EPA-520/9-73-003-C
ENVIRONMENTAL ANALYSIS
OF THE URANIUM FUEL CYCLE
i
S-SSSS
1
PART II - Nuclear Power Reactors
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U.S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
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ENVIRONMENTAL ANALYSIS
OF THE URANIUM FUEL CYCLE
\
a
*
PART II-Nuclear Power Reactors
November 1973
U.S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
Technology Assessment Division
401 M Street, S.W.
Washington, D.C. 20460
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CONTENTS
Page
FOREWORD iii
INTRODUCTION 1
PROCESS DESCRIPTION 3
Light-Water Reactor Types 7
Site Characteristics. 10
SOURCES OF RADIOACTIVE DISCHARGES TO THE ENVIRONMENT 15
Fuel Cladding Defects 17
BWR Condenser Air Ej ector Off-gas 22
PWR Gaseous Radwaste System 22
Liquid Radioactive Waste Treatment Systems 24
Primary-to-Secondary Leakage in PWRs 24
System Leakage to Building Atmosphere • 30
Containment Purging 30
Gland Seal Leakage. , 31
Other Sources of Leakage 33
Atmospheric Steam Dumps from PWRs 34
DISCHARGE CONTROL OPTION CONSIDERATIONS 37
LIQUID DISCHARGE CONTROL OPTIONS 43
BWR Liquid Radwaste Systems. 43
PWR Liquid Radwaste Systems. 53
Cost Analysis 62
NOBLE GAS DISCHARGE CONTROL OPTIONS 69
Pressurized-Water Reactors (PWRs) 69
Boiling-Water Reactors (BWRs) 77
RADIOIODINE DISCHARGE CONTROL OPTIONS 83
Pressurized-Water Reactors (PWRs) 83
Boiling-Water Reactors (BWRs) 89
DETERMINATION OF POPULATION RADIATION EXPOSURE 95
Estimation of Radiation Doses from Liquid Effluent Releases 98
Evaluation of External Whole Body Doses from Gaseous Effluents.. 104
Radioiodine Thyroid Dose Computations 110
Evaluation of the Thyroid Doses from Radioiodine Inhalation... 113
Evaluation of Thyroid Ingestion Dose 118
Evaluation of Total Thyroid Dose and Associated Health Risks.. 124
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Page
ESTIMATED RADIATION EXPOSURE FROM NUCLEAR REACTOR EFFLUENTS 129
Liquid Effluents 129
Noble Gases 134
Radioiodine 140
Worldwide Dose Contributions 144
ECONOMIC AND ENVIRONMENTAL COSTS 147
Total Costs 147
Cost-Effectiveness and the Consumer Perspective 149
REFERENCES 161
Figures
1. Appearance of a Typical Light-Water-Cooled Nuclear Power
Station Site 6
2. Pressurized-Water Reactor Schematic 8
3. Boiling-Water Reactor Schematic 9
4. BWR Reactor Water Cleanup System 25
5. PWR Chemical and Volume Control System (CVCS) 26
6. Liquid Case BWR-1: Source Term 45
7. Liquid Case BWR-2: Presently Operating 46
8. Liquid Case BWR-3: Improved Design 47
9. Liquid Case BWR-4: Maximum Treatment 48
10. Liquid Case PWR-1: Source Term 55
11. Liquid Case PWR-2: Presently Operating 56
12. Liquid Case PWR-3: Improved Design 57
13, Liquid Case PWR-4: Maximum Treatment 58
vi
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Tables
Page
1. Principal Characteristics of Representative Reactor Sites... 12
2. Atmospheric Dilution Factors for Vent Release (h = 0) at
Site Boundaries 13
3. Volatile Radionuclide Inventory in a 1,000 MWe Nuclear
Power Plant 19
4. Estimated Reactor Coolant Specific Fission Product and
Corrosion Product Activities 21
5. Estimated Air Ejector Off-gas Release Rates Following
30-Minute Holdup 23
6. Representative Estimated Gaseous Releases Associated with.
Primary-to-Secondary Leakage (20 Gallons per Day) 23
7. Comparison of Estimated Radionuclide Liquid Discharges to
the Environment with Appreciable Primary-to-secondary
Leakage 29
8. Primary-to-secondary System Leakage Experience in Pressur-
ized-Water Reactors 29
9. Estimated Gaseous Releases to Principal Radionuclides from
Miscellaneous Effluents at a BWR Station 32
10. PWR Radioactivity Releases via Atmospheric Steam Dumps 35
11. BWR Plant Parameters Used in Source Term Calculations 40
12. PWR Plant Parameters Used in Source Term Calculations 41
13. Classes of BWR Liquid Radioactive Wastes 44
14. DFs for BWR Liquid Systems 50
15. Liquid Radwaste System Component DFs 51
16. Releases of Long-Lived Radionuclides for BWR Liquid Rad-
Waste Systems 52
17. Classes of PWR Liquid Radioactive Wastes 54
18. DFs for PWR Liquid Systems 60
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Page
