway sector
Airborne 6 ± 4 x 10" pCi/m3 C(a) = 2 3 x
pCi/s
Waterborne 4 x 10"6 pCi/m3 C(w) = 0.1Oc
pCi/s
aRadius of 80 km
bRiver flow - 280 m3/s; for mills in western states where rivers are smaller,
assume a river flow of 14 m3/s and multiply this factor by 20.
cApparent length of river - 300 km
-------
A-7
Table A-3
Population models for air and water pathways
Pathway Area of Concern
Air
Within 80 km
Water Within 300 km
Population
Biological
effects
Within 80 km of mills
1.5 x 106 persons Somatic
0.8 x 1Q6 Genetic
.
5.5 x 104 persons Somatic
1.3 x TO* Genetic
0.6 x 106 persons Somatic
0.3 x 106 persons Genetic
For mills, within 300 km of outfall
4.4 x 104 persons Somatic
2.2 x 10 persons Genetic
-------
A-8
within 80 kin of the site. Only one-half of these populations
are used for genetically significant dose calculations.
3.2 Through the Water Pathway
It is estimated that 2,000 persons/km of river for 300 km
downstream from the outfall drink water taken from the river.
This is a total of 0.6 x 10 persons (1980). The number was
obtained by dividing the number of people in the various water-
sheds by the length of rivers over 600 miles long. The popula-
tion-at-risk for mills was reduced by a number proportional to
the population density ratios, 0.037 and multiplied by 2 because
it may be assumed that in arid territory, the population is more
concentrated around water. The resulting population at risk is
44,000. Doubling time of the population is assumed to be 40
years. Only half of the population is used for genetically
significant dose calculations.
4.0 Radiation Dose Conversion Factors
4.1 Airborne Radionuclides
The dose conversion factors for insoluble (class Y) alpha-
emitting aerosols given in table A-4 were estimated from the
dose conversion factor for an insoluble aerosol of plutonium-239
oxide by a consideration of the ratio of the alpha energies of
the radionuclides to the alpha energy of plutonium-239 (table
•MM
A-5)i.S The continuous inhalation of 1 pCi/m of insoluble pluto^"
jiium oxide is taken to deliver 12 rem/yr to the lung. \This
value is based on the ICRP Report of Committee II of 1959 plus a
-------
A-9
Table A-4
Dose conversion factors for the airborne pathway
Radionucl ide
Plutoniu;:--239
Urani'j.Ti-233
Uranium-234
Uranium- 230
Radium-226
Radon-222
Aerosol Class
Y
Y
D
Y
D
Y
Y
D
-
Organ at
Risk
Lung
Lung
Bone
Lung
Bone
Lung
Lung3
Bone
Lung (T.B.)b
Dose conversion factor
mrem/yr per
pCi/m3
12,000.
10,000
150
10,000
150
11,000
11,000
300
4
'Assume radon-222 and therefore all radium-226 daughters escape from the aerosol
particle and only radium-226 (i.e., i EF(RBE) n = 49) contributed the dose
*T.B. means tracheobronchial region.
-------
A-10
Table A-5
Dose conversion factors for airborne insoluble participates
Radionuclide
Plutonium-239
Uranium- 238
Uranium- 235
Uranium-234
Thorium -230
Radium -226
I E (RBE)na
(MeV)
53
46
46
46
48
49b
Dose conversion factor
(mrem/yr per pCi/m3)
12,000
10,000
10,000
10,000
11,000
11,000
a
ICRP report of Committee II - assume uranium-238 and uranium-234
equal in energy to uranium-235
Assume all radon escapes from the aerosol particle
-------
A-ll
correction factor of 8 (5) Jo convert from the 1959 lung model
to the newer lung model recommended by the ICRP Task Force report,
"Deposition and Retention Models for Internal Doslmetry of the
Human Respiratory Tract" (6).
Dose conversion factors for soluble aerosols were calculated
from ICRP Publication 2 values using bone as the critical organ
(table A-6).
It was recognized that doses calculated for the lymph nodes
in the tracheobronchial region of the lung are much higher (by a
factor of 35) than the lung dose if the ICBP "New Lung Model" is
used. However, the ICRP has recommended that lymphatic tissue
not be considered the critical organ in inhalation exposure to
Plutonium (7). This apparently is because at present there are
no reports of primary tumors of the tracheobronchial lymph nodes
and because it is believed that protection of the more radiosen-
sitive lung tissue will provide more than adequate protection to
the lymph nodes. Radiation dose to the lymph nodes of the tracheo-
bronchial region will not be used as a criterion for setting envi-
ronmental standards at this time; and therefore, this dose has
not been included in tables A-10 and A-ll.
