ENVIRONMENTAL ANALYSIS
OF THE URANIUM FUEL CYCLE
PART IV - Supplementary Analysis: 1976
July 1976
US.. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
Washington, D.C. 20460
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PKEFACE
In 1973 the Office of Radiation Programs issued an environmental
analysis of the uranium fuel cycle, which was isaued in three volumes
covering fuel supply, power reactors, and fuel reprocessing. Sub-
sequent to the issuance of this analysis, the Agency proposed
environmental radiation protection standards on May 29, 1975, for
nuclear power operations of the .uranium fuel cycle (40 CFR Part 190).
The Agency held public hearings on these,proposed standards, in
Washington, B.C., on March 8 - 10, 1976. As a result of the ensuing
comments, a number of areas were identified in which the development
of additional information was necessary.
It is the objective of this! new Part IV, entitled "Supplementary
Analysis - 1976," to address several technical areas in which new
information is available or which were discussed only briefly in
previous reports. In the former category are sections pertaining to
uranium milling and fuel reprocessing, while items such =as transuranic
effluents from recycled uranium and nitrogen-16 skyshine at BWRs fall
into the second category. Finally, Part IV, replaces and updates the
technical discussions presented in the January 5, 1976, Supplementary
Information document.
As in the original reports, the principal purposes of these
analyses are to project the impact on man of the environmental releases
of radioactive materials from the fuel cycle, and to assess the capa-
bilities and costs of controls available to manage environmental
releases of these materials.
Comments on this analysis would be appreciated. These should be
sent to the Director, Technology Assessment Division (AW-459), Office
of Radiation Programs.
W. D.'Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs (AW-458)
111
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CONTENTS
I. Fuel Supply i ' Page
A. Environmental Analysis of the Uranium l*uel Cycle,
Part I (Fuel Supply): Uranium Milling 1
1.0 Introduction ..;. 1
2.0 General description, 3
3.0 Releases of radioactive effluent.from uranium milling
operations 7
3,1 Airborne releases from the mill 7
3.2 Waterborne releases from the mill 9
3.3 Airborne and wa"terborne releases from the mill
tailings pond .,., •«.«.•• •>•• ••.••••••..«•• 12
4.0 The model uranium mill 17
! • • * -.•••:.. M> , .-.-.:•. ' *: •.-..-. •'
5.0 Radioactive effluents from a model uranium mill 19
6.0 Radiological.impact pf a model mill ,. 23
7.0 Health effects impact of a model mill 26
8.0 Control technology fbr uranium milling 27
8.1 Airborne effluept control technology 27
8.2 Waterborne effluent control,technology and solid
waste control technology .,...' 30
9.0 Effluent control technology for the model mill 34
10.0 Retrofitting control.technology to operating uranium
mills 36
10.1 Retrofitting control measures to operatipnal
tailings ponds , 3$
References ..i...'.'..'.".. .T.......' 44
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CONTENTS (continued)
B. Transuranium Effluents from Re-enriching or
Refabricating Reprocessed Uranium 47
1.0 Introduction ..... 47
2.0 Gaseous diffusion operating experience 51
3.0 Estimated radioactivity releases 54
References 58
II. Nuclear Power Reactors
A. An Analysis of Control Options for N-16 Offsite Skyshine
Doses at Boiling Water Reactors 59
1.0 Introduction „ .... 59
2.0 Sources 60
3.0 Turbine building configuration 62
4.0 Dose assessment 64
•>,
5.0 Shielding of components 65
References , 83
III. Nuclear Fuel Reprocessing
A. Control of Iodine Discharges from Nuclear Fuel
Reprocessing Facilities 84
1.0 Introduction 84
2.0 Source terms for iodine 85
3.0 Control technologies for iodine at reprocessing
plants 87
vi
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CONTENTS (.continued)
3.1 Caustic scrubbers 89
3.2 Mercuric nitrate scrubbers 90
3.3 Silver zeolite adsorbers 90
3.4 Maeroreticular resins. .. 91
3.5 Suppression in evaporator by mercuric nitrate ... 91
3.6 Advanced systems -, 92
4.0 Cost evaluations ,, , _ 94-
5.0 Doses and potential health impact attributable to
iodine discharges from fuel reprocessing 95
*
6.0 Cost—effectiveness considerations 97
References 106
B. Control of Krypton Discharges from Nuclear.Fuel
Reprocessing Facilities 109
1.0 Introduction 109
2.0 Source terms for krypton ._, 110
3.0 Control technologies for krypton at reprocessing
plants Ill
3.1 Cryogenic distillation *. Ill
3.2 Selective absorption 113
4.0 Cost of krypton control at fuel reprocessing
plants ...115
4.1 Direct costs 117
4.2 Indirect costs ., 118
4.3 Operating and maintenance costs X19
4.4 Present worth , .\ ....... 120
5.0 Doses and potential health impact attributable 'to
krypton discharges from fuel reprocessing 121
6.0 Cost-effectiveness of krypton control at fuel
reprocessing plants 122
References €> J26
vii
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TABLES
I. Fuel Supply Page
A. Environmental Analysis of the Uranium Fuel Cycle,
Part I- (Fuel Supply) :, Uranium Milling 1
Section 2 .
2.0-1 Uranium Mills in Operation as of March 1975 4-5
Section J
3.1-1 Predicted Airborne Releases of Radioactive
Materials from the Highland Uranium Mill, Powder
River Basin, Wyoming 10
3.2-1 Concentrations .of Radioactive Effluents in Waste
Milling Solutions from the Highland Uranium Mill .. 11
3.2-2 Analysis of Waste Milling Solution from the
Humeca Uranium Mill (Alkaline Leach Process) 13
3.3-1 Estimates of Quantities of Radio'nuclides Seeping
Through the Impoundment Dam of a. Uranium Mill
Initially and at 2-1/4 Years 15
Section 5
5.0-1 Discharge of Radiottuclides to the Air from Model
Uranium Mills and Tailings Piles with Base Case
Controls 20
Section 6
6,0-1 Radiation Doses to Individuals Due to Inhalation in
the Yieinity of a Model Mill with Base Case.
Controls , 24
6.0-2 Collective Dose to the General Population in the
Vicinity of a Model Mill with Base Case Controls .. 25
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TABLES (continued)
Page
Section 8
8.1-1 Cost and Efficiencies of Control Technology for
Mills 29 :
Section 9
9.0-1 Radiological Impact of Airborne Effluents versus
Control Costs for a Model. Uranium Mill 35
B. Transuranium Effluents from Re-enriching or Refabricating
Reprocessed Uranium 47
1.0-1 Calculated Gamma Radioactivity Distribution of
Fission Products, Gamma and Beta Radioactivity
of all Fission Products, and Alpha Radioactivity
of Transuranium and Uranium Isotopes 49
1.0-2 Calculated Fission Product and Transuranium
Isotope Annual Inputs and Equilibrium System
Concentrations 52
3.0-1 Assumed Distribution of Fission Products and
Transuranium Isotopes to Atmosphere, Primary
Holding Pond, and Burial Ground 54
3.0-2 Estimated Radioactivity Released to the Atmosphere
from an Enrichment Plant 57
II. Nuclear Power Reactors
A. An Analysis of Control Options for N-16 Offsite Skyshine • "
Doses at Boiling Water Reactors 59
Section 2
2.0-1 N Characteristics of a Standard BWR Turbine
System 69-72
2.0-2 N^ Inventories for' a Standard BWR Turbine
System 73-74
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TABLES (continued)
Page
Section 5
5.0-1 Turbine Equipment Typical Total and Net N
Inventories (Ci) for a 1200 MWe Plant 75
5.0-2 Summary of Shielding Cost Estimates 76
III. Nuclear Fuel Reprocessing
A. Control of Iodine Discharges from Nuclear Fuel
Reprocessing Facilities 84
Se ct ion 3
3.0-1 Iodine Control Cost Summary 99
Section 5
5.0-1 100-year Cumulative Environmental Dose Commitment
and Estimated Health Effects Attributable to the
Release of 1-129 from a 1500 MTHM/yr Reprocessing
Plant 100
5.0-2 Maximum Individual Thyroid Doses from 1-129
Discharged from a 1500 MfHM/yr Reprocessing
Plant 101
5.0-3 Maximum Individual Doses from 1-131 Discharged
from a 1500 KEHM/yr Reprocessing Plant 102
Section 6
6.0-1 Cost-Effectiveness of Iodine Control Systems at
Fuel Reprocessing Plants. , 103
B. Control of Krypton Discharges from Nuclear Fuel
Reprocessing Facilities 109
Section 4
4.0-1 Estimated Capital and Present Worth Costs of
Krypton Control Systems 124
Secfeion 6
6.0-1 Cost-Effectiveness of Krypton Control at Fuel
Reprocessing Plants 125
XI,
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FIGURES
II. Nuclear Power Reactors Page
A. An Analysis of Control Options for N-16 Offsite
Skysh'ine Doses at Boiling Water Reactors 59
Section 3
3-1 Typical Component Layout in Early BWR Turbine
Building Design 77
3-2 Typical Component Layout in Current BWR Turbine
Building Designs 78
3-3 Contributions to Dose Rate from N-16 in Turbine
Building Components 79
Section 5
5—1 Top View of. Turbine Component Layout Showing
Typical "Access" Shield Design Along With
Various Shield Options 80
5-2 Transverse sectional view of Nine Mile Point 2
Nuclear Plant Turbine Building, Showing Shielding
of Moisture Separators and Turbines 81
5-3 Annual Dose at 500 Meters vs. Cost of Shielding
(Turbine Parallel to Boundary) 82
III. Nuclear Fuel Reprocessing
A. Control of Iodine Discharges from Nuclear Fuel Reprocess-
ing Facilities 84
Section 3
3-1 Simplified Schematic of Current Iodine Control
Systems at Reprocessing Plants 104
3-2 Simplified Schematic of Advanced Iodine Control
Systems at Reprocessing Plants ,. 105
xiil
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I. FUEL SUPPLY
A. Environmental Analysis of the Uranium Fuel Cycle,
Part I (Fuel Supply): Uranium Milling
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1,0 Introduetion
The EPA recently completed a teehnial review (1) of the
uranium milling industry as, part of an overall analysis of the
uranium fuel cycle (2_,_3). This review included a description of.
the milling process, estimations of radioactive effluent releases,
radiological impact, health effects impact, and the costs and
effectiveness of control technologies for mills. An analysis of
the tailings piles associated with mills was also included. This
review was prepared in support of EPA's proposed standards for the
nuclear fuel cycle, 40 CFR Part 190 (4).
Since publication in 1973, considerable new information on the
uranium milling industry has become available (j> ,£,J7 ,J5,JJ,JIO,1.1) |
in particular,. the engineering survey report (j>) , "Correlation of
Radioactive Waste Treatment Costs and the Environmental Impact of
Waste Effluents in the Nuclear Fuel Cycle for Use in Establishing
'as Low as Practicable1 Guides - Milling of Uranium Ores," has been
prepared by Oak lidge National Laboratory for the Nuclear Regulatory
Commission (NEC). This report contains an extensive review of the
costs and the effectiveness of various control technology systems
for uranium mills and mill tailings piles.
The EPA believes it to be worthwhile to revise its previous
technical review of the milling industry, taking into account these
new sources of information. Because radon-222 releases from fuel
cycle facilities have been specifically excluded from EPA's proposed
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standard, analysis of radon-222 releases from uranium mills and
uranium mill tailings piles has been omitted from this document.
Radon-222 will be the subject of separate regulatory actions at a
later date.
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2.0. General description of the milling process •'
A uranium mill extracts uranium from "ore. The product is a
semi— refined uranium compound (U~00) called "yellowcake" which is
j o
the feed material for the production of uranium hexaf luoride (UF-) .
6
As of March 1975, seventeen mills (7) were operating in the United
States (table 2.0-1) with nominal capacities ranging from 250 to
7,000 tons of ore per day. These mills are characteristically
located in arid, low population regions of the west. States with
significant high grade ore reserves are (j6) Wyoming, New Mexico,
Texas, Colorado, and Utah.
Eighty— five percent .of yellowcake is currently produced by a
process that uses sulfurie acid to leach the uranium out of the orej
the remainder is produced by a sodium carbonate, alkali leach process.
Exact details vary from mill to mill, but, as an example, the principal
steps in an acid leach process mill are as follows:
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 leaching 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 circuit while the washed residues
(tailings) are sent to the tailings pond or pile.
• . c. Solvent extraction or ion exchange is used to purify and
concentrate the uranium.
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Table 2.0-1 (7)
URANIUM MILS IN OPERATION AS OF MARCH 1975
COMPANY
Anaconda Company
Atlas Corporation
Conoco & Pioneer
Nuclear , Inc ,
Cotter Corporation
Dawn Mining Company
Exxon, U.S.A.
Federal-American
Partners
Kerr-McGee Nuclear
Petrotomics Company
Rio Algom Corp.
Onion Carbide Corp.
Union Carbide Corp.
LOCATION
Grants, New Mexico
Moab, Utah
Falls City, Texas
Canon City, Colorado
Ford, Washington
Powder River Basin, Wyoming
Gas Hills, Wyoming
Grants, New Mexico-
Shirley Basin, Wyoming
La Sal, Dtah
Uravan, Colorado
Natrona County, Wyoming
YEAR OPERATIONS
INITIATED
1953
1956
1961
1958
1957
1971
1959
1958
1962
1972
1950
1960
NOMINAL CAPACITY
(Tons of Ore/Day)
3000
800-1500
220-1750
150-450
0-400
2000
500-950
3600-7000
525-1500
500
0-1300
1000
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Table 2.0-1 (Continued)
COMPANY
LOCATION
.YEAR OPERATIONS
INITIATED
NOMINAL CAPACITY
(Tons of Ore/Day)
United Nuelear-
Homestake Partners
Utah International,
Inc.
Utah International,
Inc.
Western Nuclear,.Inc.
TVA (Mines Develop-
ment, Inc.)
Grants, New Mexico
Gas Hills, Wyoming
Shirley Basin, Wyoming
Jeffrey City, Wyoming
Edgemont, South Dakota
1958
1958
1971
1957
1956
1650-3500
750-1200
1200-
400-1200
250-500
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d. The uranium is precipitated with ammonia and transferred
as a slurry.
e. Thickening and centifuging are used to separate the
uranium concentrate from residual liquids.
f. The concentrate is dried at 400°F and is sometimes
calcinated at 750 to 1100°F.
g. The concentrate or yeUowcake is packaged in 208 liter
(55 gallon) drums for shipment.
Large amounts of solid waste tailings remain following the
remo'/al of the uranium from the ore. A typical mill may generate
1,800 metric tons per day of tailings solids slurried in 2,500
metric tons of waste milling solutions* Over the lifetime of the
mill, 100 to 200 acres may permanently be committed to store this
material.
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3.0 Releaseofradioactive effluent from uranium milling operations
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: uranium
is separated and concentrated, while most of. the radioactive daughter
products of uranium remain in the uranium-depleted solid 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, and airborne radioactive-releases
include radon gas and particles of the ore and the product uranium
oxide. External gamma radiation levels associated with uranium milling
processes are low, rarely exceeding a few mrem/y even at surfaces
of process vessels.
The tailings retention system or "tailings pond" will have a
radiological impact on the environment through the air pathway by
continuous discharge of radon-222 gas (a daughter of radiiim-226),
through gamma rays given off by radium-226, radon-222 and daughters
as they undergo radioactive decay,- and finally through air and water
pathways if radium-226 and thorium-230 are blown off dried out areas
of the tailings pond by wind or are leached from the pond into surface
waters (10,11).
3.1 Airbornereleases from the mill
Airborne releases from uranium milling operations include both
particulate matter and gases. Dusts containing uranium and uranium
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daughter products (principally thorium-230 and radium-226) are released
from ore piled outside the mill. Dusts containing uranium and uranium
daughter products are released from the ore crushing and grinding
ventilation system, while a dust containing mostly uranium without
daughters is released from the yellowcake drying and packaging
operations. These dusts are discharged to the atmosphere by means
of low stacks.
Uranium discharged to the air pathway as ore dust and as calcinated
yellowcake and the radium-226 and thorium-230 discharged to the air
pathway as ore dust are all considered insoluble aerosols. If they
are inhaled, aerosols that are insoluble in. body tissue fluids tend
to remain in the pulmonary region of the lung so that the lung becomes
the critical organ when the critical radiation dose is calculated.
The air flow through a typical crushing and grinding ventilation
system is about 27,000 cfm; that through the yellowcake drying and
packaging ventilation system is about 6,000 cfm. Because of the
different air flows, dust characteristics, and locations within the
plaat, separate air cleaning equipment systems are usually required.
A mill is usually considered to have two separate airborne effluent
release streams, each with its own control systems, costs, and source
terms.
Eadon gas is released from the ore storage piles, the ore crushing
and grinding ventilation system, leach tank vents, and the tailings
retention system. There is no practical method presently identifiable
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that will prevent the release of radon gas from uranium mills.
As an example, table 3,1-1 gives the estimated maximum release
rates and conservative estimates of site boundary concentrations
considering all potential sources of airborne dust fumes and mists
as predicted for the Highland Uranium Mill in Wyoming (12^23). The
capacity of the Highland Mill is about 2000 tons of ore per day.