4.2 Waterborne Radionuclides
The uranium dose conversion factors (table A-7) for the water-
borne pathway were taken from dose calculations endorsed by the
UNSCEAR report (2) for radiation dose rates that result to soft
tissue (gonads) and bone surfaces, the tissue at risk for cancer
-------
A-12
Table A-6
Dose conversion factors for airborne soluble participates
Radionuclide
Uranium-238
Uranium-235
Uranium-234
Thorium- 2 30
Radium-226
Plutonium
168 h week
(MPC) a
(yCi/cm3)
2xlO-10
2 x lO'10
2 x 10-10
8 x 10"13
1 x ID'10
6 x 10"13
Critical
organ
Bone
Bone
Bone
Bone
Bone
Bone
Dose conversion factor
(mrem/yr per pCi/nr)
150
150
150
38 ,000
300
50,000
5ICRP Report of Committee No. II - (MPC) is equivalent to
30 rem/yr to bone.
-------
A-13
Table A-7
Dose conversion factors for the water pathway
Radionuclide
Uran1um-238
Uranium- 234
Thorium- 230
Radium- 226
Organ at
risk
Bone
Soft tissue
Marrow
Bone
Soft tissue
Marrow
Bone
Soft tissue
Marrow
Bone
Soft tissue
Harrow
Dose conversion factor
(mrem/yr per
pC1/1lter)»»t>
9
0.9
2
9
0.9
2
1
0.2
12
0.4
2.4
alt is assumed that adults consume 2 liters/day of water
and that the listed dose rates will result if activity
in the water 1s 1 pCi/liter.
bListed are equilibrium dose rates that result from
equilibrium body burdens. For example, radon-226 dose
rate is not assumed to be reached until 16 years following
start of exposure.
-------
A-14
induction, following the ingestion of known amounts of uranium
naturally present in the human diet (table A-8). While it is
recognized that uptake following ingestion in food may not neces-
sarily lead to the same rate of uptake as the ingestion of
uranium in drinking water, it is believed that this is the best
model currently available for long-term chronic ingestion of
uranium. Note that these dose conversion factors refer only to
the uranium radionuclide listed and do not include daughter
product radiations. These must be listed and calculated separ-
ately. However, because the factors are the same for uranium-
234, uranium-235, and uranium-238 and the source terms are
listed as total uranium (i.e., the sum of uranium-234, ura-
nium-235, and uranium-238), a single calculation suffices.
The thorium-230 dose conversion factor was calculated from
ICRP Publication 2 and corrected for surface dose; the radium-
226 dose conversion factor was calculated in the same fashion as
uranium (tables A-9 and A-10).
The dose to the bone marrow was calculated separately (2).
For this site, cancer risks are calculated using the health
effect factor for leukemia.
5.0 Health Effect Conversion Factors
Health effect conversion factors were abstracted from the
BEIR report and rescaled for the dose to bone surfaces. They
are listed in table A-ll.