3.2 Waterborne releases from the mill
The liquid effluent from an acid—leach process mill consists
of waste solutions from the leaching, grinding, extraction, and
washing circuits of the mill. These solutions, which have an initial
pH of 1.5 to 2, contain the unreacted portion of the sulfuric acid
used as the leaching agent in the mill process, sulfates, and some
silica as the primary dissolved solids, along with trace quantities
of toxic 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
tons per day of waste milling solutions from the Highland milling
plant are shown in table 3.2-1 (12_,J,3). 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 release of this material from
the site. The waste milling solution is, therefore, stored in the .
tailings retention pond which is constructed to prevent discharge
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Table 3.1-1 (12,13)
Predicted airborne releases of radioactive materials from the Highland Uranium Mill,
Powder River Basin, Wyoming8
Badionuelide
Uranium-natural
Thorium-230
(insoluble)
Badium-226
(insoluble)
Release rate
(Ci/y)
0.1
0.06
0.06
Site boundary A
Mr concentration
(pCi/ffl3)
0,003
0.001'
0.001
Site boundary Bc
Air concentration
(pCi/m3)
0.0004
0.0001
0.0001
Mominal mill capacity 2000 tons of ore/day (1200 MI of yellowcake per year).
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.
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Table 3.2-1
Concentrations of radioactive effluents in
waste milling solutions from the Highland uranium mill (12,13)
(acid leach process)
Radionuclide Concentration
(pCi/1)
SL
Uranium-natural 800
ladium-226 350
JThorium-230 22,000
aAbout 0.001 g/ml.
11
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into the surface water system and to minimize percolation into the
ground.
As an additional example, an analysis of waste milling solution
for the Humeca Uranium Mill, which uses the alkaline leach process,
is given in table 3.2-2 (9). The solution has a pH value of about
10 and contains sodium, sodium carbonate, sodium bicarbonate, and
sulfate as the principal dissolved solids.
3.3 Airborne and waterborne releases from the mill tailings pond
The following discussion refers to the best of current procedures
for handling mill liquid and solid wastes.
The waste milling solution is used to slurry the solid waste
tailings to a tailings retention pond system which uses an impervious
clay-cored earth dam combined with local topographic features of the
area to form an impoundment. The clay-cored dam retention system
permits the evaporation of most of the contained waste liquids and
serves as a permanent receptacle for the residual solid tailings
after the plant closes.
Toward the end of the operating lifetime of a tailings retention
'}
system, some of the tailings will no longer be under water and will
dry out to form a beach (6). Mind erosion can then carry off tailings
material as airborne particulate matter unless control measures are
taken to prevent such erosion. Considerable quantities of radon
will be emitted.
12
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Table 3.2-2 (9)
Analysis of the waste milling solution
from the Humeca Uranium. Mill
(alkaline leach process)
Rad^-onuclide ' p'Ci'/l
Radium-226 240
Thorium-230 110
Uranium-238 and 234 46,000
13
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Immediately after the retention system is put to use, it is
to be expected that there will be small losses of radioactive mill
waste liquids through and around the dam (.9 »_12) • This will be seen
as surface water seeping from the foot of the dam. The radiological
significance of this seepage will depend on the location of the
pond. In arid regions, the seepage may evaporate before leaving
the site, leaving the radioactivity entrained and absorbed on soil.
Should the tailings pond be located next to a river, minor 'leakage
might be immediately Saluted sufficiently by the additional river
water to meet relevant drinking water standards. Discharge of
pond seepage into streams providing insufficient dilution and not
under the control of the licensee would not be acceptable. In such.
cases, a secondary dam may be built below the primary dam to catch
the seepage which may then be pumped back into the tailings ponds.
It is sometimes stated 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 (12).
Examples of estimates of' the total quantities of radionuclides
that would be released through and around the dam to surface waters
are shown in table 3.3-1. Radium-226 is a radionuclide of concern
because levels as high as 32 pCi/1 (14) have been found in seepage
from current operating mills. Assuming a seepage rate of 300 liters
per minute, 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 1/5 of EPA's proposed interim Primary Drinking
14
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Table 3. 3-1
Estimates of quantities of radionuclides seeping through the
impoundment dam of a uranium mill initially and at 2-1/4 years (12,13)
Initial seepage Seepage per day'?'
Radionuclide per day after 2-1/4 years
Uranium 350 yCi 35 pCi to 3.5 pCi
Thorium-230 9,600 yGi 960 pCi to 96 pCi
Radium-226 150 pCi 15 yd to 1 .5 pCi
'^Seepage assumed to be inhibited due to seal ings effect from
accumulation of fines between sandstone grains.
15
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Water Regulations for radium-226 (15). In the applicant's environ-
mental report for the Highland Uranium Mil GL2_».13) , a seepage
concentration of 350 pCi/1 radium—226 was assumed.
Considerable quantities of mill waste solution seep downward
inGb the soil beneath the impoundment area. Ordinarily this is
not expected to result in offsite releases of radioactive materials
because the radionuclides are strongly absorbed onto clay soil
particles. They are removed from solution and considered to be
permanently retained on the mill site. However, this is a continuing
potential problem requiring monitoring programs to insure that there
is no significant movement of contaminated liquids into the offsite
environment (10).
16
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4.0 The Model uranium mill
A model plant has been assumed 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 design and
projected commercial industry practice (j>) . However, it is not
necessarily representative of presently operating facilities.
Characteristics of the model mill are assumed to be:
a. 600,000 MT ore milled per year,
b. 1,140 MT UJ30 as yellowcake produced per year,
j o
c. .use of the acid leach process,
d. a tailings retention pond system which uses a clay—core
earth dam and local topographic features of the area to form the
impoundment,
e. collection and return of any seepage through the dam to the
tailings pond, and
f. location in a western State in an arid, low—population density
region.
While reference (1) considered the radiological impact of
seepage through a model clay core impoundment dam, it is now believed
to be standard practice (6) to collect and return any such seepage,
to the tailings pond so that there are no routine liquid discharges
of radionuclides to water pathways from mills. The cost of a seepage
control system is nominal compared to the cost of the tailings
17
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impoundment system Itself. .......
Radiation dose rates and health effects that -might result from
the discharges of airborne radioactive effluents from the model mill
are calculated using representative X/Q values, dose conversion factors,
model pathways, and health effect conversion factors that are
similar to those for other facilities in the previous discussion
of the fuel supply cycle. These factors and assumptions are discussed
in Appendix A of reference (1) .
Values of (X/Q) given in the ALAP Guides for milling of uranium
ores (6) as derived from meteorological data near actual uranium mills
f\ — *3
range from 2.3 to 8.7 x 10 a*m for a New Mexico site and range
7 g" r»
from 5.1x10 to 5. Ox 10 s-m for a Wyoming site. The maximum
values for these sites are in agreement with the value used in
f _ O
reference (1) of (X/Q)m«._. of 6 x 10 s»m . This value would apply
— iUciA
to individuals living from 0.5 to 1.5 kilometers downwind from the
mil site. Values of (X/Q) for individuals living outside the
sector containing the prevailing wind wJ.ll be up to 3 to 12 times
lower. The committed lung dose will also be lower in direct proportion,
The operating lifetime of a uranium mill is commonly from 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 the model mill, an operating lifetime of 20 years
has been selected. After the mill shuts down, it is assumed that
the tailings pond will be allowed to dry out and that the resulting
pile will be stabilized and placed under perpetual care.
18
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5.0 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 model mills are based on somewhat limited
operational information (6). Source terms listed in table 5.0-1
are believed to be reasonably accurate estimates of the quantities
of radioactive materials discharged to air pathways from model mills
with base case controls. The controls assumed as the base case
consist of an orifice scrubber on the crusher and fine ore bins,
and a wet impingement scrubber in the yellowcake drying and packaging
areas. The milling procedures are so similar for acid and alkaline
leach processes that source terms for the two types of mills are
considered identical, except that the alkaline leach process does
not remove thorium from the ore so that, in this case, there is very
little thorium-230 as an impurity in the yellowcake dust.
The model mill ±s assumed to use clay—core dam impoundment
technology for tailings with a catch basin if required to contain
seepage through the dam. Unless the impoundment area is lined with
an impervious material, considerable quantities (as much as 10 percent)
of the liquid effluent from the mill will leak out through the bottom
of the pond. However, because of the ion-exchange properties of most
soils, radionuclides dissolved in this effluent will attach to soil
particles and will not reach offsite locations or ground water. The
model mill is considered, therefore, to'deliver no radiation exposure
to members of the general population through liquid pathways.
19.
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Table 5.0-1
(a)
Discharge of Radionuclides .to the Air from Model Uranium Mills and Tailings Piles (6)
With Base Case Controls
Radionuclide
Uranium-238 and 234
Radium-226
Thorium-230
Uranium-238 and 234
Badium-226
Thorium-230
Draaium-238 and 234
Radium-226
Thorium-230
Chemical or
Physical State
ore dust (oxides)
ore dust
ore dust
yellowcake (oxides)
yellowcake
yellowcake
tailings sand (0-10 vim)
tailings sand (0-10 vim)
tailings sand (0-10 urn)
Acid Leach Mill Alkaline Leach Mill
Source Term Source Term
(mCi/y) (mCi/y)
9.0 9,0
4.5 4.5
4.5 4,5
170. 170.
0.2 1.7
4.7 \
0.2 - 0.8 0.3 - 2.2.
1.3 - 4.2 2.3 - 1,5
1,4 - 4.5 2.4 - 1.5
moisture ore, radion-222 releases excluded
-------
Each site must be evaluated Individually. If the ground water
table is high and the soil is low in ion exchange capacity so that
it becomes likely that radium.-226" and thorium-~23Q will escape from
the tailings impoundment into underground waters, then the pond area .
could be lined with an impervious membrane of asphalt to minimize
seepage. Acid wastes would have to be neutralized beforehand to
prevent damage to this type of liner.
"Ehe amount of radioactive particulate material removed from the
tailings beach by wind erosion is believed to depend pn the area and topog-
raphy of the beach, the wind'velocity, and particle size distribution of the
tailings (J5) . Estimates of this source term are shown in table 5-.0-1
and include only the alpha emitting radionuclides U-238, U-234,
Th-230, and Ra-226 which are the significant contributers to the
lung dose. While this, estimate is derived from theoretical considera-
tion rather than experimental measurements at actual tailings beaches,
it is believed to be the best available estimation for this source
term. Particles greater than 10pm in, diameter are not considered to
be respiratole particles and are not included in the inhalation source
term pathway. Historically, windblown tailings, have caused elevated
gamma exposure levels around piles, but the inhalation pathway is
usually considered to be the critical pathway because levels of
control sufficient to limit radiation exposure through the inhalation
pathway will also prevent, to a significantly greater degree,
exposures through the ground deposition whole body exposure pathway.
21
-------
The ALAP document developed for the Nuclear Regulatory Commission (6)
provides an estimation of the relative ratio of the respirable
particles (< 10 ym) to larger particles (10-80 ym) blown off the
tailings beach of a well-managed tailings impoundment system. This
ratio averages about 1 and varies from 0.4 to 1.4 depending on
specifics of the milling process and other variables. It can be
estimated that 1 mCi/y of alpha emitting insoluble 0-10 ym particles
removed by wind from a typical pile would deliver a dose equivalent
of approximately 1 mrem/y to the lungs of a person living one
kilometer downwind of the pile. At the same time, if it is assumed
that 1 raCi/y of 10-80 ym particles are deposited in a ring % to 1% km
from the pile, there would result a surface contamination level of
2
about 0.2 nCi/m . The Ra-226 component of this, surface contamination
would cause a. whole body gamma-ray exposure level of about 10 yrem/y.
After 20 years' of operations, each contributing to surface contamination
at such a rate, this exposure might increase to as much as approxi-
mately 0.2 mrem/y. This is still a factor of 5 smaller than the
lung dose from the inhalation pathway indicating that inhalation is
the critical exposure pathway.
22
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6.0 Radiological Impact of a model mill
Estimates of -the. radiation doses to individuals • through the air
pathway in the vicinity of an acid leach model mill using base case
controls from routine emissions are shown in table 6.0-1. The
estimated collective lung doses to the population in the vicinity of
an acid leach mill are given in table 6.0-2. The collective lung
dose is determined by summing the average individual radiation dose
equivalent to individuals living within 80 kilometers of,.the mill over
the total population within 80 kilometers of the mill. The models
for the dispersion and dose calculations are discussed in detail in
Appendix A of reference (_1) . Based on the information available at
the time that analysis was performed, an effective -half—life of ' "
1,000 days was used for insoluble class Y compounds in the pulmonary
region of the lung in calculating the lung-doses from mill emissions.
f
In accordance with what is now becoming accepted practice, in this
report all dose conversion factors are calculated using a 500-day
effective half-life (20) and are, therefore, reduced by a factor of
two from the previously used values.
It is assumed that food consumed by -individuals living near the
mill is not produced locally so that exposure through food chains is
not significant compared to lung exposures resulting from the direct
inhalation of radioactive particulate matter. The radon exposure.
pathway was excluded from this report.
Because there are no liquid releases from the model mill, there
is no projected radiological impact through water pathways.
23
-------
fable 6.0-1
Radiation Doses to Individuals due to Inhalation
in the Vicinity of a Model Mill with Base Case Controls
Radionuclide
Source
Term
(mCi/y)
Critical
Organ
Dose Equivalent
Individual at Plant
Boundary
(tar em/ y)
to Critical Organ
Average Individual
Within 80 km
(mrem/y)
Oraniwn-234 180
and 238
Thorium-230
Radium-226
15
10
Lung
Lung
Lung
Total 205
170
15
15
200
3.9 x 10~2
3.4 x 10~3
2.2x 10~3
4.5 x 10~2
-------
Table 6.0-2
Collective Dose to the General Population in the
Vicinity of a Model Mill with Base Case Controls
o
„ .. ,., mu _. , Critical Collective Critical Organ Dose
Radionuclide Term Pathway •
(mCi/y) rgan (person rem/y)
Uranium-234 and 238 180 Air Lung 2.2.
Thorium-230 15 Air Lung 0.2
Radium-226 10 Air Lung . 0.1
Total 2.5
aReleases to water pathways assumed equal to zero, and doses from radon-222 are
not included.
v 4
The population model for the model mills assumes that 5.5 x 10 persons are
exposed within 80 km of the mill site.
25
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7.0 Health effects impact of a model mill
Potential health effects to members of the general population
in the vicinity of a model mill using base case controls are
estimated to be 0.0002 lung cancers per year of operation or 0.005
lung cancers for 30 years of operation. The models used for the
calculation of health effects are given in Appendix A of reference (1)
26
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8.0 Control technology for, uranium milling • •
8,1 Airborne effluent control technology
Hazardous airborne gaseous and particulate wastes are generated
in the milling operation from a number of different sources. The
major areas of the milling operations in which gaseous and particulate
matter effluents must be controlled are the ore crushing area, the
fine ore bins, and the yellowcake drying and packaging areas. Mills
often prefer to use multiple dust collection systems rather than
design a. single, more elaborate system. There will usually be two
or more ore dust collectors and separate systems for the. yellowcake
dryer and for the yellowcake packaging rooms.
Dust collector systems that are currently used or that can be
adapted for use by uranium mills are discussed in reference (_§_) .
They are for the most part control technologies, that have.been proven
and are standard industrial equipment.
Briefly, these treatment methods are:
a. Orifice Scrubbers - The dusty air flows through a stationary
baffle system coated with a sheet of water. The.dust, particles
penetrate the water film and are captured-.
b. Wet Impingement Scrubber - The dusty air carrying water .
droplets added by preconditioning sprays passes through perforated
plates to atomize the water and to wet the dust. Particles are then
collected by impingement on baffle plates and a vaned demister.
c. Venturi Scrubber - The dusty air is passed through a venturi,
increasing its velocity. Water is added which atomizes in the gas
27
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stream and collects the dust by impingement. The wetted dust is
removed by demisters. Raising the pressure drop across the venturi
increases the collection efficiency, but this requires higher energy
levels and raises the costs.
d. Bag Filters - These filters are made of woven or felted
fabric and have high collection efficiencies provided the air being
filtered is cool and dry.
e. HEPA Filters - These filters are made of fiber glass.
They have very high efficiencies but have a number of limitations;
in particular, they can only be used in conjunction with a prefliter
and on dry air streams.
Current practice involves the use of wet dust control systems
although several mills use bag filters for air flows from ore
handling and from the yellowcake packaging area. The costs and
percent effluent reduction for the various control systems suitable
for effluent streams of the model mill are given in table 8.1-1 (6).
Particulate material can be prevented from being windblown off
the tailings pile beach by back filling with overburden and, as an
interim measure, by chemical stabilization by spraying with various
polymers or petroleum derivatives. Chemical stabilization is
expected to last about a year and must be repeated on a regular
schedule (j6) . Although no specific value is given for the percent
reduction of airborne effluent by these control measures, it is
assumed that they would reasonably reduce the tailings beach source
28
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Cost and Efficiencies of Control Technology for Mills'*'
A.
B.
Control Method
Gaseous (Crusher -and line Ore Bins)
1. Orifice Scrubber
2. Wet Impingement Scrubber
3. Low Energy Venturi Scrubber
4, Bag Filters
Gaseous (Yellowcake Drying and Packaging)
1. Wet Impingement Scrubber^0'
2. Low Energy Venturi Scrubber t°)
3. High Energy ¥enturi Scrubber
4, High Energy Venturi Scrubber + HEPA
Filters
Capital Cost
(dollars)
101,000
116,000
173,000
300,000
(35,000)
(35,000)
46,000
106,000
Annual
Operating Costs
(dollars)
7,200
8,600
17,000
21,000
(3,500)
(6,900)
15,000
22,000
Present WorthP1)
(dollars)
,
172,000
200,000
340,000
506,000
(69,000)
(103,000)
193,000
322,000
Percent
Effluent
Reduction
(%)
93.6
97.9
99:5
99.9
97.9
99.5
99.9
>99.99
C. Liquids, Solids, and Windblown Particulate
Matter '
1. Clay Core Dam Retention System with 2,250,000
Seepage Return and 0.6 Meters (2 feet)
of Earth Cover Plus Rock Stabilization^6^
2. Chemical Control of Windblown Dust from 63,000
Tailings Pond Beach
3. Asphalt Liner for Tailings Pond^ , 800,000
50,000
8,000
0
(d)
2,750,000
142,000
800,000
100.00
100.00
(a)l974 dollars; radon-222 emissions not included,
Wpresent Worth = Capital Cost + (Annual Cost x 9.818); 8% Discount Rate, 20 yr. Plant Lifetime.