The total integrated radiation dose is multiplied by these
-------
A-15
Table A-8
Dose conversion factors for uranium for the water pathway (2J
Intake 1 yg/day natural uranium „,,
0.68 pCi/day total uranium (238U, Z35U, U)
Equilibrium body 0.1 to 0.9 ng/g soft tissue
burdens 20 to 30 ng/g bone ash
Dose delivered (calculated by the method of Spiers)
Total uranium at 9^2 pCi/kg bone yields:
0.3 mrad/yr to bone (tabecular bone, surfaces)
0.06 mrad/yr to bone marrow
0.03 mrad/yr to soft tissue (gonads)
Assuming a quality factor of 10, 0.68 pCi total uranium intake per day yields
3 mrem/yr to bone (tabecular bone, surfaces)
0.6 mrem/yr to bone marrow
0.3 mrem/yr to soft tissue (gonads)
For an intake of 2 liters of liquids per day containing 1 pCi/1 total
uranium, yields:
8.8 mrem/yr to bone (tabecular bone, surfaces)
1.8 mrem/yr to bone marrow
0.9 mrem/yr to soft tissue (gonads)
-------
A-16
Table A-9
Dose conversion factors for radium-226 for the water pathway (2)
Intake 1 pCi/day radium-226
Body burden 7.6 pd'/kg bone
40 pCi/skeleton
Dose delivered (calculated by the method of Spiers)
0.6 mrad/yr to bone (tabecular bone, surfaces)
0.1 mrad/yr to bone marrow
0.02 mrad/yr to soft tissue (gonads)
Assuming a quality factor of 10, 1 pCi 226Ra/day intake yields:
6 mrem/yr to bone (tabecular bone, surfaces)
1 mrem/yr to bone marrow
0.2 mrem/yr to soft tissue (gonads)
An intake of 2 liters of liquids per day containing 1 pCi 226Ra/l
yields:
12 mrem/yr to bone (tabecular bone, surfaces)
2.4 mrem/yr to bone marrow
0.4 mrem/yr to soft tissue (gonads)
-------
A-17
Table A-10
Dose conversion factors for the water pathway for soluble radionuclides
calculated from the recommendations of the ICRP
Radionuclide
Thorium-230
168 h week
(MPC)W
(yCi/cm3)
2 x ID"5
Critical
organ
Bone
Bone surfaces
Marrow
Dose conversion factor
(mrem/yr per pCi/1)
2.5
1.0a
0.2a
Ratio of bone dose to bone surface dose and bone marrow dose assumed to be the
same as for uranium.
-------
Table A-ll
Health effects conversion factors
Critical
organ
Lung3
Boneb
Bone marrowb
Total soft
tissue
organs
other than
bone
Health
effect
Cancer of lung
Cancer of skeleton
Leukemia
Cancers and
genetic effects
Mortal i
50
16
54a
150
Health effects conversion factor
Events per 10b rem aggregate dose
ty Nonfatal cancers Genetic effects
0 0
16 o
0 0
150 300
Insoluble aerosol exposure pathway
blngestion exposure pathway
00
-------
A-19
factors to give the total health effects that are expected
(committed) as the result of the radiation exposure. These
effects occur over a period of years following exposure.
For uranium ingested through the water pathway, the majority
of this uranium body burden is located in the bone. This ura-
nium is considered to irradiate the bone with the corresponding
health effect of cancer of the skeleton. In addition, 20% of
the dose is assumed to irradiate the bone marrow with the re-
sulting health effect being leukemia. The mortality risk from
leukemia is 54 cases per 10 rems to the bone marrow. The risk
of mortality from leukemia from bone irradiation by alpha-emit-
ting radionuclides in the bone is therefore 0.2 x 54 cases per
10 rems.
For soft tissue organs, including the gonads, uranium bur-
dens are considered to provide an average organ dose with the
corresponding health effects assumed to be the sum of all ef-
fects on each individual soft tissue organ. In addition, this
dose is delivered to the gonads causing genetic health effects.
6.0 Radiological Effects Calculations
Given the source term and the factors discussed above, the
various radiation doses to individuals in the general popula-
tion, and the expected committed health effects resulting from
these doses can be calculated. These calculations are presented
in tables A-12, A-13, and A-14. A simplified flow sheet for
radiation dose and health effects calculations is given in table
-------
Table A-12
Airborne pathway dose calculations for node! facilities - current best technology
Model
facility
Mill
Conversion
(wet solvent
extraction)
Conversion
(hydrofluor)
Enrichment
Fabrication
Source /./m
Radio- term 'x/Q/max Critical
nuclide (Ci/vr) (s/n>3) Orqan
U total 0.
226Ra 0.
23°Th 0.
U total 0.
insoluble
U total 0.
soluble
U total 0.
insoluble
U total 0.
soluble
U total 0,
soluble
U total 0,
insoluble
a Ci 1 yr _ „ lo12 pi
yr x 3.15.
mrem/yr „
x 10' " Ci
1
06
06
015
008
,038
,019
,045
.005
— x
10"3 rem .. persons
6 x 10"6 Lung
(To. 1904)) Lung
Lung
*?
Lung
Bone
Lung
Bone
Bone
Lung
Dose
conversion
factor Maximum exposure Average exposure Aggregate
/mrem/yr-, at boundary dose within 80 km Persons Somatic Organ Dose
Sci/rrv* ' (mrem/yr) C(a) (mrem/vr) Exposed (rem-per yr/facility-yr)
1
1
1
1
1
1
1
1
1
.0 x
.1 x
.1 x
.0 x
.5 x
.0 x
.5 x
.5 x
.0 x
104
104
104
Total
104
10?
104
102
102
104
1.9
1.3
1.3
4.5
2.9
2.3
7.2
5.4
1.3
9.5
x 102 2.3 x lO"4
x 102
x 102
x 102 Total
x 101
x 10'1
x 101
x 10'1
x 10°
x 10°
4.4 x 10-2
2.9 x lO-2
2.9 x 10'2
10.2 x 10"2
6.6 x ID"3
5.3 x 10~5
1.7 x 10"2
1.2 x 10'4
3.0 x 10'4
2.2 x 1Q'3
5.5 x 104 2.4 x 100
1.6 x 10°
1.6 x 10°
Total 5.6 x 10°
1.5 x 105 9.9 x 10°
7.9 x 10"2
1.5 x 106 2.5 x 101
1.9 x 10'1
1.5 x Id6 4.4 x 10"1
1.5 x 106 3.3 x 10°
pCi/m3 x mrem/yr = mrem/yr;
pCi/s pCi/m3
x facility-yr = rem/yr
aggregate dose
per facility-yr
-------
Table A-13
Waterborne pathway dose calculations for model facilities - current best technology
(x/Q)R Dose conver-
Source (B£j/l) sion ^actor Maximum exposure
Model Radio- terra pCTTs" Critical ,mrem/yr. dose
facility nuclide (Ci/yr) (0.127) organ l pCi/1 ; (mrem/yr) C(w)
Mill U total 0.1
226Ra 0.06
230Th 3.5
Conversion U total 2
(wet solvent)
226Ra 0.006
230Th 0.0006
Conversion U total 0.8
(hydrofluor)
Enrichment U total 0.6
Fabrication U total 0.5
1 x 10'6 Bone 9
x 20a Soft
tissue 0.9
Bone 12
Soft
tissue 0.4
Bone 1
Soft
tissue
4 x 10'6 Bone 9
Soft
tissue 0.9
Bone 1 2
Soft
tissue 0.4
Bone 1
4 x 10"6 Bone 9
Soft
tissue 0.9
4 x 10-6 Bone 9
Soft
tissue 0.9
4 x 10'6 Bone 9
Soft
tissue 0.9
2.2 x 10° 0.10
2.2 x 10"1
1.8 x 10°
6.2 x 10'2
8.9 x 10°
-
2.3 x 10°
2.3 x 10"1
9.1 x 10"3
3.1 x 10"4
7.6 x 10"5
9.1 x 10'1
9.1 x 10"2
6.9 x 10-'
6.9 x ID'2
5.7 x 10''
5.7 x ID'2
Average exposure Aggregate
dose Persons somatic organ dose
(mrem/yr) exposed (rem per yr/facility-yr)
2.2 x 10"1 4.4 x 104
2.2 x 10"2
1.8 x 10'1
6.1 x 10" 3
8.9 x 10"1
-
2.3 x 10"1 6.9 x 105
2.3 x 10'2
9.1 x 10'4
3.0 x 10'5
7.6 x 10"6
9.1 x 10"2 6.0 x 105
9.1 x 10'3
6.9 x 10-2 6.0 x 105
6.9 x ID"3
5.7 x 10~2 6.9 x 105
5.7 x 10"3
1.0 x 101
1.0 x 10°
8.1 x 10°
2.7 x 10'1
3.9 x 101
-
1.4 x 102
1.4 x 101
5.5 x 10'1
1.8 x 10'2
4.6 x 10'3
5.5 x 101
5.5 x 10°
4.1 x 101
4.1 x 10°
3.4 x 1C1
3.4 x 10°
Correction factor for smaller size of the western rivers - mill only
-------
Table A-14
Health effect calculations for model facilities - current best technology
Model
facility
Mill
Conversion
(wet solvents)
Conversion
(hydrofluor)
Enrichment
Fabrication
H.E.F. - healtt
Aggregate somatic
Critical dose H-E'Fi Genetic
organ (rem/facility-yr) x 10"6 correction Mortalities
Lung
Bone
Bone3
Soft
tissue
Lung
Bone
Bone3
Soft
tissue
Lung
Bone
Bonea
Soft
tissue
Bone
Bone
Bone3
Soft
tissue
Lung
Bone
Bone3
Soft
tissue
;e is 20%
5.6 x 10°
5.7 x 10'
1.3 x 10°
9.9 x 10°
1.4 x IO2
1.4 x IO1
2.5 x 10]
5.5 x 10'
5.5 x 10°
4.4 x 10"1
4.1 x 10'
4.1 x 10°
3.3 x 10°
3.4 x 101
3.4 x 10°
of bone dose; the
factor
50/0/0
16/16/0
n/o/oa