*-c'Costs for all yellowcake effluent control are shown for completeness. In actual practice, the value of
recovered product more than compensates the cost of control options Bl and B2.
WIncludes investment to provide for perpetual care.
acre tailings pile.
-------
term by greater than a factor of 10 (i.e., to < 1 mCi/y).
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 moisture content of the ore as mined may be
sufficiently high to eliminate most dust formation in the ore
storage and blending area; due to insufficient information, this
case will not be considered at present beyond stating that the
problem appears potentially significant and that it can be controlled
in principle through sprinkling and by use of wind breaks. Dust
generation on ore haul roads can also be controlled by sprinkling.
8.2 Waterborne effluent control technology and solid waste control
technology
New mills in the Rocky Mountain area are' using impoundment
technology in order to approach 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:
(a) limited runoff, (b) dammable downstream openings, and (c) 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 postulated
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 p.umped to the impoundment
reservoir for storage and liquid reduction. Liquid reduction is
accomplished primarily by evaporation, but also by seepage through
30
-------
the dam, the reservoir walls and floor. By filling a dammed natural
depression with tailings, a relatively flat, stable contour is
achieved. There usually will be a continuing problem with control
of upstream drainage. Diversion ditches to control this drainage
will require perpetual maintenance.
Two methods for seepage collection and return are being
considered for new mills. In that situation when 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 release to the offsite environment. In that situation
where either an underlying impermeable geological formation is not
existent or is .not continuous, vertical seepage may occur to the
underlying ground water formation. Wells may be drilled downstream
of the retention system into the subsurface formations where seepage
will' collect, and this-, water is pumped, back to the - retention system..
Such a system requires specific favorable subsurface conditions. In
both cases, these control costs are small compared to the cost of the
clay core dam retention system (1).
Impoundment of solids is being accomplished at many older mills • -
by construction of a dike with local material and then 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
31
-------
through the dike have sometimes been routed through a treatment
system and discharged to the environment. The diking procedure,
which is less costly initially, creates an above-ground pile of
tailings which is difficult and costly to stabilize. While the
mill is operating, this type of pile is also subject to wind and
water erosion. Field studies at tailings piles after mill shutdown
have shown high gamma radiation levels in the vicinity of such piles,
elevated radium-226 levels in water supplies, and high airborne
levels of thorium-230 and radium-226 due to wind blown tailings
(i§»,iZ.».l§. ».!£}• ^or these reasons, new mills are not likely to be
built using this type of solid waste control.
After the mill shuts down, stabilization of the tailings pile
after it has dried out requires contouring of the tailings area to
lessen side slopes, establishing drainage diversion, covering with
nonradioactive material, and revegetating the area. In semiarid
regions it may be necessary to initially irrigate the pile to achieve
vegetation growth; in arid areas, vegetative cover without perpetual
irrigation will not be possible. 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 significantly reducing
or eliminating migration of the tailings outside the controlled area,
The costs of such stabilization has recently been estimated (_6) at
$350/acre-ft for earth, and $2,000/aere-ft for rock. The stabilization
32
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of a grade level diked tailings pile is more costly and is probably
less effective compared to a depression fill tailings pile because
of difficulties faced in contouring, covering, and revegetating the
potentially steep side slopes.
Uranium mill tailings piles are long half-life, low-level
radioactive wastes. As such, they will require perpetual care.
This will include occasional inspection and maintenance to insure
integrity of the stabilizing cover, fencing,, and. of the warning signs
/
around the pile. A perpetual care fund should be included as part
of the cost of the control technology to pay for this care. The
maintenance associated with perpetual care of a stabilized dike
system would probably be higher than that for the depression fill
system, since there is tendency toward collapse of side slopes and
possibly inadequate drainage of precipitation from the pile.
33
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9.0 Effluent controltechnology for the modelmill
Typical current effluent control systems were assumed for
the model mill. They were;
a. Ore Crusher and Ore Bin Dust - Orifice Scrubber.
b. Yellowcake Dryer and Packaging Dust - Wet Impingement
Scrubber.
c. Liquid and Solid Waste - Clay core dam retention system *
(160 acres) with seepage return and exposed beach. To be stabilized
with 2 feet of earth cover and 6 inches of rock cover.
The radiological impact of total airborne effluent versus
successively more effective control systems for a model uranium
mill are listed in table 9.0-1. Each improvement in control is the
most cost-effective available at that level of control.
The output of the model plant using base case contols is 1,140
tons of yellowcake per year of which approximately 1 percent is
recovered by the wet impingement dust collector system during
drying and packaging operations (6). The value of 11,000 kilograms
(24,000 Ibs) of recovered yellowcake more than compensates for the
cost of this control system. The low energy venturi scrubber is
1.6 percent more efficient than the wet impingement scrubber and
will recover an estimated additional 200 kilograms (440 Ibs) of
yellowcake per year. The value of this additional recovered yellow-
cake is approximately equal to the increased annual operating costs
of the low energy venturi scrubber as compared to the wet impinger.
The present worth of these systems are, therefore, not included as
a control cost for the model mill.
34
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Table 9.0-1
Radiological Impact of Airborne Iffluents versus Control Costs for a Model Uranium Mill
Controls
(Table 8.1-1)
None
Al;
Al;
Al;
A2;
A2;
A2;
A3;
A4;
Cl<<
Bi20,000
205
75
35
25
15
6
1.5
0.3
0
Maximum .Lung
Dose to an
Individual 0>)
(mrem/y)
>2 0,000
200
73
34
24
15
6
1.5
0.3
.0
Present Worth
(1974 $/facility)
0
172,000
172,000
262,000
290,000
432,000
561,000
701,000
867,000
2,750,000
WAlpha emitting radionuclides as insoluble, xespirable jpartieulate matter,, excluding radon and daughters
Cp)For the assumed worst case of an individual permanently occupying a location exhibiting
a x/Q of 6 x 10~6 s/m?. ' .-••'.-
lc'Assumed current level of controls for new mills.
Wcosts for control options Bl and B2 not included, since they are more than compensated for by
the value of product recovered.
-------
10.0 Retrofitting control technology to operating uranium mills
The cost and practicality of retrofitting control technology
systems to an operating uranium mill if it should be required to
comply with EPA1 s proposed standards (40 CFR 190) was not included
in reference (6). The costs are judged to be approximately the
same order of magnitude as the costs to install the same control
systems in a new mill.
10.1 Retrofitting control measures to operational tailings ponds
The cost and practicality of retrofitting control measures to
operational tailings ponds that do not use clay core dam impoundment
technologies must be considered on an individual basis. EPA has
reviewed the available literature concerning 17 operational uranium
mills. Based on this survey, it was concluded that of the 17 mills,
the presumption of evidence indicated that 7 would be in compliance
with the Agency's proposed (4.0 CFR 190) standards while 10 mills would
require remedial measures of varying degrees to comply with the
standards.
Three mills, opened since 1971, use advanced impoundment
technology designed to prevent loss of tailings material. This
includes use of a natural basin with a clay core earth dam across
the opening to impound the tailings. The tailings are below grade,
protected from wind erosion, and depending on the season, are often
either moist or actually covered with water which effectively provides
additional protection against wind erosion. These mills are in
36
-------
remote locations with no residence within one mile. The use of
advanced tailings impoundmeiit techniques and the remoteness of
the sites should be sufficient to insure compliance during the
active life of these mills.
Four mills are located in remote areas where no one is
believed living within about one mile of the site. In addition,
the active tailings ponds are either impoundments.in natural basins
or, if above ground, the sides are stabilized with rock. There
may be inactive tailings pile areas on several of these sites that
could be stabilized at this time. The combination of reasonable
tailings impoundment techniques and large distance to the nearest
resident should be sufficient to insure compliance as long as
these conditions are in effect.
For the remaining 10 mills, members of the general population
are believed to reside within 1 mile of the sites. An evaluation
of each tailings pile and pond will therefore be required to
determine compliance with EPA standards because a recent study (21)
has indicated that windfelown tailings from inactive unstabilized
tailings piles has caused elevated gamma exposures > 25 mrem/yr at
distances up to one mile from the pile. Critical pathways to be
considered are inhalation of insoluble alpha-emitting radioactive•,
particles windblown from the pile (11), deposition of radioactive
particles windblown from the pile causing whole body exposure from
gamma rays (21), and radioactive contamination of drinking water by
seepage from the tailings pond or by discharge of mill process water (10)
37
-------
Five of these tailings piles are judged to require slight, if
any, remedial measures to comply with the standard. These are
relatively small piles in remote locations where tailings pile
dikes have been constructed of earth and clay rather than tailings
sand.
The other five tailings piles are judged to require major
remedial measures to comply with EPA standards. These are, in
general, large tailings piles located above grade with dikes
constructed of tailings sand and where persons live in close
proximity to the pile.
It is not appropriate for 1PA to specify in detail an implementa-
tion plan for each mill to comply with the proposed standards. Hie
Agency is on record as stating that the standards should be imple-
mented with regard to operational tailings piles by requiring proper
and reasonable dust control measures. In practice, this means that
all tailings material should be stabilized, covered, or otherwise
controlled by chemical stabilization or by keeping the tailings under
water or at least moist. In the absence of very large controlled
areas, or unless individuals live more than a mile from the tailings
pile, the tailings pile source term must be kept very low (<1 mCi/yr)
by use of these procedures. Otherwise, a detailed site specific
dose assessment (modeling) effort and perhaps environmental monitoring
will be required to demonstrate compliance.
"38
-------
Xn the event the implementation proceedings conducted by the NEC or
an agreement State determine that a specific tailings- pile is not in
compliance, a variety of reasonable remedial measures are available to
the mill operator at reasonable cost. These measures include:
1. Enlarge the restricted area around the site and move people
living near the site to more distant locations.
In some instances, the closest residents are employees of:the
company and their families living in trailers next to the site •
boundary. It would appear that moving these people: would .be a practical
protective action to take.
2. Cover and stabilize all unused tailings piles and ponds.
There are piles and ponds at some sites that have been filled
to capacity. These can be stabilized immediately to reduce wind blown
tailings. This is especially important.for carbonate leach process
tailings piles which contain finer material and are believed to be
more susceptible to wind erosion.
3. Cover and stabilize tailings pile dikes constructed of.
tailings sands. This may be accomplished by covering with earth
and use of rock as rip rap or, temporarily,'by chemical sprays.
The sand,dikes at one active tailings pile have been stabilized
using crushed rock from local sources. The dikes at the inactive
tailings pile at Tuba City (22) were temporarily stabilized at.
reasonable cost using chemicals that bound the surface sands together
to form a hard crust. They were sprayed with an elastomeric polymer
39
-------
forming a 2" crust cover for about $760 an acre (1975 dollars) of
dike. While this cover eventually broke up, due in part to lack
of pedestrian access control, chemical stabilization of dikes should
be effective under more controlled conditions for several years.
Additional applications would be necessary. Continual maintenance
consisting of patching small holes before they become large holes
would probably be effective in increasing the overall lifetime of
the chemical stabilization.
When a mill is shut down and before the license is terminated,
it is NIC policy that the tailings pile must be stabilized. At the
present time, this entails covering the pile with earth and either
establishing vegetation or using rock rip rap to protect the cover
from wind erosion. Because it must be done eventually, it may be
more cost effective to use earth stabilization of sand dikes at
operational piles rather than use temporary chemical stabilizers
that must be reapplied every few years.
4. Stabilize the tailings pond beaches, i.e., the material
contained inside the dikes. This may be done with chemical sprays,
by sprinkling with water, or by covering with water or backfill.
Tailings ponds are often so large that only a portion of them
are under water continuously. Large areas may dry out and become
susceptible to wind erosion. If these dry areas are firm enough
to hold heavy equipment, it should be possible to cover them with
backfill. Otherwise, chemical stabilizing applied by sprinklers
40
-------
can be used. This will "be a temporary measure requiring reappliea—
tion every few.years. The tailings beach at the Tuba City pile (22)
was stabilized with calcium magnesium lignosulfonate at a cost of
about $430 per acre (1975 dollars). If enough water is available,
continuous.sprinkling can be used to keep the surface wet and prevent
wind erosion,.
The State of Texas, which is an agreement State, has determined
that wind blown tailings from an active tailings pile near Fall City,
Texas, must be controlled. As the dikes for this pile were constructed
using sandy clay rather than tailings sand, this will prove to be an
f
example .of control of a tailings beach by some, means as sprinkling,
backfill, or chemical cover. • ...
5. Close down the tailings pond and stabilize it; construct
a new tailings pond using advanced tailings impoundment techniques.
This may be the best procedure when the tailings pond- is of
such configuration (i.e., very high dike walls) that it must be
reshaped before stabilization procedures are .effective and where the
mill is expected to continue in operation for some time. Multiple
tailings ponds on a single site are common practice.
The reasonableness and cost of stabilizing an active- uranium
mill tailings pond may be examined by considering a "model" tailings
pond. A. model pond is assumed to be 100 acres in total area and,
contained .by tailings sand dikes 7 meters high and 20 acres in area
with a dry beach of 50 acres. The remainder of the area inside the
41
-------
dikes will be under water or continuously wet. All of the following
costs are givea in 1975 dollars.
By analogy with the experience with the Tuba City pile (22),
it would require $15,000 to chemically stabilize the dikes and
$22,000 to chemically stabilize the beach. Five stabilizations
(biannually over a 10 year period) would imply total costs of $110,000
to stabilize the beach and $75,000 to stabilize the dikes.
As an alternative, the dikes could be permanently stabilized
by earth. If it is assumed that this would require the covering
3
of one side of a 2,600 meter long dike by 50m of earth per meter
3
of dike at a cost of $1 per m , then the cost would be about
$130,000. There would be additional costs of establishing a
vegetation cover or for rock rip rap. If the cost of stabilizing the
dikes is considered as part of the final stabilization costs, the net
cost of complying with the standard would then be $110,000, the cost
to chemically stabilize the beach.
Costs (1975 dollars) for stabilizing inactive piles vary (6,22).
Arizona Copper procedures report that costs of stabilizing with a
12" soil cover were about $1,600 per acre. Stabilization of the
Monticello, Utah, pile which involved considerable moving and
contouring of the tailings sand, with 12" to 24" of soil and with
vegetative planting, cost $7,300 per acre. Union Carbide has
calculated their cost of stabilization at $1,300 to $5,100 per acre
for a minimum cover depth of 6" with costs depending on grading and
42
-------
distance that rock and rip rap must be hauled. The ALA'P Guide for
Milling of Uranium Ores (6) estimated cost of $510 per acre foot
for earth and $3,000 per acre foot for rock.
The Agency concludes that tailings piles at active, uranium
mills can meet the proposed standard 40 CFR 190 by the application
of reasonable and proper remedial measures. The cost of implementing
the standard will be small compared to.the eventual overall costs
of stabilizing the tailings sands.
43
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REFERENCES
(1) U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis
of the Uranium Fuel Cycle, Part I - Fuel Supply, EPA-520/9-73-
003-B. Office of Radiation Programs, Environmental Protection
Agency, Washington, B.C. 20460 (October 1973).
(2) U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis
of the Uranium Fuel Cycle, Part II - Nuclear Power Reactors,
EPA-520/9-73-Q03-C. Office of Radiation Programs, Environmental
Protection Agency, Washington, D.C. 20460 (November 1973).
(_3) U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis
of the Uranium Fuel Cycle, Part III - Nuclear Fuel Reprocessing,
IPA-520/9-73-003-D. Office of Radiation Programs, Environmental
Protection Agency, Washington, D.C. 20460 (October 1973).
(4) U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Radiation
Protection for Nuclear Power Operations, 40 CFR Part 190.
Federal Register, Volume 40, No. 109 (Thursday, May 29, 1975).
(5) U.S. ATOMIC ENERGY COMMISSION. Draft Environmental Statement
Related to the Utah International, Inc., Shirley Basin Uranium
Mill, Shirley Basin, Wyoming, Docket No. 40-6622. Fuels and
Materials Directorate of Licensing. U.S. Atomic Energy Commission
(June 1974).
(jj) SEARS, M. B., et al. "Correlation of Radioactive Waste Treatment
Costs and the Environmental Impact* of Waste Effluents in the
Nuclear Fuel Cycle for Use in Establishing "as Low as Practicable1
Guides - Milling of Uranium Ores," ORNL-TM-4903, Two Volumes.
Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830
(May 1975).
(2) 1EKNEKRON, INC. "Scoping Assessment of the Environmental Health
Risk Associated with Accidents in the LWR Supporting Fuel Cycle -
Draft Report," EPA Contract No. 68-01-2237. Teknekron, Inc.,
Washington, D.C. 20036 (September 2, 1975).
(8) "Controlling the Radiation Hazard from Uranium Mill Tailings,"
Report of the Congress by the Comptroller General of the
United States, RED-75-365 (May 21, 1975).
(9.) U.S. NUCLEAR REGULATORY COMMISSION. Final Environmental Statement
Related to the Operation of the Humeca Uranium Mill, NU1EG-0046,
Docket No. 40-8084. Fuels and Materials Directorate of Licensing,
U.S. Nuclear Regulatory Commission, Washington, D.C. 20545
(April 1976).
44
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(10) U.S. ENVIRONMENTAL PROTECTION AGENCY. Water Quality Impacts
of Uranium Mining and Milling Activities in the Grants Mineral
Belt, New Mexico, EPA-906/9-75-002. U.S. Environmental Protection
Agency, Region VI, Dallas, Texas 75201 (September 1975).
(11) SWIFT, J. J., J. M. HARDIN, AND H. W. GALLEY. Potential
Radiological Impact of Airborne Releases and Direct Gamma
Radiation to Individuals Living Near Inactive Uranium.Mill
Tailings Piles, EPA-520/1-76-001. Office of Radiation Programs,
U.S. Environmental Protection Agency, Washington, D.C. 20460
(January 1976).
(12) HUMBLE OIL AND REFINING COMPANY. ^Applicant's Environmental
Report, Highland Uranium Mill, Converse County, Wyoming. Minerals
Department, P.O. Box 2180, Houston, Texas 77001 (July 1971).
(13) HUMBLE OIL AND REFINING COMPANY. Supplement to Applicant's
Environmental Report, Highland Uranium Mill, Converse County,
Wyoming. Minerals Department, P.O. Box 2180, Houston, Texas
77001 (January 1972).
(14) U.S. ENVIRONMENTAL PROTECTION AGENCY. Evaluation of the Impact
•of the Mines Development, Inc., Mill on Water Quality Conditions
in the Cheyenne River. EPA Region VIII, Denver, Colorado. 80203
(September 1971). .. . .
f
(15) U.S. ENVIRONMENTAL PROTECTION AGENCY. Interim Primary Drinking
Water Regulations - 40 CFR Part 141. Federal Register, Volume 40,'
No. 158 (Thursday, August 14, 1975).
(16) SHELLING, R. N. AND S. .D.•SHEARER, JR. Environmental Survey of
Uranium Mill Tailings Pile, Tuba City, Arizona. Radiological
Health Data and Report 10:475-487 (November 1969).
(17> "SHELLING, R. N. Environmental Survey of Uranium Mill Tailings
Pile, Monument Valley, Arizona. Radiological Health Data and
Report 11:511-517 (Ocotber 1970).
(18) SHELLING, R. S. Environmental Survey of Uranium Mill Tailings
Pile, Mexican Hat, Utah. Radiological Health Data and Report.
12:17-28 (January 1971).
(19) U.S. ENVIRONMENTAL PROTECTION AGENCY. Radium-226, Uranium, and
Other Radiological Data from Water Quality Surveillance Stations
Located in the Colorado River Basin of Colorado, Utah, New Mexico,
and Arizona, January 1961 through June 1972. 8SA/TIB-24,-EPA
Region VIII, Denver, Colorado (July 1973).
45
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(20) INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION. The
Metabolism of Compounds of Plutonium and Other Actinides,
Adopted "May 1972, IC1P Publication 19. Perganmon Press, New York
(1972).
(21) DOUGLAS, R. L. AND J. M. HANS, JR. "Gamma Radiation Surveys at
Inactive Uranium Mill Sites." Technical Note ORP/LV-75-5.
U.S. Environmental Protection Agency, Washington, D.C. 20460
(August 1975).
(22) HAVENS, R. AMD K. C. DEAN. Chemical Stabilization of the
Uranium Tailings at Tuba City, Arizona. Report of Investigation
7288 (RI). Bureau of Mines, U.S. Department of Transportation,
Washington, D.C. (August 1969).
46
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I. FUEL SUPPLY
B, Transuranium Effluents from Re-Enriching
or Refabrlcating Reprocessed Uranium
-------
1.0 Introduction
Uranium feed material, either to an enrichment plant or to a
fabrication plant, which has been previously used as fuel, in a nu-
clear power plant may still contain trace amounts of radioactive
impurities after decontamination at fuel reprocessing.
Spent reactor fuel is typically allowed to decay either at the
reactor plant site or at the chemical reprocessing plant site a
minimum decay time of 150 to 180 days. The fuel is then dissolved
in nitric acid and processed by solvent extraction.
The UFg product from chemical reprocessing will contain small
quantities of fission products and transuranium-isotopes. Specifi-
cations have been published by the Atomic Energy Commission (1) which
indicate the maximum acceptable limits for radioactivity resulting
from these impurities. These are: gross alpha due to transuranium
isotopes — 1500 dis/tnin/ (g of U); gross beta due to fission pro-
ducts and transuranium isotopes — 10% of the beta activity of aged
normal uranium; and gross gamma due to fission products and trans-
uranium isotopes — 20% of the gamma activity of aged normal uranium.
Such processed uranium may then be sent to the enriching plant.
The above maximum acceptable limit for -gross alpha radioactivity can
be translated into the following typical distribution (assuming total
solvent extraction plus conversion decontamination factors (2) for
4?
-------
neptunium of 10 , plutonium - 10', and transplutonium - 10°);
0 0
neptunium - 9 x 10 alpha dis/min/(g of 0) , plutonium - 5 x 10
*5
alpha dis/min/(g qf JJ) and transplutonium - 1 x 10^ alpha dis/min/
(g of U). The actual alpha activity distribution will depend on
reactor type, fuel irradiation history, type of chemical process,
and the additional conversion and purification operations used in
converting uranyl nitrate hexahydrate to UFg, but should not vary
significantly from these typical values.
The above beta-gamma^radioactivity limits are based on gross
radioactivity measurements related to the background of aged normal
uranium. The beta activity limit is based on direct measurement of
the beta counting ratio, and ..therefore depends upon the variation of
counting efficiency with energy. The gamma specification is based on
a comparative measurement using aged natural uranium and a high pres-
sure ion chamber. A reasonable gamma comparison with natural uranium
can. therefore be equated to 20% of the gamma power of aged normal
uranium. The gamma power of aged normal uranium can be calculated
to be 269 MeV/sec/(g of U), which results in a gamma specification of
approximately 54 Me?/sec/(g of TJ) .
Typical reactor return material has shown the fission product
gamma radioactivity distribution given in Table 1.0-1. Technetiym
and uranium beta and uranium and transuranium alpha radioactivity
levels found are also indicated.
48
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TABLE 1.0-1
CALCULATED GAMMA RADIOACTIVITY DISTRIBUTION OF FISSION-PRODUCTS; GAMMA
AND BETA RADIOACTIVITY OF ALL FISSION PRODUCTS, AND ALPHA RADIOACTIVITY
OF TRANSURANIUM AND URANIUM ISOTOPESa(2)
Isotope
Ru-106
Zr-95-Nb-95
Cs-137
Ce-144
% of Gamma
75
22
1
1
Other fission products 1
Tc-99
U-237
c
Transneptunium
Np-237
U-232
U-233
U-234
U-235
U-236
U-238
Typical distribution
based on
gamma specification •
(Y MeV/sec/g U)
Radioactivity
(Ci/g> U)
Y Radioactivity
4Q.O
12.0
0.054
0.054
0.054
42.2 X 10
-10
9.3 X 10
-1-0
^6.9 X 10
-11
^6.9 X 10
-11
X 10
-11
(3 Radioactivity
3.16 X 10
2.41 X 10
a Radioactivity
2.43 X 10
4.32 X 10
9.01 X 10
4.70 X 10
7.59 X 10
1.71 X 10
2.88 X 10
3.14 X 10
-8
-6
-10
-10
-9
-11
-7
-8-
-7
-7
aPower reactor returns are based on an initial feed of 3.2% U-235,
specific power 30 MW/metric ton uranium, exposure 33,000 MW day/metric
ton, decay 180 days.
These fission products consist principally of Sr, Sb, Sn, and Te.
cPu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Cm-242, Cm-244
49
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These radioactivities can be used to determine the annual
inputs and system equilibrium concentrations at an enrichment plant
(Table 1.0-2). The technetiura-99 beta will contribute the remaining
beta radioactivity and is also included. Plutonium and neptunium
concentrations are based on the above specifications for transuranium-
isotopes in the reactor return material.
50
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2.0 Gaseous Diffusion Operating Experience
Gaseous diffusion operating experience, although of almost 30
years duration, has been very limited in terms of large throughputs
of power reactor returns. Although there has been considerable produc-
tion reactor material returned to the cascade, irradiation exposure
of that material has been ten- to twenty-fold less than that for power
reactors. Experience to date has indicated the following:(2)
1, A significant quantity of all non-uranium radioactivity
(neptunium, plutonium, and fission products) is retained in the
feed cylinder (UFg tank) and will be removed when and where the
returned cylinder is washed.
2. PuF, and NpF, are easily reduced and therefore removed by .
trapping with Cop2 MgF2, NaF, Cryolite,- etc. "*
3. Fission product removal (except technetium) by these traps may
also be significant. However, good data based on low-level radio-
activity feed materials have not been obtained.
4. Technetium, compared to other fission or alpha emission
products, is less likely to be removed by any process. Experience at
ORGDP* indicates that technetium release to the environment would be
10% of'feed to the liquid effluent and 1% of feed to the gaseous
effluent.
*0ak Ridge Gaseous Diffusion Plant
51
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TABLE 1.0-2
CALCULATED FISSION PRODUCT AND TRANSURANIUM ISOTOPE3
ANNUAL INPUTS AND EQUILIBRIUM SYSTEM6 CONCENTRATIONS(2)
Annual Input Equilibrium System
Isotope (Ci/year) burden
(Ci)
Ru-106 9.3 13.5
Zr-95-Nb-95 2.0 0.5
Cs-137 0.16 -0.0266T b
0.16 (1-e )
0.0266
Ce-144 0.16 0.17
£
Other fission products 0.16 0.7
Tc-99 (g only} 70.0 70.0Td
Np-237 0.9 0.9Td
Transneptuniura 0.5 0.5T
aBased on fuel specifications of Table 1.0-1.
Not an equilibrium condition since Cs-137 has a 26-year half-life
and true equilibrium would only be approached in 130 years-. Therefore,
activity depends on time, T (years of operation) .
cAssuming an average effective half-life of 3 years.
long half-life, never reaches equilibrium.
e8.75 MSWU
52
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5. Experience also indicates that other fission products and
alpha radioactivity release fractions should be no more than one; tenth
of that for teehnetium. Measurements of gaseous and liquid effluents
have failed to identify any other fission products. However release
fractions of 1% to the liquid effluent and 0.1% to the gaseous
effluent for other fission products will be used below to estimate
environmental releases.
6. Cobaltous .fluoride traps exhibit decontamination factors of
400 for neptunium and 10 for plutonium prior to feeding to the
cascade or conversion facility. Releases for the system after
trapping can then be proportioned to those exhibited for uranium in
ORGDP release data. Thus", alpha release fractions will be 4 X 10
to the liquid and 2 X 10" to the gaseous effluents for neptunium
-8 -in •
and 1.6 X 10 to the liquid and 8.0 X 10 1U to the gaseous effluents
for plutonium.
7. A large portion of the radioactivity entering a settling pond
will be entrained in the sludge of the pond.
53
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3.0 Estimated Radioactivity Releases
Releases to the environment can occur in three physical states
(gas, liquid, and solid). The bulk of the radioactivity will be
released as solids, either entrained on adsorbate or equipment
removed from service for disposal. Liquid waste will be generated
by rinsing (.decontamination) of recycled equipment. The first rinse
solution, which contains the bulk of the radioactivity, are saved to
be used as the dilute acid wash solution. Subsequent rinses are sent
to the primary holding pond.
Gaseous wastes can result from purge system venting, venting of
evaporator overheads at the uranium recovery facility, and venting of
decontamination hoods in the recycle facility. However, the exact
breakdown for retention and release factors for each step is not known.
One can only make assumptions based on experience with gaseous diffusion.
The limited experience available was used to arrive at the following
estimates (see Table 3.0-1) about gaseous, liquid, and solid discharges
for non-uranium radioactivity (2).
TABLE 3.0-1
ASSUMED DISTRIBUTION OF FISSION PRODUCTS AND TRANSURANIUM ISOTOPES
TO ATMOSPHERE, PRIMARY HOLDING POND, AND BURIAL GROUND
Isotope
Np-237
Other Transuranium
Tc-99
Fission Products
Fraction released
to atmosphere
2 X 10"
8 X 10-10
0.01
0.001
Fraction released
to primary
holding pond
-6
4 X 10
1.6 X 10"8
0.10
0.01
Fraction input
to burial ground
^1.0
VI. 0
0.89
0.989
54
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Primary enrichment plant sources of gaseous radioactive wastes
are the product and waste purge systems. Uranium particulates are
removed from these process streams by the high-efficiency-particulate
absolute (HEPA) filter, which has an efficiency greater than 99.95%.
Removal of gaseous uranium is achieved through the use of two chemical
traps in the product and waste withdrawal systems, in series, between
the cold trap and point of discharge into the air.
The .first trap contains sodium fluoride that provides for the .
adsorption of uranium and certain fission or alpha emitting products.
Through heating and proper valving, the trapped uranium may be
desorbed and subsequently returned to the cascade. The second trap
in the series contains alumina that is used for further removal of
uranium .prior to discharge of the gas stream to the atmosphere. This
trap is nohreversible and uranium recovery is accomplished by leaching
with nitric acid.
The fraction of the.feed made up of.reactor returns is passed
through cobaltous fluoride traps prior to being fed into the cascade(2);
the traps remove plutonium, neptunium, and a major fraction of the
fission products. These products are removed from the gas stream
by reduction with CoF2 to the tetraflouride forms that, being particulates,
are entrained within- the traps.
Quantification of potential gaseous effluents is difficult because
of uncertainties about the behavior of certain fission products in
feed cylinders, traps, piping, and equipment. In attempting to analyze
55
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possible releases to the environment, all assumptions, where necessary,
have been made so as to overestimate the magnitude of the source term.
Uranium and technetium releases were estimated by comparison with
operating experience and extrapolated to higher operating levels.
Fission product releases were based on current fission product
specifications, with releases being assumed proportional to that of
technetium, with the exception that a decontamination factor (DP) and/or
retention factor 10 times that for technetium was assumed. This
assumption is very conservative, since current experimental, investigations
indicate that the actual factor might be as high as 100 to 1000 (2)•
Releases of the alpha emitters, neptunium and plutonium, were estimated by
assuming an alpha specification of 1500 dis/min/(g of U) in reactor returns,
with a neptunium DF of 400 and a plutonium DP of 10 through Cop2 traps.
Once fed into the cascade, neptunium and plutonium are assumed to be
released to the environment in the same proportions as uranium.
The estimated constituents of an effluent under the above assumptions
are listed in Table 3.0-2.
It may be concluded that recycled uranium which has been re-enriched
will present no particular problem at the fabrication plant because most
of the impurities of higher isotopes have been taken out in the enriching
process, and could not make a significant contribution to an industry
limit of 0.5 raCi/GW(e) for alpha-emitting transuranics of half-life
greater than one year.
56
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TABLE 3.0-2
ESTIMATED RADIOACTIVITY RELEASED TO THE ATMOSPHERE'FROM
AN ENRICHMENT PLANTd
(Transuranic alpha specification = 1,500 dis/mih/g U}
Isotope Radioactivity
(Ci/year)/Gw(e)
U-232 2.75 X 10~8
-10
U-233 1.5 X 10,
U-234 3.25 X10~5
U-235 • 1.25 X 10"6
U-236 • 0.92 X 10"6
U-238 5.3 X 10~6
Transneptunium • • 3.3 X 10~
c ~10
Np-237 1.7 X 10
Tc-99 ' ' 4.5-X 10~4
Ru-106 6.0 X 10~6
Zr-95-Nb-95 1.25X10~
Cs-137 . 0.92 X 10~7
Ce-144 0.92 X 10~?
• -7
Other fission, products-. 0.92 X 10
>3
Relative to Tc-99, the retention of all fission
products in equipment or traps is greater by a factor of 10.
TU
Cobaltous fluoride trap decontamination factor for
Pu-239 = 10 .
cCobaltous fluoride trap decontamination factor for
Np-237 = 400.
d8.75 MSWU Plant
57
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If, however, recycled material goes directly from reprocessing
to fabrication, cleanup systems will have to be designed and installed
to collect the impurities as the material is converted from UFg to
U02 for blending and/or pelletizing. These systems should have
efficiencies and decontamination factors similar to those described
above for the enrichment plant. They would, therefore, be expected
to also reduce transuranium isotopes in the U02 to levels resulting in
negligible releases compared to the proposed standard of 0.5 mCi/GW(e).
REFERENCES
(1) 32 m 16289. (November 29, 1967),
(2) U.S. ENERGY RESEARCH AND DEVELOPMENT ADMINISTRATION. Environmental
Statement - Expansion of U.S. Uranium Enrichment Capacity, DRAFT
ERDA-1543 (June 1975).
58
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II. NUCULE. PQW1R REAG10RS
An Analysis of Control Options for N-16 Offsite
Skyshlne Doses at Boiling Water Reactors
-------
1.0 Introduction
The turbine system at a boiling water reactor (BOT.) is a
potentially significant source of radiation due to the presence of
nitrogen-16» a relatively short-lived (t =7.14 sec), high energy (2.75
Mev (1%), 6.13 MeV (69%), and 7.11 MeV (4.9%) gamma emitter in the
steam leaving the reactor. Nitrogen-16 is produced in the reactor
core by neutron activation of oxygen in water, and, although short-
lived, can be present in the turbine system in significant quantities
due to the rapid transit of steam from the reactor vessel' through the
turbine system and to the condenser. The result is a flux of direct
and scattered gammas which can result in high occupational exposure
rates in and close to the turbine building, as well as potentially
significant exposure rates to members of the public beyond site
boundaries near the turbine building.
59
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2.0 Sources
Detailed expositions of nitrogen-16 sources are presented In the
safety analysis report for the General Electric standard boiling water
reactor, the BWR/6 (3L) and for operating BWR's in a comprehensive
report recently released by General Electric (2). In these reports a
nitrogen-16 activity concentration of 50 MGi/gm of steam at the
reactor nozzles is assumed, based on experimental measurements of
contact dose rates on cross-around pipe sections of operating Ills.
Other analyses Q.,4) have assumed nitrogen-16 activities of up to 100
/iCi/gm of steam at the nozzles; however, this is probably due to the
desire for conservatism in the design of shielding.