150/150/300
50/0/0
16/16/0
11/0/oa
150/150/300
50/0/0
16/16/0
11/0/03
150/150/300
27/16/0
16/16/0
ll/O/O3
150/150/300
50/0/0
16/16/0
ll/O/O3
150/150/300
H.E.F. for leukemia
2.8
9.1
6.3
-/-/0.5 2.0
Totals 20.2
5.0
22.4
15.4
-/-/0.5 21.0
Totals 63.8
12.5
8.8
6.1
-/-/Q.5 8.3
Totals 35.7
0.1
6.6
4.5
-/-/0.5 6.2
Totals 17.4
1.7
5.5
3.7
-/-/0.5 5.1
Totals 16.0
is multiplied
x TO'4
x 10-4
x ID'4
x ID'4
x 10-4
x ID'4
x TO'4
x IO-4
x IO-4
x IO-4
x 10-4
x 10-4
x ID'4
x 10-4
x IO-4
x ID'4
x TO'4
x ID"4
x 10-4
x 10-4
x ID"4
x 10-4
x TO'4
x 10-4
x 10-4
by 0.2
Nonfatal
cancers
0
9.1 x
0
2.0 x
11.1 x
0
22.4 x
0
21.0 x
43.4 x
0
8.8 x
0
8.3 x
17.1 X
7.0 x
6.6 x
0
6.2 x
12.8 x
0
5.5 x
0
5.1 x
10.6 x
to give
10-4
ID"4 2.
lO-4 2.
10-4
TO"4 21.
ID"4 21.
10-4
TO'4 8.
ID"4 8.
lO'6
10-4
10-4 6.
TO'4 6.
io-4
10-4 5.
IO-4 5.
a H.E.F.
Genetic
events
0
0
0
0 x TO'4
0 x ID"4
0
0
0
0 x 10-4
0 x ID'4
0
0
' 0
3 x 10-4
3 X ID'4
0
0
0
2 x 10-4
? * TO"4
0
0
0
1 x ID'4
1 x ID"4
for use
Pathway
Air
Water
Water
Water
Air
Water
Water
Water
Air
Water
Water
Water
Air
Water
Water
Water
Air
Water
Water
Water
with bone
Total effects per
Total effects 30 yr exposure
per exposure (effects/faci lity-
(effects/facility-yr) 30 yr)
2.8 x 10-4
3.1 x 10-3
3.4 x 10~3
5.0 x ID'4
12.3 x lO-3
12.8 x 10-3
1.3 x IO-3
4.9 x ID'3
6.1 x ID"3
0.1 x ID'4
3.6 x IO-3
3.6 x ID"3
1.7 x TO'4
3.0 x ID'3
3.2 x 10-3
dose.
8.4 x ID'3
9.2 x 10'2
10.0 x TO'2
1.5 x 10'2
3.7 x 10'1
3.9 x 10"1
3.8 x 10"2
1.5 x 10'1
1.9 x 10'1
3.0 x TO'4
1.1 x 10"1
1.1 x 10'1
5.0 x TO'3
9.0 x 1(T2
9.5 x 10"2
-------
A-23
A-15.
7.0 The 100-Year Radiation Dose Commitment
7.1 Introduction
A year's release of radioactive material from a fuel supply
facility causes an immediate commitment of radiation dose to
members of the general population in the vicinity of the faci-
lity. This dose is delivered when the radioactive effluents,
moving quickly through air and water pathways, are taken up by
individuals. This takes place during the year in which the ef-
fluent is released; and these radiation doses and the resulting
health effects have been calculated in previous sections.
The same radioactive material upon release to the biosphere
may, over longer periods of time, find its way back to man
through slow, secondary pathways such as resuspension-inhalation
and food chains. This causes additional radiation dose and
health effects. Although these dose rates are usually quite
snail, the number of people exposed may be very large so that
because of the linear, nonthreshold health effects model, the
number of predicted health effects may become significant.
The 100-year dose commitment is an attempt to calculate the
radiation dose and health effects that will result when a single
year's release of radioactive effluent interacts with man during
the following 100 years. To estimate the effect of 30 years of
facility operations, the 100-year dose commitment for each year
of operations is added together. The health effects that result
-------
A-24
Table A-15
Radiation dose and health effect calculations - flowsheet
Source term x (%-)
(PC1/S) W
Dose
x conversion
factor
x C( a or w)
x Number of
people exposed
Maximum exposure
concentration
(pCi/1 or m3)
Dose to maximum
exposed individual
(organ mrem per yr/facility-yr)
Dose to average
exposed individual
(organ mrem per yr/facility-yr)
Total integrated dose
(organ rems per yr/facility-yr)
x Health effects
factor
Total number of health effects
(effects/facility-yr)
-------
A-25
from the 100-year dose commitment are in addition to the effects
that occur from the exposure to or immediate uptake of radio-
active effluents by people during the year of release.
7.2 Source Terms
Source terms are identical to those used for each model
facility in the proceeding sections.
7.3 Exposure Pathway
Natural uranium is released from model facilities of the ura-
nium fuel supply in liquids in a soluble form and to air in both
soluble and insoluble forms. The uranium released in solution
form is washed to the ocean or is deposited in silt along the
river bed. There are presently no known or postulated exposure
pathways from uranium buildup in the ocean or silt. The assump-
tion is made that they represent an infinite sink; thus, liquid
releases do not contribute to the 100-year dose commitment.
Air releases of natural uranium, whether soluble gases or
insoluble particulates, settle out at some point after release
rather than remain airborne. The exact pattern of settling
depends on the meteorology at the time of release, the geology
around the plant, stack height, and other related parameters.
It is assumed that 20% of the release will be uniformly dis-
tributed within 80 km. of the model plant, the remaining 80% will
be uniformly distributed across the eastern half of the United
States, and that the deposited uranium will be uniformly dis-
tributed throughout the top 15 cm of soil.
-------
A-26
Uranium occurs naturally in the soil and has an average
concentration of 2.8 micrograms per gram of soil, (yg/g) (2).
The average daily intake is estimated to be 1 ug; less than 1%
of this intake is from inhalation, of resuspended particles (2).
The major pathway is ingestion through the food chains.
It may be assumed that the increase in concentration of ura-
nium in the soil due to releases from the uranium fuel supply
facilities causes a proportional increase in the average daily
Intake of uranium. Since the radiation dose from background
levels of uranium has been calculated (2) , the additional dose
due to the released uranium can also be calculated.
7.4 Exposure Term
One curie of natural uranium is equivalent to 1.48 x 10
14
grams. The land area within 80 km of a plant is 2.01 x 10
2
cm , and of the eastern half of the United States, 3.88 x
16 2
10 cm . Assuming uniform mixing within the top 15 cm of soil
(2) and a soil density of 1.5 g/cm , a one curie release would
result in soil concentrations above background of 6.54 x 10~~
and 1.35 x 10" vg/g for the local and eastern U.S. land areas,
respectively.
Table A-16 gives the background dose to critical organs from
naturally occurring uranium and the additional calculated doses
due to deposition of a one curie release from a model plant.
The additional doses were determined by multiplying the doses
from natural uranium by the ratios (6.54 x 10~ pg/g)7(2.8 yg/g)
-------
A-27
Table A-16
Doses to critical organs from naturally occurring uranium and an air
release of one curie of uranium from a model plant
Critical
organ
Soft tissue
Gonads
Endosteal bone
Bone marrow
Background
dose
(mrad/yr)
0.03
0.03
3
0.06
Additional dose
Local
(mrad/yr)
7.0 x 10~7
7.0 x 10'7
7.0 x ID'6
1.4 x 10'5
from a 1 Ci re1e:.:e
Eastern
United States
(mrad/vr)
1.4 x 10'8
1.4 x 10'8
1.4 x 10"7
2.9 x 10'8
-------
A-28
and (1.35 x 10~ yg/g)/(2.8 yg/g). To convert each critical
organ dose to a dose equivalent (mrads to mrem) it is necessary
to multiply the doses by a quality factor. For alpha emitters,
the quality factor is defined as 10.