In & typical modern boiling water reactor, steam flows directly
from the reactor nozzles through the main steam, header to the high
pressure turbine (HPT). Steam extraction is also made from this flow
path for steam to the steam jet air ejector (SJAE), feed water heaters
(I¥H), gland seal system, and the moisture separator/reheater units
(MS1H). Steam leaving the HPT is routed through the shell side of the
MSIH's, where it ie dewatered and reheated for injection into the low
pressure turbines (LFT). Steam extractions are also made at the HPT,
MSHR's, and in several places along the LPT for the various feedwater
heater stages (usually 6).
Typical delay times to and transit times through these components
are shown in Table 2.0-1, At a concentration of 50 MCl/gm of steam,
the nitrogen-16 source term at the nozzles is 100 Cl/sec. Thus, it is
60
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obvious that the potential exists for considerable equilibrium
activity to be present in these turbine system components.
Table 2.0-2 lists the calculated Inventories for the various
turbine building components. The doslmetrlc significance of these
sources depends on the shielding (both exterior and self-shielding of
components) as well as the geometry of the component layout. The
typical order of the dose significance by component is (a) moisture
separator/reheaters, b) intermediate piping, c) high pressure turbine,
and d) all other components.
61
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3.0 Turbine Building Configurations
The configuration in which components are placed in a turbine
building has undergone several changes in recent years. Several
different turbine manufacturers have supplied turbines for BWR reactor
plants and component layout has varied as a function of both turbine
manufacturer and of architect-engineer. Turbines have been supplied
by General Electric, Westlnghouse, and Kraftwerk-Union, for example,
and facilities using BWR's have been engineered by a variety of
architect-engineering firms. The major significant system design
changes have been with respect to the placement of moisture separators
and reheaters. Earlier BWR designs had vertically-oriented moisture
separators and separate reheaters located on the mezzanine level of
the turbine building (below the operating floor) as shown in Figure 3-
1 (5). Considerable shielding was afforded by the concrete structure
of the turbine building around these components, and, particularly
above, by the operating floor.
For a variety of engineering reasons, including increased
efficiency of turbine operation, reduction in building size, and
reduction in time of construction, recent designs have incorporated
horizontally-oriented combined moisture separators and reheaters
located above the turbine building operating floor level, as shown in
Figure 3-2, The high equilibrium nltrogen-16 activity levels in tube
and shell side of these systems, combined with the relative lack of
self-shielding, compared to that of the thick steel shells and massive
internals of turbines, result in these "exposed" MSKH's and their
62
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supply and return piping producing a potentially high gamma flux In
comparison with all other components. '
A system which can perhaps be considered an example of a "worst
case" is the combination of a General Electric BWB. with a Westlnghouse
turbine system. In this case the steam piping runs overhead from the
top of the HPT to the top or side of the MSBH. Silice there is ,
considerable nltrogen-16 activity in these pipes, they can provide a
significant additional source of gamma exposure beyond the MSKH's
themselves.
63
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4.0 DOBS Assessment
The gamma .flux existing at a point outside a turbine building due
to sources of nitrogen-16 inside is difficult to calculate. Gammas
may arrive at a. given point by direct paths, by scattering in
shielding and other components, or from air scattering, as shown in
Figure 3-3. The shielding geometry is complicated due to the variety
of component shapes and locations, and each component also has
different self-shielding factors for the gammas involved.
A variety of types of computer codes have been developed to
calculate the air-scattered contribution to the gamma exposure field
(see, for example, refs. .2.».6.».D« ^e potentially most accurate of
these are Monte Carlo transport codes. However, these models have not
been verified by ,EPA, and they are sufficiently complex and expensive
to prohibit performing such analyses on a case-by-case basis. No
discussion of analytical techniques for quantitatively analyzing these
exposure rates based on transport codes was undertaken, although the
results of some calculations performed by industry (5) provide the
basis for the present comparison of several options.
Insight into the relation between various shielding options and
anticipated dose rates can be obtained, however, through an
examination of existing shielding studies in conjunction with field
measurement studies. This examination indicates the principal
contributors to and magnitudes of potential doses and permits an
informed, if not detailed, understanding of what might be required to
reduce such doses.
64
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5.0 Shielding of Components
Because of the high radiation field resulting from nitrogen-16
activity, existing turbine systems are already veil-shielded. This is
not primarily because of consideration of doses beyond site
boundaries, but due to the need to comply with existing occupational
exposure limits. In order to restrict the extent of high radiation
areas adjacent to turbines and to allow more frequent or even
uncontrolled access to other areas in the turbine building, the
turbines and MSIH's are heavily shielded. Usually this shielding
consists of a thick concrete "shadow shield" surrounding the turbine
(as much as 4 ft thick), and upward extension of the turbine building
lower side walls (up to 3 ft thick) to shadow-shield the MSEH's.
While such shielding substantially reduces, the direct components of
the gamma flux, air-scattered contributions from gammas leaving the
unshielded top of the. turbines and MSKH's can still produce
considerable exposure rates. Thereforea often as a design option,
many recent designs have included concrete shields (up to 20" thick)
over the MSIH's and vertical steel plating running between the
turbines and MSKH's to reduce this air-scattered flux (see Figs. 5-
1,5-2). In order to assess the effectiveness of such additional
shielding as a means to reduce site boundary doses we have chosen to
analyze a variety of such shielding options for the turbine building
component configuration shown in Figure 5-1. The assumption is made
that concrete walls are already in place around the MSRH/turbine area
as shown to allow required access in the remainder of the turbine
65
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building area within applicable limits for occupational eKposure.
These walls are assumed to consist of three feet of reinforced
concrete, this thickness will provide an attenuation of approximately
99.7% of the incident gamma flux (neglecting buildup), leaving only
the scattered flux as a potentially significant contributor to the
off-site dose.
Such a characterization of skyshlne as the principal source of
exposure from nitrogen-16 at distances greater than a few hundred
meters from the turbine building is supported by a recent field study
performed at the Cooper Nuclear Station by EPA and ERDA (8). Cooper
station is a BWR with a Westinghouse turbine and horizontally-oriented
moisture separators located on the turbine building operating floor.
Field measurements were made by EPA in February, 1975, and by IRDA'e
Health and Safety Laboratory in April, 1975. Cooper is a reasonable
example of the "base" case turbine building discussed above, since
shielding consists of side walls only, although in this case these
consist of 3 ft of high density concrete. A significant finding of
the study was that nearly 100% of the dose measured was due to air-
scattered (skyshine) gamntaa. The contribution to dose of the direct
flux was negligible.
Referring to Table 5.0-1, it can be seen that for the base case
the total net equivalent activity above the turbine operating floor is
34 Ci. Out of this total, 21 Ci are associated with the moisture
separator/reheater and 10.3 CI are associated with the intermediate
piping.
66
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The shielding options considered, calculated doses, and
anticipated costs are presented in Table 5•0-2. These have been
derived in part from information provided the Agency by General
Electric (5). With these options and their associated dose rates as a
basis, and using Means 1975 Building Construction Cost Data (9)» we
have made independent cost estimates for installing the additional
shielding required by each of the options considered... The costs
presented do not include any additional basic building structure which
might be required within the turbine building to support the
additional weight of the shielding, because for most of. the cases
considered the additional weight involved does not, appear to require
any additional support beyond that already available in the basic
structure supporting the turbine and other components. The costs
presented here are appropriate to plants in the design stage, and
would not necessarily apply to retrofit situations.
All cases above the base case include the cost of poured-in-place.
reinforced concrete, which is supported by an assembly of steel
girders bridging the MSRH's between the exterior turbine shielding
wall and inside panel wall. The inside panel Includes steel columns
to provide additional support for the overhead assembly. The
dimensions required for each of two overhead shields 'are
conservatively estimated to be 140' long by 35* wide, The inside
panel walls are assumed to be 140' long by 25' high. The concrete for
exterior side walls and end walls is assumed to be already present as
the "base case." Costs of materials, installation, engineering,
67
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financing, overhead, and profit, were based on standard estimating
assumptions (10). Details of the estimation procedure used are
available upon request. Table 5.0-2 provides a summary of costs for
the various shield options, and Figure 5-3 displays annual dose at 500
meters vs. cost of shielding.
Doses are presented for the various shielding options both as
calculated by the industry and as projected from values measured in
the field. The data provided by General Electric was calculated using
a source term of 100 Ci/gm and has therefore been divided by two to
be consistent with the currently accepted source term of 50 MCi/gm.
In addition, the assumption of 100% occupancy, no additional shielding
by offsite building structures, and annual operation at 100% power are
considered to be unreasonably conservative assumptions for estimating
real doses to individuals at real sites. It is concluded, therefore,
that it should be readily possible to restrict the dose from nitrogen-
16 skyshine to a real individual located at reasonable distances from
the center of the turbine building for realistic occupancy times to
less than 2 mrem/yr. These dose levels should be attainable for no
more than approximately $250,000 and even these costs should be
incurred only in those few instances where actual site boundaries are
so close to turbine buildings as to create the possibility of
significant offsite exposures from nitrogen-16 sources.
68
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Table 2.0-1
N16 afflRACTEKCSTICS OF A STANDARD 3WR TURBINE
Conpca-ient
Main Steam Line and Header System
a. Beactx»r Nozzle to W&in Steam Header
b. Main Stream Header to HPT
High Pressure Turbine
tow Reassure Turbines
Maisture Separator Shell-Side (Steam)
a.. Inlet to Vanes
b. Vanes
c. Vanes to Outlet
Moisture Separator Shell-Side (Liquid)
{Vanes, Drain Trough) • •''
Decay Tine
at Inlet
(seconds)
0.00
2.09
3.18
5.86
4.29
4.64
4.73
4.64
Estimated
Mass Inventory
(Its)
8. 933x10 3
4. 464x10 3
13. 397x10 3
3.784xlf)2
7. 611x10 2
1. 256x10 3
a.ooxiry2
2. 119x10 3
3. 67 5x10 3
4. 059x10 3
Ilass Plowrate
(Ib/hr x 10" 6)
15.396
14.764
14.748
10.678
13.171
11.460
10.904-
1.712
Component
Transit Time
(seconds)
.2.-09
1.09
0.0924
0.257 .
0.343 -
0.0942
0,700
8.54
-------
labla 2,0-1 (Continued)
-g
o
Component
ttoisture Separator Drain System
a. Steam
b. Liquid
First Stage Reheat System
a. Supply Pipe - HPT to fube Inlet
b. Tubes
Second Stage Reheat System
a. Supply Pipe-Main Header to Tube
Inlet
b. Tubes
First Stage Keheat Drain System-
Second Stage Reheat Drain System
Decay Tine
at Inlet
(seconds)
4.73
13.18
3.27
4.33
2.09
3.73
37.3
37.8
Estimated
Mass Inventory
(Ibs)
2.058xl02
6.424xl03
6.630xl03
2.80xl02
s.Biixio3
6. 091x10*
Mass Flowrate
(Ib/hr x It)"6)
0.5554
1.712
0.7011
0,7011
0.6145
0.6145
0.7011
0.6145
Ganponent
Transit Time
(seconds)
1.06
33.0
1.64
34.0
-------
Table 2,0-1 (Continued)
Qcnponent
Piping System - HPT to MS/MR
Piping System - MS/BHR to U?T
a. NB/RHR to civ
b, ciy
c. CIV to 1PT
First Stage FWH and Extraction System
a. Extraction Point 4
b." Extraction Point 5
Second Stage FWH and Extraction System
Third Stage PMH and Extraction System
Fourth Stage FWH and Extraction System
Fifth Stage FWH and Extraction System
(Excluding MS Drain System)
Decay Time
at Inlet
(seconds)
3.27
5.43
5.66
5.75
6.12
6.12
6.12
6.12
6.12
3.18
Estimated
Mass Inventory
(Ibs)
3. 717x10 3
6.857xl02
2,852xl02
2.812xl02
1.252X103
Mass Flowrate
(Ib/hr x 10"6)
13.171
10.904
10.678
10.678
0.1016
0.6017
,0.6301
0.7344
0.4016
0.0126
Component
Transit Time
(seconds)
,1.02
0.227
8.0962
0.0948
-------
Table 2.0-1 (Continued)
Oraponent
Sixth Stage BWH and Extraction System
(Excluding Bsheater Drain Systems)
Condenser
(Ixclixttng return from FW Turbine)
Hotwall
(Bxalufling return from FW Haters, etc.)
SJAE First Stage System
a. Off-Gas
b. Driving Steam Supply Line
c. First Stage Driving Steam
Itecoiribiner System
(Second Stage Mr Ejector Driving Steam)
Gland Seal Systan
a. From HPT
b. Fran Valve Stan
Feedwater Turbine System
Decay Time
at Inlet
(seconds)
3.27
6.12
"36
"7
2.09
4.33
4.33
3.27
3.18
5.66
Estimated
Mass Inventory Mass Flowrate
(Ibs) (lb/hr x 10~B)
0.857
8.207
0.0016
8.207
0.0016
1.12X101 0.0180
0.0080
0.0100
0.0186
0.0029
0.2259
Ocnpment
Transit Time
(seconds)
"30 (liqui*
* 1 (gas)
2.24
-------
Table 2.0-2
N16 Inventories For A Standard BWR Turbine
Conponent
Main Steam Line and Header System
High Pressure Turbine
Low Pressure Turbines (1)
MDisture Separator and Keheater Shell-side Steam
Moisture Separator Shell-side Liquid
Moisture Separator Drain System
First Stage Raheat System (2)
Second Stage Iteheat System (2)
First Stage Reheat Drain System (3)
Second Stage Reheat Drain System (3)
Intermediate Piping System - HPT to MS/BH
Intermediate Piping System — MS/SH to I£T
First Stage - FWH & Extraction System. (4)
Second Stage - FWH & Extraction System (4)
Third Stage - FWH & Extraction System (4)
Foiirth Stage - FWH & Extraction System (4)
Fifth Stage - Pvffi & Extraction System
(Excluding Moisture Separator Drain
System activity Listed above) .
Sixth Stage - FWH & Extraction System
(Excluding First and Second Stage Reheat
Drain System Activities Listed above)
N-16
Inventory
(Curies)
263
6.3
9.8
53
41
56
33-
32
1.4
1.1
59
17
26
23
27
15
.6
,42
73
-------
Table 2.0-2 (Continued)
N-16
Inventory
Component (Curies^
Condenser 287
(Excluding Residual Activity Returned from
Feedwater Turbine).
Hotwell 18
(Excluding Residual Jk^dvity Beturned from
Feedwater Heaters and Gland Seal System)
SJAE First Stage System (5) .6
SJZffi Off-gas System .4
Gland Seal System (6) 1.0
F.W. Turbine System (6) 8.8
•total 1822.0
Notes
(1) 6-Flcw machine.
(2) Includes inventory in liquid and steam in reheat tubes and in steam
supply line.
(3) Includes total inventory beyond reheater outlet.
(4) Includes total inventory beyound extraction point. Distribution of this
will depend on equipment arrangement and sizing,
(5) Includes inventory in steam supply line.
i
(6) Includes total inventory beyond inlet at steam supply line.
74
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Table 5.0-1
Turbine equipment typical total and net
16N inventories (Ci) for a 1200 fife plant.
Main Steam lines
HP Turbine
HPT to MS/R Piping
MS/R
MS/R to LPT Piping
IP Turbines
FW Heaters & Extraction
Condenser
Hotwell
SJAE & Gland Seal
IW Turbine
TOESL
CPERfiUNG FDOOR
260
6
60
220
17
10
130
290
18
2
9
GROSS
5 •
6
2
150
17
10
—
—
—
—
_ u.
NET
EQPTOMSSIT
1.6
0.3
1.3
21'
9
0.5
—
—
• —
—
MMMM*
1022
190
34
75
-------
Table 5,0-2 Summary of Shielding Cost Estimates
ShieldDesign Estimated Dose at 500
Estimated Cost of Shielding (k$)
Meters (mrem/yr) , Based
_. M on Calculational Models
r-i
£
-------
NN\
n—ir-n
BWR TURBINE BUILDING LAYOUT WITH
MOISTURE SEPARATORS LOCATED
BELOW THE OPERATING FLOOR
FIGURE 3-1. TYPICAL COMPONENT LAYOUT IN EARLY BWR TURBINE BUILDING DESIGNS.
(5)
-------
N\l\
ROOF SLAB LOCATION
WHEN USED
LOW PRESSURE
FEED WATER
HEATERS
3-2. TYPICAL COMPONENT LAYOUT IN CURRENT BWR TURBINE BUILDING DESIGNS.
.(5)
-------
AIR-SCATTERED
DIRECT (EQUIPMiNT
BELOW FLOOR)
DETECTOF
FIGURE 3-3. CONTRIBUTIONS TO DOSE RATE FROM N-16 IN TURBINE BUILDING COMPONENTS.
-------
I m
10"
TO 19,6'
ABOVE
MOISTURE SEPARATOR
REHEATiRJf
STEEL '
LOCATION OF INSIDE PANEL
WHEN USED
HP TURBINE
STiEL
19.B'
LOCAfiON OF INSIDE PANEL
WHEN USED
"V
MOISTURE
SEPABATOB-
\' HEHEATER
TBSvT
\ LOCATION OF ROOF
N SLAB WHEN USED
V T
8!4" TO
19.5'
BK" TO
ABOVE
1QK-
3,3-
STL,
STL.
r
FIGURE 5-1. TOP VIEW OF TURBINE COMPONENT LAYOUT SHOWING TYPICAL "ACCESS" SHIELD
DESIGN ALONG WITH VARIOUS SHIELD OPTION. (5)
-------
.» ,21111 PtHTR. DRAIN RIO,
Jjr-JT-17 |T-I6 3"-0!|
FIGURE 5-2. Transverse sectional view of Nine Mile Point 2 nuclear plant turbine building,
showing shielding of moisture, separators and turbines, v^
-------
8.0 J
1
i-l
O
4J
3
O
T<
-------
REFERENCES
(1) GENERAL ELECTRIC COMPANY. BWR/6 Standard Safety Analysis Report,
HEBO 10741.