Estimates of the number of health effects expected for a
given population dose are based on data presented in the BEIR
report (3). Health effects will be defined as the summation of
the number of lethal cancers, nonlethal cancers, and, if appli-
cable, genetic effects. The health effects conversion factors
for the critical organs are:
Soft tissues 300 health effects per 10fi person-rem
Gonads 150 health effects per 10,. person-rent
Endosteal bone 32 health effects per 10, person-rem
Bone marrow 54 health effects per 10 person-rem
The probability that an average individual in the population
will incur a health effect due to a release of uranium is cal-
culated by multiplying the health effects conversion factors by
the respective critical organ dose equivalents and summing.
This summation results in a single conversion factor relating
the anticipated probability of a health effect occurring fol-
lowing the release of one curie of uranium from a model facility.
The conversion factors for local and eastern U.S. populations
-12 —13
are 5.5 x 10 and 1.2 x 10 health effects per person per
curie released, respectively.
7.5 fopulation Term
The population of the United States is projected to grow
-------
A-29
linearly from 205 million in 1970 to 300 million in 2020.
Beyond 2020 it is assumed to remain constant at 300 million
persons. The population within 80 kia of a model facility is
assumed to be 0.7% of the total U.S. population and is assumed
to grow as the U.S. population grows. An exception is the
population around a model mill. It is assumed to remain con-
stant at 54,000 people. The population of the eastern half of
the U.S. is assumed to be 80% of the total U.S. population. It
will also grow as the U.S. population grows.
7.6 Evaluation of the 100-Year Dose Commitment
Table A-17 gives the health effects expected to result from
the 100-year dose commitment for 30 years of facility operations.
For comparison, the health effects expected to result from the
immediate dose commitment for 30 years of facility operations is
included.
For all types of fuel supply facilities, the 100-year dose
commitment causes less than 10% additional health effects when
compared to the immediate dose commitment. These calculations
refer to uranium discharges only.
-------
Table A-17
Health effects resulting from the 100-year dose commitment from uranium
- 30 years of facility operations -
Model
facility
Mill
Conversion
(wet solvent)
(Hydrofluor)
Enrichment
Fabrication
Airborne pathway
source term
(Ci/yr per facility)
0.1
0.023
0.057
0.045
0.005
Health effects from
100-yr dose commitment
(effects ^er facility-30_yr)
0.006 (10%)a
0.002 (0.5%)a
0.005 (3%)a
0.004 (4%)a
0.0005 (0.5%)a
Health effects from
immediate dose
commitment
(effects per facility-30
0.05a
0.4
0.2
0.1
0.1
yr)
00
o
aPercent of total health effects.
bHealth effects from uranium immediate dose commitment.
-------
B-l
Appendix B
Costs of Control Technology
Capital and operating costs of control technology were obtained
from literature sources where available. If not available, an esti-
mate was made based on the judgment of the factors involved in oper-
ation of the control systems for a different purpose. An example
would be waste settling ponds. Estimates of the annualized costs
(combined yearly capital and operating costs for accounting purposes)
were made by multiplying the capital cost of a control system by an
annual fixed charge rate. The fixed charge rate includes deprecia-
tion, interest, taxes, and insurance for a sinking method of depre-
ciation based on a 30-year plant life (1). The fixed charge rate used
was 16.6% per year (2). The product of the annual fixed charge rate
and the capital cost yields the annualized capital cost which is then
added to the annual operating costs. This sum is an estimate of the
total annualized costs for the control system. The present worth of a
treatment system was calculated from the annualized cost (equivalent
to the present worth of the annualized costs using a 7.5% interest
rate). Present worth is any future payment or series of payments that
will repay a present sum with interest at a given rate. The present
worth factor (pwf) for 7.5% for 30 years is tabulated as 11.81 (1).
The present worth cost of the system is the equivalent to 11.81 times
the total annualized costs.
-------
R-l
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R-2
11. KERR-MCGEE CORPORATION. Uranium Rexafluoride Plant, Sup-
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-------
R-3
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Section 3
1. U.S. ENVIRONMENTAL PROTECTION AGENCY. Fuel Cycles for Elec-
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-------
R-4
4. KERR-MCGEE CORPORATION. Uranium Hexafluoride Plant, Sup-
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Section 4
1. U.S. ATOMIC ENERGY COMMISSION. Environmental Survey of the
Nuclear Fuel Cycle. Directorate of Licensing, Fuels, and
Materials. U.S. Atomic Energy Commission, Washington, D.C.