(2) ROGERS* D.R. BWR Turbine Equipment Nitrogenrl6 Radiation
Shielding Studies, General Electric Report NEDO-20206 (December
1973).
(3) STONE AND WEBSTER ENGINEERING COMPANY, Radiation Shielding Design
and Analysis Report - Nine Mile Point Nuclear Station Unit 2, RP-6
(January 1974).
(4) PUBLIC SERVICE ELECTRIC AND GAS COMPANY OF NEW JERSEY. Newbold
Island Nuclear Station Preliminary Safety Analysis Report
(February 1970).
(J|) Information provided EPA by General Electric and Bechtel
Engineering Staff (January 1975).
(6) WOOLSEN, W.A. , A.E. Profio, D.L. Huffman. Calculation of the Dose
at Site Boundaries from Nitrogen-16 Radiation in Plant Components,
JRB 72-507 LJ, JRB Associates (December 1972).
(7_) WAHDi J.T., Jr. A Dose Bate Kernel for Air-Scattered Nitrogen-16
Decay Gamma Rays, Ph.D. Thesis, University of California, Berkley,
California.
(8) PHILLIPS, C.» W. Lowder, C. Nelson, S. Windham, and J. Partridge.
Nitrogen-16 Skyshine Survey at a 2400 MH(t) Power Plant, EPA-
520/5-75-018 (December 1975).
(9) GODFREY, R.G., Editor. Building Construction Cost Data 1975, 33rd
Ed., Robert Snow Means Company, Inc. (1974).
(10) The following markups were applied to materials and
installations: 251 overhead and profit, 2.5% engineering, 10%
contingency. A short term financing factor of 1.375 was then
applied to the total, representing a 10% per annum financing cost
over a period of three years.
83
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III. NUCLEAR FUEL REPROCESSING ,
A. Control of Iodine Discharges From
Nuclear Fuel Reprocessing Facilities
-------
I'O Introduction
Iodine-129 in spent fuel has been recognized as a potentially
significant environmental contaminant, and efforts have been made in
the past to control the discharge of this;species of radioactive
iodine. These efforts were only partially successful, however, and it
has become increasingly apparent that improved control of long-lived
radioiodine discharges from fuel reprocessing facilities is necessary
(1^2)' Current estimates of the costs and control efficiencies of a
variety of Improved control systems for iodine-129 and iodine-131 are
reviewed below. The benefits to be gained by reducing the
environmental dose commitments associated with releases of iodine-129
through installation of such systems are then set forth. Finally, the
level of cost-effectiveness of each of the control options is
determined.
85
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2.0 Source Terms for Iodine
The quantities of iodine-129 and iodine-131 present in spent
uranium fuel have been previously reported, based on calculations
using the computer code ORIGIN (.3). These values, expressed in curies
per metric ton of heavy metal in the fuel, are:
1-129: 0.04 Ci/MTHM
1-131: 0.9 Ci/MTHM
for the following fuel parameters, used in this report:
Burnup = 33,000 Mtfd/MTHM
Average Specific Power = 30 MW/MTHM
Cooling Time = 160 days.
It is assumed that a light-water-cooled power reactor operates at
33% thermal efficiency, producing approximately 33 MTHM of spent fuel
with this burnup for each gigawatt-year of electric power(JGW(e)-yr},
and that a typical fuel reprocessing plant has a throughput capacity
of 1500 MTHM per year. Such a plant would be capable of processing
the spent fuel from about 45 such reactors each year.
If no iodine control systems were installed at a 1500 MT plant,
the number of curies discharged annually would be:
1-129: 60 Ci
1-131: 1,400 Ci
It is assumed that these contaminants are discharged to the
atmosphere, rather than into liquid pathways, since currently
projected plants use complete recycle of process liquids and thus no
liquid discharges are planned.
86
-------
Although the source term for 1-131 could theoretically approach
•1400 Ci per year, it is highly unlikely that such quantities will be
available for discharge in actual operations because of its relatively
short half-life (8.08 days). Even if all spent fuel was processed at
160 days cooling time, any delay of iodine-131 in the various inplant
processes or off—gas streams would permit additional decay and reduce.
the quantity available for discharge. Other factors that would reduce
the quantity of iodine-131 available for discharge include: a) the
existing large backlog of spent fuel, which indicates there is no
need, at least
-------
3.0 Control Technologies for Iodine at Reprocessing Plants
The control of iodine at reprocessing plants is a significant
technical challenge. During the last few years a number of promising
systems for control of iodine in gaseous waste streams have been
investigated and most are now in various stages of final demonstration
for commercial use. The principal remaining problem, as pointed out
in the previous EPA report concerning fuel reprocessing (1), is that,
until recently, inadequate attention has been given to the control of
iodine in low-level liquid waste streams. Any iodine present in these
liquid streams, whether from off-gas scrubber solutions or from other
sources, can potentially be discharged to the environment because of
its high volatility. Evaporative processes are used to reduce the
volume of these low-level liquid wastes and to provide for discharge
of tritium to the atmosphere. Such processes will, of course, also
drive off any iodine present for subsequent discharge to the
atmosphere, and systems developed for removal of iodine from gaseous
streams are not, in general, applicable to evaporator discharges
because of their high water content.
A simplified schematic of waste streams appropriate to the
discussion of iodine control systems for current designs of
reprocessing plants is shown in Figure 3-1. Most of the iodine
present in spent fuel is released to the off-gas system during the
fuel dissolution and initial processing steps. The fraction released
to the off-gas has been estimated at no less than 90% (5). The
balance is collected in liquid waste streams. The off-gas system for
88
-------
a specific plant will .not necessarily be designed just,as shown In.the
schematic, since the detailed design can vary due to the order in
which contaminants are removed. For example».it may be advantageous
to remove the oxides of nitrogen from the dissolver off-gas stream
before dilution by process off-gas inputs.
The chemical form or species is an important characteristic of the
iodine when considering cleaning efficiencies, environmental
transport, and iodine dosimetry. In general, it is believed that
iodine evolved during the dissolution process will be in the elemental
form (.7). However, any iodine discharged to the off~gas system during
or following the separation processes is considered likely to have a
large organic component (8). The relative fractions of iodine evolved
from the dissolution process step and from the various subsequent -
separation processes is not known, nor is the organic component of
either fraction (5). Estimates of these fractions vary widely (5_,J5)
and these differences will probably not be resolved until studies are ,
conducted during actual operations of a large facility (9). For the
purposes of this analysis it is assumed that 90% of iodine is
discharged to the off-gas system, with the balance going to liquid
waste streams (5). The fraction .of. the iodine discharged to the
atmosphere following all control .systems is assumed to be about 50%
organic and 50% elemental. Factors contributing,to an expectation of
a significant organic component of the final discharges are: a)
iodine from the low-level liquid pathway has passed through organic
processing steps and thus can be .expected to have a significant
89
-------
organic component, b) iodine in the off-gas stream is expected to
contain a significant organic contribution from separation processes,
and c) most iodine cleaning systems are more efficient in removing
elemental than organic iodine, and thus selectively allow passage of
organic iodides.
Table 3.0-1 summarizes iodine control system capabilities and
costs. The iodine control system J)F*s assumed are, for the most part,
those used in a recent study of effluent controls for fuel
reprocessing by ORNL (4). The difference in control efficiencies for
1-129 and 1-131 shown in Table 3.0-1 for Ag-Z and macroreticular
resins are due primarily to the differences in half-lives of these ,
radionuclides, as discussed in detail by Davis (6). This difference
is to be expected in any system which relies upon delay as part or all
of its operating principal. Thus, it is essential to both isolate and
contain long-lived radionuclides to insure that they will not
eventually re-enter a discharge stream. A brief description of each
of the radioiodine control systems is given in the following sections.
3.1 Caustic Scrubbers
Caustic scrubbers are widely used in the chemical industry to
remove contaminants from off-gas streams (10). They have been used in
*
the nuclear industry to control both ruthenium and iodine (11). Tests
have indicated that DF's of 100 and greater for elemental iodine are
attained (11), but DI's are less for organic iodine species. The
fraction of organic iodine in the primary off-gas stream is not known,
but is predicted to be low (5_). It has been assumed that the organic
90
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fraction Is less than 10% and that caustic scrubbers will, therefore,
operate routinely with a removal efficiency of no less than 90%.
Capital cost estimates for a caustic scrubber, are abstracted from the
OBHL work (4).
3.2 Mercuric Nitrate Scrubbers
Mercuric nitrate-nitric acid scrubbers have been used at the ABC
(now IRDA) reprocessing facilities at Idaho Falls to control the
discharge of iodine. While this type of scrubber removes both
elemental iodine and organic iodides, .tests have indicated that it is ,
also more efficient in removing iodine in the elemental form (12).
Based on the predicted relative fractions of organic iodides present
(5), it is assumed to remove about 90% of all iodine from the off-gas
stream (12_,]L3). Costs for mercuric nitrate scrubbers are expected to
be similar to those for caustic scrubbers (l^j^Q) •
3.3 Silver Zeolite Adsorbers ........
Silver zeolite adsorbers have not been used to treat reprocessing
plant off-gas, but are scheduled to be installed in future plants.
Most of the development work for this system was conducted at the
Idaho National Engineering Laboratories (14). Silver nitrate is
impregnated into an alumina-silica matrix and the resulting material
is arranged' in a relatively deep bed, since a longer residence time of
the iodine in the adsorber appears to enhance its efficiency. High
removal efficiencies have been observed for all chemical species of
iodine using this process (14). Although considerably higher values
are reported for small-scale systems, OHNL assigned a BF of 10 for
-91
-------
1-129 and a DF of 100 for 1-131 for a silver zeolite adsorber, pending
the development of additional data for plant-scale usage (15^16), and
these conservative values have been assumed here. The costs are
subject to some uncertainty related to the loading rate of the system
and thus the quantities of silver required (l.».2Q) •
3.4 Macroreticular Resins
Adsorption of iodine from both neutral and slightly acidic
solutions on macroreticular resins has been shown to be about 99%
efficient in laboratory studies (1?). However, performance of this
system has not been demonstrated in commercial-scale practice and,
until proven under operating conditions, a conservative DF of 10 for
1-129 and a DF of 100 for 1-131 are assigned. Costs for this system
are estimated to be small (20).
3,5 Suppression in Evaporator by_ Mercuric Nitrate
Mercuric nitrate, when added to liquid evaporators, will suppress
the evolution of iodine into the overheads. The Barnwell Facility
includes provision (18) for this method of iodine emissions control
from liquid waste streams. Yarbro has estimated a DF of 2 to 10
across the waste evaporators, including the final vaporizers, for this
addition (5). A conservative value of 2 is assumed for this analysis.
Costs are estimated to be similar to those for a macroreticular resin
system.
92
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3/6.Advanced Systems
i , . Figure 3-2 displays a simplified schematic of an advanced iodine
control system. The basic principle of this system is to force
essentially all of the iodine into the off-gas system so as to avoid
the difficulty of removing iodine from liquid streams, and then to use
highly efficient systems to remove and retain iodine from the off-gas.
In the schematic this objective is achieved by using an iodine
evolution process at the dissolver to drive the iodine into the off-
gas, and the iodox system to efficiently remove the iodine from the
off-gas. The voloxidation step is primarily used for tritium control.
.However, a significant fraction of both the iodine and krypton present
in the spent fuel will also be driven off by this process. After
tritium has been removed from the voloxidation off-gas, this stream is
routed to the dissolver off-gas stream for subsequent krypton and
iodine removal.
The iodox process itself effectively scrubs both elemental and
organic iodine from off-gas streams with concentrated (~2QM) nitric
acid (]_>¥Q • Laboratory-scale studies have indicated that DF's in
excess of 10,000 for methyl iodine have been obtained in multi-staged
bubble-cap columns (8). The efficiency with which iodine is scrubbed
from off-gas streams with nitric acid is dependent on the oxidizing
x
power of the concentrated nitric acid, which converts the volatile
iodine species to the nonvolatile HI308 form. The capital cost
estimates in Table 3.0-1 are abstracted from the»OBNL work (4) ; there
is no provision made at this time for the additional cost of a
93
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fractlonation system to permit recovery of the acid at low
concentrations for recycle to the dissolver and iodox systems.
The voloxidation process effectively removes such volatile fission
products as iodine and krypton from sheared fuel, by heating the fuel
to about 550 °C in air or oxygen to release these fission products by
thermal evolution or by oxidation (21). The process equipment would
consist of; a) a rotary kiln to oxidize the fuel, b) a recombiner to
form tritiated water, and c) a drier to collect the water and separate
it from iodine and krypton which then flow to the iodox equipment
(20). Laboratory-scale tests1 with highly-irradiated sheared fuel show
that up to 75% of the iodine and 45% of the krypton are volatilized.
The costs shown are based on the OKNL work (4).
OENL is currently conducting development work on these advanced
systems. Capital cost estimates and projected DF's are abstracted
from their recent summary. OEKL has projected that these systems will
be demonstrated and available for installation in new reprocessing
plants by about 1983, assuming that an orderly program of engineering
development, construction, and demonstration is pursued (4).
94
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4.0 Cost Evaluations
Estimated capital costs and annual operating costs for the various
iodine control systems described are listed in Table 3.0-1, The
Agency's capital cost estimates for iodine control are based on
work at ORNL (4) and recently released actual cost figures for
mercuric nitrate scrubbers and silver zeolite beds at the Barnwell
plant (20). Both of these analyses considered iodine control as
applied to a 1500 MTHM per year fuel reprocessing plant similar to
the Barnwell plant in design features. Therefore, the Agency
feels that costs from the Barnwell experience are more appropriate
for use in determining the cost-effectiveness of iodine control
systems. In general operating costs have been estimated since no
operating experience is available. Storage costs and disposal
costs have been neglected in the analysis since meaningful data
cannot be developed until a determination is made on the final
disposition of fuel cycle waste. However, since the additional
iodine-129 waste that the proposed standard will require be
collected is very small compared to that which will be collected
under current practices, the incremental cost of storage and
disposal are expected to be insignificant.
95
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5.0 Poses and Potential Health Impact Attributable to Iodine
Discharges from Fuel Beprocessing
lartial cumulative environmental dose commitments to the thyroid
and estimated potential health effects attributable to discharges of
iodine-129 from a model 1500 MTHM/yr plant were calculated using the
specific activity method (1)» and are presented in Table 5.0-1. These
values represent a partial assessment of the total potential, dose and
health impact of iodlne-129 in that the period of assessment following
release of this extremely long-lived material (17 million years half-
life) i«3 limited to 100 years. Dose commitments were cumulated for
releases over an assumed control equipment lifetime of 20 years
commencing in 1983. These partial cumulative environmental dose
commitments and their associated health impacts are shown for
representative values of overall plant decontamination factors
obtainable using the control methods described above. The dose-effect
assumptions used were derived from more recent values (22_,j24) than
those used in the original analysis (1); a population age weighted
value of 60 thyroid cancers per million rems to thyroid was used.
Health effects may also result from exposure of local populations
immediately following release of both iodine-131 and iodine-129, in
addition to the long-term effects described above. Using methods
described previously (1) and short term pathway parameters noted
below, it is estimated that uncontrolled release of 1400 Ci/yr of I-
131 could result in 35 health effects and the release of 60 Ci/yr of
iodine-129 could result in 30 health effects over a 20-year period of
plant operation commencing in 1983. These values should be added to
96
-------
Listed in. Table 5.0-1 to obtain a .complete estimate of potential
health effects attributable to the uncontrolled release of radioactive
iodines for the first 100 years following release.
In addition to the population doses and impacts calculated above,
maximum potential thyroid doses to individuals may also be
significant. Tables 5.0-2 and 5.0-3 list calculated maximum
individual thyroid doses from iodine-129 and iodine-131 discharges for
a variety of age groups and release fractions. , The values for iodine-
131 were calculated using dose conversion factors previously described
(23). Dose conversion factors for iodine-129 were based upon those
used for iodine-131, corrected for differences in pathway and
dosimetry dependent upon half-life and effective energy of decay
products (1). It is assumed that 501 of the iodine released is in
elemental form and 50% is in organic form, and that X/Q is equal to 5
—s , ' * -; . • '
x 10 sec/nr. Although specific sites could vary significantly from
this assumption; it is expected that site selection criteria for fuel
reprocessing facilities will reflect particular attention to
minimization of the possibility of dose to the thyroid of'nearby
individuals.
97
-------
6.0 CoBt-effectiveaess Considerations
Analysis of the options available for control of iodine is
complicated by a) the multitude of alternatives available, and b) the
variability of the current stage of development of the different
processes. It is clear that iodine evolution and the iodox cleanup
process represent the most effective improvements over the basic
cleanup of gas streams by scrubbers (with or without backup by Ag-Z)
and the cleanup of liquid waste streams by macroreticular resins
characteristic of current design practice. Unfortunately, reduction
to commercial practice of these systems has not been projected to be
completed before 1983. However, with the exception of some secondary
systems for liquid cleanup (HgN03 suppression and, in the case of
iodine evolution, macroreticular resin), all of the options display
good cost-effectiveness, as shown in Table 6.0-1. It should also be
noted that a second scrubber has apparently better cost-effectiveness
than does Ag-Z, which is more appropriate as a polishing method for a
bulk method of iodine removal. Finally, cost-effectiveness has been
determined on a dollar per man-rem thyroid basis, shown in the last
column of Table 6.0-1. It is readily seen that the cost of just
about all systems listed, in terms of dollars spent to avoid one man-
rem to the thyroid, is rather small, especially when compared to the
NRC's interim value of $1,000/whole body or thyroid man-rem applicable
to light water power reactors (26).