20545 (November 1972).
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-------
R-5
Section 5
1. U.S. ATOMIC ENERGY COMMISSION. Environmental Survey of the
Nuclear Fuel Cycle. Directorate of Licensing, Fuels, and
Materials, U.S. Atomic Energy Commission, Washington, D. C.
20545 (November 1972).
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1257 (September 1970).
JERSEY NUCLEAR COMPANY. Applicant's Supplemental Environ-
mental Report, Uranium Oxide Fuel Plant. No. JN-14ADD1.
USAEC Docket No. 70-1257 (October 1971).
3. KERR-MCGEE CORPORATION. USAEC Docket No. 70-1113, Letter
dated October 11, 1971.
4. NUMEC, USAEC Docket No. 70-135, Letter dated April 13, 1972.
5. GENERAL ELECTRIC CO., USAEC Docket No. 70-1113, Letter dated
November 29, 1971.
6. GULF UNITED NUCLEAR FUELS CORPORATION, USAEC Docket No. 70-
36, Letter dated May 3, 1972.
7. HITTMAN ASSOCIATES. Radioactive Waste Management - A Survey
for: U.S. Environmental Protection Agency, Office of Radia-
tion Programs, Contract 68-04-0052, HIT-516 (May 1972).
8. HETLAND, N. and J. C. RUSSELL, Jr. Adequacy of Ventilation
Exhaust Filtering System for New Plutonium Facilities. Paper
presented at 12th AEC Air Cleaning Conference (1972).
9. LEONARD, H. H., T. S. BAER and L. E. EKART. Techniques for
Reducing Routine Release of Radionuclides from Nuclear Power
Plants for U.S. Public Health Service, Bureau of Radiologi-
cal Health, Contract No. CPE-R-70-0015, p. 41 (January 1971).
10. BATTELLE, PACIFIC NORTHWEST LABORATORY. Data for Preliminary
Demonstration Phase of the Environmental Quality Information
and Planning System (EQUIPS) for U.S. Atomic Energy Commis-
sion BNWL-B-141. (December 1971).
11. Public Law 92-500, 92nd Congress, S.2770, An Act to Amend
the Federal Pollution Control Act, (October 1972).
12. KEMMER, F. N. and J. H. BEADSLEY. Chemical Treatment of Waste
from Mines and Mineral Processing. Engineering and Mine J.
172. No. 4 (April 1971).
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13. RYAN, E. S., J. N. VANCE, and M. E. MAAS. Aqueous Radio-
active-waste-treatment plant at Rocky Flats. Proceedings.
Symposium on Practices in the treatment of low and inter-
mediate level radioactive wastes. Jointly organized by JAEA
and European Nuclear Energy Agency, Vienna, Austria (Decem-
ber 6-10, 1965).
Appendix A
1. INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION. Report
of Committee II on Permissible Dose for Internal Radiation
Pergamon Press, New York, New York (1959).
2. UNITED NATIONS. Ionizing Radiation: Levels and Effects, A
Report of the United Nations Scientific Committee on the
I no7->?f At0mic Radiation to the General Assembly, Volume
-*- • \^y / £) •
3. NATIONAL ACADEMY OF SCIENCES. The Effects on Populations of
Exposure to Low Levels of Ionizing Radiation, Report of the
Advisory Committee on the Biological Effects of Ionizing Rad-
iations (BEIR), National Academy of Sciences (November 1972).
4* "l?I iT°MIC !NERGY COMMISSION. Final Environmental State-
proposed Rule Making Action. Volumes I and
D.C.
5' Protf ;
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2. U.S. ATOMIC ENERGY COMMISSION. Draft Environmental State-
ment Concerning Proposed Rule Making Action: Numerical
Guides for Design Objectives and Limiting Conditions for
Operation to Meet the Criterion "As Low as Practicable" for
Radioactive Material in Light-Water-Cooled Nuclear Power
Reactor Effluents. Prepared by the Directorate of Regu-
latory Standards, U.S. Atomic Energy Commission, Appendix A,
p. 162 (January 1973).
*US. GOVERNMENT PRINTING OFFICE: 1973 546-311/112 1-3
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