Although Table 6.0-1 does not display overall plant
decontamination factors, it can be seen from Tables 3.0-1, 5.0-2, and
98
-------
5.0-3 that conforman.ee with the proposed thyroid dose limit of 75
mrem/yr can be readily achieved through use of a variety of
combinations of systems exhibiting DF's of 100 or more. However,
eonformance with the proposed limit of 5 mCi/Gff(e)-yr or 1.4 kg/yr for
iodine-129 (0.225 Ci/yr from a 1500 MTHM facility) by 1983 will
require a plant DF of no less than 300. This would be readily
achieved by utilization of iodine evolution followed by the iodox
process. Successful achievement'of this level of cleanup without use
of the iodox process will depend to some, extent upon future operating
experience with less sophisticated systems. Present estimates of
their performance are quite conservative because of a paucity of
operating experience, especially with respect,to 1-129. However, it
is anticipated and highly probable that DF's greater than 300 for
iodinei-129 could be achieved by 1983 using appropriate combinations of
scrubbers and Ag-Z, since a variety of options are available for
improving, if necessary, the conservative levels of performance
currently projected. These include a) tandem operation of systems, b)
additives, such as thiosulfate to caustic scrubbers, to improve their
'efficiency (33) c) use of iodine evolution to reduce the fraction of
iodine in the liquid waste stream and increase the efficiency of
scrubbers by reducing the organic content of the gas streams, and d)
demonstration of more efficient cleanup of liquid streams than
currently assumed.
99
-------
Table 3,0-1 Iodine Control Cost Summary
(a)
Process
1,
2.
3.
H 4.
C3
O
5,
6,
A.
B.
Caustic Scrubbing
Mercuric Nitrate Scrubbing
Silver Zeolite Beds
Adsorption on Macroreticular
lias itis
Mercuric Nitrate Suppression
lodox
VoloxidatioiT ^
Iodine Evolution
Capital
DP Cost (M$)
10
10
10 (1-129)
100(1-131)
10 (1-129)
100(1-131)
2
10,000
4«
200(a)
0,60
0.60
1.25
0.4
0.4
2.07
2.74
0.75
Annual
Operating
Cost (M$)
0,04
0.12
0.15
0.04
0.04
0.22
0.29
0.08
Present Worth:'*
Operating Cost
(M$)
0.34
1.02
1.28
0.-34
0.34
1.87
2.47
0.68
') Total
Present ,
Worth (M$)
0,94
1.62
2.53
0,74
0.74
3.94
5.21.
1.43
(a") All costs are expressed in millions of 1975 dollars.
(b) 10% & 20 years; present worth factor - 8.51356
(c) Total Present Worth. = Capital Cost + (Annual Operating Cost x 8.51356)
(d) 'This system is not installed, primarily, to facilitate iodine control, and is listed only
for coifileteness.
(e) These values do not represent actual DP's but represent a process efficiency factor.
-------
Table 5,0-1 100-Year Cumulative Environmental Dose Commitment and Estimated Health Effects
Attributable to Release of 1-129 from a 1500 MTHM/yr Reprocessing Plant (a,b)
Source Term (Ci/yr)
60
6
1.2
0.6
0.2
0.06
DF
1
10
50
100
300 .
1000
Thyroid Dose Commitment (man-kilorems)
1700
170
34
17
5.7
1.7
Health Effects
100
10
2 .
1
0.33
0.1
(a) Partial environmental dose commitment and health effects are calculated for 100 years
following release only and for a plant operating life of 20 years, eomnencing in 1983
(b) Doses and health effects do not include short term, local impact of either iodine-129
or iodine-131. These are estimated to be 30 and 35 health effects, respectively, for
a DF of 1. ' .
-------
Table 5.0-2 Maximum Individual Thyroid Doses from 1-129 Discharged from a 1500 MTHM/yr Reprocessing Plat
(for average consumptive levels)
DF
1
10
50
100
300
1000
Source Term (Cl/yr)
60
6
1.2
0.6
0.2
0.06
Maximum
6 month
1100
110
22
11
3
1
Individual 1-129 Thyroid Dose
old 4 year old 14 year
1600
160
32
16
.? 5.3
.1 1.6
600
60
12
6
2
0
(mrem/yr)
old adult
140
14
2.8
1.4
.0 0.47
.6 0.14
(a) The elemental iodine fraction is assumed to be 50%.
Q
(b) Atmospheric dispersion coefficient equals 5 x 10 seconds per cubic meter; only the milk
pathway is considered.
-------
Table 5.0-3 Maximum Individual Doses from 1-131 Discharged from a 1500 MTHM/yr Reprocessing Plant
(for average consumptive levels)
DF
1
10
100
300
500
1000
10000
Source Term (Ci/yr) ^ ^
1400
140
14
4.7
2,8
1.4
0,14
Maximum Individual 1-131 Ihyroid
6 month old 4 year old 14
1900
190
19
6.3
3,8
1.9
0.19
2300
230
23
7.7
4.6
2.3
0.23
Dose (mrem/yr)
year old adult
430
43
4.3
1,4
0.86
0,43
0.043
110
11
1.1
0.37
0.22
0.11
0,011
(a) Fuel cooled for 160 days before processing; the elemental Iodine fraction is assumed to be 50%.
<3
(b) Atmospheric dispersion coefficient equals 5 x 10 seconds per cubic meter; all pathways are
considered. . •
-------
fable 6,0-1 Cost Effectiveness of Iodine Control Systems
at Fuel Reprocessing Plants
Cost
Increment
System (M$)
A.' Gaseous Phase Iodine
1. Without Iodine Evolution (a) HgKL Scrubber
(b) lodox (no scrubbers)
(c) Second Caustic Scrubber
(d) Silver Zeolite
(one scrubber)
f> 2. With Iodine Evolution (a) HgNO Scrubber
_^, (b) lodox (no scrubbers)
(c) Second Caustic Scrubber
(d) Silver Zeolite
(one scrubber)
B. Liquid Phase Iodine
1. Without Iodine Evolution (a) Macroreticular Resin
(b) Mercuric Nitrate Suppression
2 . With Iodine Evolution (a) Macroreticular Resin
(b) Mercuric Nitrate Suppression
1.62
3.94
0.94
2.53
3.05
5.37
0.94
2.53
0.74
0.74
0.74
0.74
Health
Effects
Averted
134
149
13
14
148
164
15
15
15
0.7
0.8
0.03
Cost per
Health
Effect
(M$/HE)
0.012
0.26
0.072
0.181
0.021
0.033
0.063
0.169
0.049
1.06
0.93
24.7
Cost per
Unit Thyroid
Dose
($/man-rem)
0.71
1.6
4.3
11
1.2
1.9
3.6
9.7
2.9
62
53
1,450
* Add incremental iodine evolution cost
-------
907, of Jodine (7) or® @
Krypton ^
I"*" Control • •*-• — • — ••Hfc. Scrubbers ••*• Silver
f J Zeolite
1 „, * * „,.,,„
i ! S jr
« i i ° "™""" "
«l «, " Iodine
CQ| Up CO*
!lf vil Si S|qrage
»H| 'Hi 4-?
tj" m ! w
• °j °l ' °|
•__ _ PRODUCT
SHEftfi "^ DISSOLVE *" PROCESS STEPS "*" LQADOUT
System DF ©
•) Caustic Scrubber 10 V r Low Level Liquid
^HgN03 Scrubber 10 (10% of Iodlne)
Si^^lwr 7pn1if*» 1^^"T.^9Q^ If
100(1-131) "" " "
DMacroreticular .10(1-129)
Resin 100(1-131) ^fSL
DHgNO. Suppression 2 ^plp,, - (OPTION)
I'
INTERMEDIATE
l£VEL HASTE
STORAGE
J
/
—*" HEPA " — "— "~W \
Filters / \
/ Stack\
"7
•
i
i
H
I
. • I
Macroreticular r
Resin I
„ „_. f
r 1
Final
"* Vaporize
®
sr
Figure 3-1. SIMPLIFIED SCHEMATIC OF CURRENT IODINE COHTSOL SYSTEMS AT REPROCESSING PLANTS
-------
®
G.HFAH —Hi
f"
1
o*
I
Tritium
Control
1
*
I
TOLOXIDATIOM
System
® Silver Zeolite
<2» Macroreticular
Resin
® HgNOq Suppression
© lodox
©Iodine Evolution
99.5% of Iodine / \ @
\
.-.-..->_>. lodox ^ Kryptoa ^Silver _.
4 Control Zeolite
i Y
' Iodine
3 I Storage
M-l"
o|
j
^ 1 ^ PRODUCT
DISSOLVE *" LOADOUT
©
^ l' H LOW Level Liquid
10(1-129") (0.5% of Iodine)
100(1-131) r
1 C\fT 1 'X'n — . _,
100(1-131) HIGH mmL
2 WASTE ^ fOPTIOlO
10'000 STOBAGE 1
Efficiency INTERMEDIATE
LEVEL WASH
STORAGE
/ \
+- HEPA _, . _ ^J \
filters / \
1 Stackl
^
i
i
j
i
j
@ i
Macroreticular |
Resin •
1
®'f 1
Final
Vaporizer
Figure 3-2. SIMPLIFIED SCHEMATIC OF ADVANCED IODINE CONTROL SYSTEMS AT REPROCESSING PLANTS
-------
REFERENCES"
(1), ' U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis
of the Uranium Fuel Cycle, Part III - Nuclear Fuel Reprocessing,
EPA-520/9-73-003-D. Office of Radiation Programs, Environmental-
Protection Agency, Washington, D.C. 20460 (October 1973).
(2) MA.GNQ, P.J., et al.. Liquid Waste Effluents from a Nuclear
Fuel Reprocessing Plant, BRH/NIRHL 70-2 (November 1970).
(3) • OAK RIDGE NATIONAL LABORATORY. Siting of Fuel Reprocessing Plants
and Waste Management Facilities, ORNL-4451 (July 1970).
(4) FINNEY, B.C., et al.. Correlation of Radioactive Waste Treatment
Costs and the Environmental Impact of Waste Effluents in the Nu-
clear Fuel Cycle for Use in Establishing "As Low as Practicable"
Guides - Nuclear Fuel Reprocessing, QRNL-TM-4901 (May 1975).
(5) YARBRO, O.O.. Supplementary Testimony Regarding the State of
Technology for and Practicality of Control and Retention of
Iodine in a Nuclear Fuel Reprocessing Plant, Barnwell Hearings,
AEC- Docket No. 50-332 (October 1974).
(6) DAVIS, W., Jr.. Models for Calculating the Effects of Isotopic
Exchange, Radioactive Decay, and of Recycle in Removing Iodine
from Gas and Liquid Streams, ORNL-5060 (September 1975).
(7) YARBRO, O.O., J.C. MA.1LEN, AND W.S. GROENIER, Iodine Scrubbing
From Off-Gas With Concentrated Nitric Acid, 13th AEC Air
Cleaning Conference (1974).
(8) GROENIER, W.S.. An Engineering Evaluation of the lodox Process:
Removal of Iodine from Air Using a Nitric Acid Scrubbing in a
Packed Column, ORNL-TM-4125 (August 1973).
(9) NEWJMAN, R.I.. Fourth Supplement to Direct Testimony of Robert
I. Newman, Barnwell Hearings, AEC Docket No. 50-332.
(10) U.S. PUBLIC HEALTH SERVICE. Air .Pollution Engineering Manual,
999-AP-40 (1967).
(11) OAK RIDGE NATIONAL LABORATORY. Aqueous Processing of LMFBR
Fuels - Technical Assessment and Experimental Program Definition,
ORNL-4436 (June 1970).
(12) OAK RIDGE NATIONAL LABORATORY. Aqueous Fuel Reprocessing
Quarterly Report for Period Ending June 30, 1973, ORNL-TM-4301
(August 1973).
107
-------
(13) OAK RIDGE NATIONAL LABORATORY, Aqueous Fuel Reprocessing
Quarterly Report for Period Ending March 31, 1973, ORNL- •
TM-4240 (June 1973).
(14) PENCE, D.T., et al.. Application of Metal Zeolites to
Nuclear Fuel Reprocessing Plant Off-Gas Treatment, ANS
Trans. 15_, 1, Las Vegas (1972).
(15) ACKLEY, R.D, AND R.J. DAVIS. Effect of Extended Exposure
to Simulated LMFBR Fuel Reprocessing Off-Gas on Radioactive
Trapping Performance of Sorbates, ORNL-TM-4529.
(16) ALLIED-GULF NUCLEAR SERVICES. Barnwell Nuclear Fuel Plant -
Environmental Report, Docket No. 50-332 (November 1971).
(17) UNGER, W.E., et al.. LMFBR Fuel Cycle Studies Progress
Report for August, November and December 1970, ORNL-TM-3281,
ORNL-TM-3127, and ORML-TM-3250.
(18) ALLIED-GENERAL NUCLEAR SERVICES. BarngellNuclear Fuel Plant
Final Safety Analysis Report (October 1973).
(19) OAK RIDGE NATIONAL LABORATORY. Aqueous Fuel Reprocessing
Quarterly Report for Period Ending March 31, 1974, ORNL-
TM-4587 (June 1974).
(20) ATOMIC INDUSTRIAL FORUM, INC. Technical Assessment of Specific
Aspects of EPA Proposed Environmental Radiation Standard for
the Uranium Fuel Cycle (40 CFR 190) and Its Associated Docu-
mentation, AIF/NESP-011 (February 1976).
(21) OAK RIDGE NATIONAL LABORATORY. Voloxidation-Removal of Volatile
Fission Products from Spent LMFBR Fuels, ORNL-TM-3723 (January
1973).
(22) NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS.
Krypton-85 in the Atmosphere-Accumulation, Biological Signifi-
cance, and Control Technology, Report No. 44 (July 1975).
(23) U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis
of the Uranium Fuel Cycle, Part II - Nuclear Power Reactors,
EPA-520/9-73-003-C. Office of Radiation Programs, Environmental
Protection Agency, Washington, D.C. 20460 (November 1973).
(24) UNITED NATIONS SCIENTIFIC COMMITTEE ON THE EFFECTS OF ATOMIC
RADIATIONS. Ionizing Radiation: Levels and Effects, Vol. II,
United Nations Publication E.72.IX.18, New York (1972).
108
-------
(25) CEDERBERG,. G.K-. AND D.K. MACQUEEN.. Containment of Iodine-131
Released by the 1AIA Process-, IDO-14566 (October 1961).
(26) U.S. NUCLEAR REGULATORY COMMISSION. Opinion of the Commission:
In the Matter of Rulemaking Hearing, 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, Docket No.
RM-50-2 (May 5, 1975).
109
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III. MJCLEM. FUEL REPROCESSING
B. Control of Krypton Discharges From
Nuclear Fuel Reprocessing Facilities
-------
1,0 Introduction
The Environmental Protection Agency has undertaken an exhaustive
review of the technology and economics of krypton control at nuclear
fuel reprocessing plants. During this review, EPA has contacted
krypton control equipment vendors, visited national laboratories where
krypton control equipment is being developed or applied, and discussed
a variety of aspects of krypton control with individuals knowledgable
in the techniques of fuel reprocessing.
In the following discussion, current estimates of the costs and
control efficiencies of control systems for Kr~85 are reviewed. The
benefits to be gained by reducing the environmental dose commitments
associated with the release of krypton through installation of such
systems are then set forth. Finally, the level of cost-effectiveness
of cryogenic distillation applied to different fuel reprocessing plant
designs is determined.
Ill
-------
2*0 Source Terms for Krypton
The quantities of fission products present in spent uranium fuel
have been previously reported, based on calculations using the
computer code ORIGIN (1). For krypton-85 this value is 10,500 Cl/MIHM
(expressed in curies per metric ton of heavy metal In the fuel). The
following fuel parameters were used in this report;
Burnup - 33,000 MWd/MTHM
Average Specific Power = 30 MW/MTHM
Cooling Time = 160 days.
It Is assumed that a light-water-cooled power reactor operates at
331 thermal efficiency, producing approximately 33 MTHM of spent fuel
with this burnup for each gigawatt-year of electric power (GW(e)-yr),
and that a typical fuel reprocessing plant has a throughput capacity
of 2100 MTHM per year. Such a plant would be capable of processing
the spent fuel from about 64 such reactors each year. '
If no krypton control systems were installed at a 2100 MT plant,
22 million curies of krypton-85 would be discharged annually. It is
assumed that krypton-85 is discharged to the atmosphere, rather than
into liquid pathways, since currently projected plants uee complete
recycle of process liquids and thus no liquid discharges are planned.
112
-------
3.0 Control Technologies for Krypton at Reprocessing Plants
Since krypton is a chemically inert noble gas, it follows the
process off-gas stream in the fuel reprocessing plant and will be
discharged to the atmosphere unless specially designed air-cleaning
systems are used to capture it. Standard air-cleaning systems based
on chemical processes are ineffective in collecting noble gases. Most
of the krypton produced by the fission process in the reactor is
released to the off-gas stream during dissolution of the spent fuel
(2_t3). A small fraction is also released during the shearing
operation, but this fraction is also routed to the main off-gas
stream. Thus, all of the krypton-85 present in the spent fuel is
collected in one stream, along with other.contaminants, such as oxides
of nitrogen, hydrocarbons, and other radioactive materials.
Two basic systems are in advanced stages of development for the
control of krypton-85: the cryogenic distillation system and the
selective absorption system. These are discussed in turn, briefly,
below:
3.1 Cryogenic Distillation
This process is widely used in industry, where it is better known
as the "liquid air" process and is used to condense and separate the
various gaseous components of air. Heat is removed from air in the
gaseous form in a closed system until the boiling points of the
various gaseous components are reached. As the boiling point of each
component is reached, it liquifies and can be separated from the
remaining gaseous components having lower boiling points. Since
krypton has a boiling point of minus 224°F and the two major gases in
113
-------
air, nitrogen and oxygen, have boiling points of minus 322° F and minus
297° F, respectively, liquif action and separation of the krypton poses
no serious technical problem, Several descriptions of the application
of cryogenic distillation for the removal of noble gases from the off-
gas at nuclear power plants are available (4-11) .
The most serious potential difficulty associated with cryogenic
systems Is the possibility of explosions due to a buildup of hydrogen,
acetylene, hydrocarbons, and oxygen (or ozone) in the system (8).
This can be avoided by chemically removing all oxygen before the gas,
stream is introduced into the qryogenic apparatus (4_). Thus, in order
to use this process, two additional systems are required; a) a
catalytic converter system to convert oxygen to water, hydrocarbons to
carbon dioxide, followed by, b) a system for removal of these products
as well as the oxides of nitrogen. In addition to determining that
the explosion potential of the cryogenic systems is effectively
removed by precleaning the gas stream following use of a catalytic
converter, a full assessment of the remote operation and maintenance
capabilities of this system must be completed in the interim. It
should be noted that the Japanese are installing a cryogenic
distillation system on the fokai-Mura fuel reprocessing plant so that
operating data will be available within the next one or two years
The cryogenic system itself is expected to exhibit a
decontamination factor (DF) of at least 1000 (£-£). However, the
overall efficiency for removal of krypton from the plant is expected
114
-------
to lie somewhat lover because of potential leakage through the system
during startup and shutdown operations, maintenance, etc. Therefore,
an effective plant DF of between 10 and 100 has.been projected for
routine operation of such a system (13).
3.2 Selective Absorption
This process was developed at the Oak Ridge Gaseous Diffusion
Plant (ORGDP), initially for reactors, and more recently specifically
for the control of krypton-85 at fuel reprocessing plants (14,15).
The process is based on preferential dissolution of noble gases in a
fluorocarbon sorbent, such as the refrigerant freon-12. The off-gas
stream is passed through the sorbent in an absorber column at a
relatively low temperature and high pressure. Essentially all of the -
krypton and xenon present are dissolved in the sorbent, along with
other components of the gas stream. The other components are then
removed in a fractionating desorption system and, essentially free of
krypton and xenon, recycled to the off-gas stream. The sorbent is
then transferred to a stripper system where a product gas concentrated
in krypton and xenon is evolved and collected. The pure sorbent is
then regenerated and returned to the absorber column.
The selective absorption process has exhibited a decontamination
factor greater than 1000 in tests with nitrogen oxides and carbon
dioxide (8). However, further investigations are expected to be
accomplished to define the relevant auxiliary systems required for
successful application. Although the selective absorption system is free
from chemical explosion and fire hazards, however, the selective absorption
system does operate at positive pressures of from 50 psig to 350 psig
115
-------
(32). This system has also not been demonstrated at an operating
commercial reprocessing plant. However, it has been offered
commercially for use on the gaseous effluents from nuclear power
reactors (16). A recent review concluded that additional process
development is needed to determine long-term impurity effects, process
reliability, and optimum operating parameters (32). Selective
absorption could be reduced to practice by 1983 provided that an.
orderly program of engineering development, construction, and
demonstration is pursued (,8).
In order to satisfy the proposed standards, storage for 40-70
years would be required, depending upon the degree of initial
decontamination achieved, in order to insure adequate decay. The
management of krypton-85 following its collection has been addressed
by Foster and Pence (17) and appears to present no serious problems.
They reviewed the advantages and disadvantages of long-term storage of
krypton-85 in high pressure steel cylinders and concluded that this
appears to be a practical method for the storage of radioactive gases.
Other methods that appear to offer more safety for comparable^cost are
encapsulation by Sodallte and metal film deposition, which are under
evaluation at Idaho and Hanford, Both methods convert the recovered
lr~85 into a low probability release form for increased safety during
transport and storage (22). ;
116
-------
4.0 Cost of Krypton Control at-Fuel-Reprocessing Plants
Over the past few years, many individual estimates of the cost of
removing krypton from the off-gas at fuel reprocessing plants have
been offered (8,1.9,2£,24_,27}. Typically, each cost given includes or
excludes items relative to other cost estimates so that comparison is
rather difficult. Costs have been given for retrofit situations and
for different krypton control alternatives. The Agency has therefor.e
undertaken an in-depth review of the technology and economics of
krypton control at nuclear fuel reprocessing plants. During this
review, equipment vendors, national laboratories, and experts in fuel
reprocessing technology have been consulted.
In considering the cost of krypton control at reprocessing
plants, it is appropriate to determine such costs on a generic basis.
Therefore, certain parameters applicable to future reprocessing plants
that would affect krypton control costs have been assessed and typical
anticipated values determined:
(1) Plant size: 2100 MTHM per year. Past experience has shown
an increase in the capacity of fuel reprocessing plants, from
the 1 MTU/day NFS plant to the 5 MTU/day Barnwell plant. Exxon
has recently submitted an application for a plant with an
expected capacity of 2100 MTHM per year, or about 7 MTU/day (21,29)
117
-------
(2) Total Gas Flow for Kr-85 Processings 50-100 sefm. The total
off-gas flow that must be treated is-determined by shear
enclosure design and the use of air or other gases for sparging
the diesolver tanks. Review of the state of the art and
discussions with personnel regarding optimum and realistic
operational flow rates indicate that future plants can be
designed with total air flow considerably lower than estimated
for Barnwell (550 scfm) but not as low as the 25 scfis anticipated
in the Exxon application (2JD-.22). The 25 scfm estimated flow
rate estimated for the Exxon plant probably would require
additional costs for leak tightness in the shear and dissolver
sections. Allowing for realistic leakages, a flow of 75 scfm to
100 scfm could be achieved such that the costs for leak tightness
at this level would be offset by a reduction in the size of non
Kr-85 control equipment (such as iodine scrubbers, adsorbers,
particulate filters, etc.) (22).
Although both the cryogenic distillation and the selective
absorption systems are in advanced stages of development it has become
clear that the cryogenic approach to krypton control is much closer to
reality than selective absorption. Cryogenic systems are presently
offered for reactor off-gas cleanup and one such system has been
purchased for use at the Tokai-Mura fuel reprocessing plant in Japan.
Selective absorption is still undergoing development at the Oak Ridge
Gaseous Diffusion Plant and will not be ready for testing with
radioactive materials until 1980. Therefore, the most detailed and
reliable cost estimates for krypton control are available for the
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cryogenic distillation approach* In the following sections, cost
estimates are developed for ,a generic fuel reprocessing plant at off-
gas flow rates of 50 scfm and 100 scfm, and also for the Barnwell
.plant, using a partially redundant system, For comparison! & recent
cost estimate for the Barnwell plant, using a. fully redundant system,
has also been included (20). Table 4.0-1 summarizes these estimates
while the following sections describe in detail the basis for them.
It should "be noted that the cost estimates for a generic plant are
considered appropriate to the great majority of future reprocessing
plants; for the first facility which incorporates krypton control,
higher costs are anticipated to be incurred (on the order of 10-15%
higher overall) (22).
4.1 Direct Costs
Direct costs include the cost of the processing equipment itself,
costs associated with the labor and materials necessary to install the
equipment in the plant, and finally, the price of structures and
buildings needed to properly house the equipment. All costs are given
in first quarter 1976 dollars and are based on the most recent
information available (1§.~^»Z2_24.).
Equipment costs may be influenced greatly by the degree and .type
of redundancy presumed. Complete redundancy of all components may be
achieved by providing an exact duplicate of the primary processing
equipment train. Alternatively, duplicates of only certain, equipment
items may be provided on an, installed basis, or kept on the site for
ready installation. Ezcept for the fully redundant "Earnwell"
estimate, the equipment cost estimates in Table 4.0-1 presume
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installed redundancy of key components, including gas cleanup and
compressors, and are based on the most recent Information available
(18^^,22-24). Gas cleanup includes hydrogen-oxygen catalytic
recombination and catalytic removal of the oxides of nitrogen. The
cold box contains the distillation columns for the recovery and
purification of krypton while the US system is sized according to the
distillation column requirements. Costs for product handling are
appropriate to storage in steel cylinders for a few years. Storage in
Sodalite or via metal film deposition would be approximately $715,000
more expensive in direct costs but offer greater safety in storage and
transport (22). Redundant compressors are provided for all systems as
these contain many moving parts under high stress.
Installation includes all of the labor and materials needed at
the site to integrate the krypton control system into the fuel
reprocessing plant. Such Items as installed piping, instrumentation,
electrical equipment, and the various control equipment are considered
as installation costs; altogether these costs are estimated to be
equivalent to 75% of the equipment cost (22). Finally, costs for the
necessary structures and buildings to properly house the equipment are
included as s direct cost.
4.2 Indirect Costs
Indirect costs include engineering design, field erection costs,
owners costs, interest during construction, and a contingency
allowance. For the generic design and partially redundant Barnwell
design cost estimates, these indirect cost factors were estimated to
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be equivalent to certain percentages of the direct cost (22):
Engineering Design. 15%
Field Erection 50%
Owners Costs 5%
Contingency, 25%
Interest During Construction..... 30%
As shown In Table 4.0-1, the estimate for the fully redundant Barnwell
system also Includes $12,500,000 for escalation to account for
Inflationary trends between now and the time when the money is spent
(1979-1980). Since this cost factor is not appropriate for a present
worth determination and is not considered in the other cost estimates,
it has been deleted from the fully redundant cost estimate to maintain
consistency. The other estimates presume that the money is spent in
the first quarter of 1976.
Contingency is Included as an indirect cost for the generic
designs and the partially redundant Barnwell system; for these systems
contingency represents a cost of 25% of the direct costs. For the
fully redundant Barnwell estimate (20), contingency was presumed to be
40% of all direct and Indirect costs, excluding escalation.
Total capital cost is the sum of the direct and indirect costs.
4«3 Operating and Maintenance Costs
Operating and maintenance (O&M) costs entail costs for utilities
and the labor and equipment necessary for maintenance. For krypton
removal equipment utility costs include electricity, liquid nitrogen,
hydrogen, cooling water, and operating labor. A number of cost
estimates have been made for krypton removal equipment O&M costs and
these have been used to determine the O&M costs shown in Table 4.0-1
121.
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(18-20,22).
4.4 Present Worth
For present worth calculations, a 10% discount rate was used
along with an assumed 20 year equipment lifetime. Under these
conditions, the present worth factor is 8.51356. In order to
calculate present worth for the krypton removal systems, the present
worth of the annual operating and maintenance costs was added to the
total capital cost. As shown in Table 4.0-1 the present worth of the
generic fuel reprocessing plant krypton removal systems ranges between
18 and 24 million dollars, while for the Barnwell design, estimated
present worth costs range from 38.3 to 44.6 million dollars.
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Table 4.0-1
CAPITAL AND WORTH
COSTS OF CONTROL
ESTIMATED • COSTS (.$ 1. OOP)
(a)
Cost Item
DIRECT COSTS
Equipment
Gas Cleanup
Cold Box
LNo System
Product Lqadout
Transfer Cask
Compressors
Installation
Structures, Buildings
Sub -Total: DIRECT COSTS
INDIRECT COSTS
Escalation
Contingency
TOTAL CAPITAL COST
Annual O&M Cost
WORTH: ANNUAL COST
TOTAL PRESENT WORTH (d)
GENERIC DESIGN
ESTIMATES (b)
,50 scfm
1,200
2,000
50
265
20
100
2,900
400
6,940
6,940
1,740
15,620
300
2,550
18,200
100 scfm
1,500
2,500
75
265
•20
100
3,880
750
9,100
. 9,100
2,300
20,500
425-
3.620
24,100
"BARNWELL1' DESIGN
,550 scfm (o)
Partially
Redundant
2,600
3,830
, 93
265
; 20
100
4,080
900
11,900
11,900
3,000
26,800
1,350
11,500
38,300
Fully
Redundant
2,600
7,660
;. 93
265
:..•• 20
100
5,100
1,500
17,300
4,900
^ i y \J )
8,800
31,000
1,600
13,600
•44,600
(a) First quarter 1976 dollars
(b) 2100 MCHM per year
(c) 1500 MTiM per year; fully redundant cost, estimate from reference 20.
(d) Present Worth = Capital Cost -I- (Annual .Cost x 8.51356); 10% Discount
Rate, 20 yr. Control System Lifetime.
(e) Escalation to 1983 not applicable to this present worth determination
'123
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5*0 Doses, and Potential Health. Impact Attributable to Krypton
Discharges^ from Fuel Reprocessing
It la estimated that 157 potential health effects would result
from the uncontrolled release of krypton-85 for 20 years from a 2100
MTHM/yr fuel reprocessing plant. This includes 84 whole body health
effectsi 56 gonadal health effects, and the remainder from exposure of
the lungs to krypton-85 in the atmosphere. For a 1500 MTHM/yr plant
such as Barnwell, the Kr-85 source term and health effects would be
proportionately smaller. The distribution of potential health effects
is shown below for the two types of plants:
•\
2100 MTHM/yr 1500 MTHM/yr
Health Effects "GenericPlant" Barnwell
Whole Body 84 60
Gonads 56 40
Lungs 17_ 12_
157 112
Plant startup in 1983 and a useful lifetime of control equipment
of 20 years is assumed. A simple model for krypton transport which
assumes immediate and uniform dispersion into the world's atmosphere
was used to estimate worldwide doses. Total doses calculated using
this simple model agree with results from a more detailed
multicompartment treatment described by Machta, Ferber, and Hefter
(25, 26) within a few percent, although the two models do differ
regarding the regional distribution of doses delivered immediately
following release, Other parameters, such as population growth and
distribution, dosimetry, and dose-effect relationships, were handled
as described in the previous analysis (27).
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6 • ° _CojB,t-Ef_fecitvene8B of Krypton Control at Fuel Reprocessing Plants
Previous sections have detailed the krypton-85 source term and
potential health impacts of a 2100 MTHM per year fuel reprocessing
plant; additionally, cost estimates for the control of fcrypton-'SS from
such a plant, as well as the 1500 MTHM per year Barnwell plant, have
been made. Table 6.0-1 pulls together the principal data needed to
perform a cost-effectiveness evaluation for the control of krypton of
nuclear fuel reprocessing plants. As shown, cost-effectiveness may be
analyzed either with respect to dollars spent to avoid health effects
or in terms of dollars spent to avoid population exposure in man-rems.
In evaluating krypton control costs, therefore, the EPA has
considered the cost of applying cryogenic distillation at "generic"
plants (2100 MIHM/yr) with off-gas flow rates of,50-100 sefm and at
the Barnwell plant, which is a retrofit case. It should be noted that
although the Barnwell plant has been designed so that krypton control
can be applied, it was not designed to minimize the cost of such
krypton control and as a result has a very large off-gas flow, on the
order of 550 acfm (20). This large (550 scfm) flow is a maximum flow
rate and operating experience may show that lower flow rates are
achievable with minor changes in the shear and dissolving enclosures.
Costs for krypton control and the associated reduction in population
doses and potential health effects are shown in Table 6.0-1. In
considering averted health effects and man-rem, it was assumed that
the cryogenic system would operate 90% of the time needed at a
decontamination factor of 100 (i.e., 99% removal). The fully
redundant Barnwell system, however, is assumed to operate 95% of the
tine, also with 99% removal efficiency. In order to determine the
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Table 6.0-1
COST-EFFECTIVENESS OF KRYPTON C01TROL AT FUEL REPROCESSING PLANTS
Plant Design
GSM1RIG DESIGNS ^
50 SCFM
100 SCFM
"BARNW1LL" DESIGNS ^
Partially Redundant
Fully Redundant(c)
Total
Present
Worth
($1,000)
18,200
24,100
38,300
44,600
POPULATION DOSE
AVERTED (man-kilorem)
Whole
Body Gonads Lungs
187 249 374
187 249 374
131 178 267
141 188 282
CQ
4-1
O
Q)
M-l
tf_l
Health EJ
Averted
140
140
100
105
$/MAN-REM
Whole
Body Gonads Lungs
52 26 5
69 35 7
157 77 15
169 85 17
$/H.E.
AVERTED
130,000
170,000
380,000
425,000
(a) 2100 MUM per year (the design capacity of the proposed Exxon facility,
which projects an offgas flow rate of 25 scfm.)
(b) 1500 MTHM per year; 550 scfm is the reported maximum offgas flow rate
for Banwell (see text)
(c) From Reference 20.
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fraction of present worth costs spent to avoid population doses,
breakdown of potential health effects given in the previous section
was used. Thus-the fraction 84/157 was applied to the $18,200,000
present worth cost of the 50 scfn generic design system tQ calculate
the amount Of money spent to avoid whole body dose. This result was
then used to determine the amount of money spent per man-rem to the
whole body avoided. It can be seen that the costs per man retn for all
of the systems and organs considered are rather small especially when
compared to the MC's interim value of $1,000/whole body or thyroid
man-rem applicable to light water power reactors (30).
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(1) OAK RIDGE NATIONAL LABORATORY. Siting of Fuel Reprocessing
and Waste Management Facilities, ORNL-4451 (July 1970).
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active Effluent from an Operating Nuclear Fuel Reprocessing
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(3) GOODE, J.H.. Hot Cell Evaluation of the Release of Tritium
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(4) DAVIS, J.S., AND J.R. MARTIN. A Cryogenic Approach to Fuel
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(13} BUCKHAH, James A.. Second Supplement to the Direct Testimony
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(24) ATOMIC INDUSTRIAL FORUM, INC. Technical Assessment of Specific
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(29) NUCLEONICS WEEK. (February 5, 1976).
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«tu. aramm PRINTIM OFFK& 1971-626-773/910
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