19. Releases of Long-Lived Radionuclides for PWR Liquid Rad-
waste Systems 61
20A. Equipment, Annual, and Capital Costs (BWR) 63
20B. Equipment, Annual, and Capital Costs (PWR) 64
21. Liquid Radioactive Waste Summary Table: BWRs and PWRs 65
22. PWR Noble Gas Source Term: 2 Units, 1,000 MWe Each 70
23. Effectiveness of Charcoal Delay Beds on Air Ejector (PWR)... 75
24. Summary Table: PWR Noble Gas Discharge Control Options 76
25. BWR Noble Gas Source Term: 2 Units, 1,000 MWe Each 78
26. Summary Table: BWR Noble Gas Discharge Control Options 82
27. PWR Radioiodine Source Term: 2 Units, 1,000 MWe Each 86
28. Annual Costs for Radioiodine (Elemental) Removal from PWR
Gaseous Effluents 87
29. Annual Costs for Radioiodine (Organic) Removal from PWR
Gaseous Effluents 88
30. BWR Radioiodine Source Term: 2 Units, 1,000 MWe Each 92
31. Annual Costs for Radioiodine (Elemental) Removal from BWR
Gaseous Effluents 93
32. Annual Costs for Radioiodine (Organic) Removal from BWR
Gaseous Effluents 94
33. Principal Exposure Pathways for Radiation Exposure from
Nuclear Reactor Effluents 97
34. Radionuclide Dependent Factors for Liquid Effluent Dose
Calculations 102
35. Parameters Used for the Calculation of the Radiation Dose
from Liquid Effluents 103
36. Population Groups Used for Radioiodine Dose and Risk Evalu-
ations
37. Thyroid Dose Parameters 115
viii
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FOREWORD
The generation of electricity by light-water-cooled nuclear power
reactors using enriched uranium for fuel is experiencing rapid growth in
the United States. This increase in nuclear power reactors will require
similar growth in the other activities that must exist to support these
reactors. These activities, the sum total of which comprises the uranium
fuel cycle, can be conveniently separated into three parts: 1) the
operations of milling, conversion, enrichment, fuel fabrication and
transportation that convert mined uranium ore into reactor fuel, 2) the
light-water-cooled reactor that burns this fuel, and 3) the reprocessing
of spent fuel after it leaves the reactor.
This report is one part of a three-part analysis of the impact of
the various operations within the uranium fuel cycle. The complete
analysis comprises three reports: The Fuel Supply (Part I), Light-Water
Reactors (Part II), and Fuel Reprocessing (Part III). High-level waste
disposal operations have not been included in this analysis since these
have no planned discharges to the environment. Similarly, accidents,
although of potential environmental risk significance, have also not been
included. Other fuel cycles such as plutonium recycle, plutonium, and
thorium have been excluded. Insofar as uranium may be used in high-
temperature gas-cooled reactors, this use has also been excluded.
The principal purposes of the analysis are to project what effects
the total uranium fuel cycle may have on public health and to indicate
where, when, and how standards limiting environmental releases could be
effectively applied to mitigate these effects. The growth of nuclear
energy has been managed so that environmental contamination is minimal
at the present time; however, the projected growth of this industry and
its anticipated releases of radioactivity to the environment warrant a
careful examination of potential health effects. Considerable emphasis
has been placed on the long-term health consequences of radioactivity
releases from the various operations, especially in terms of expected
persistence in the environment and for any regional, national or world-
wide migration that may occur. It is believed that these perspectives
are important in judging the potential impact of radiation-related
activities and should be used in public policy decisions for their
control.
Comments on this analysis would be appreciated. These should be
sent to the Director, Criteria and Standards Division of the Office
of Radiation Programs.
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs
iii
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Page
38. Radioiodine Dose Conversion Factors Per Unit Activity
Intake ng
39. Pathway Transfer Coefficient, Pathway: Inhalation 117
40. Product of the Transfer Coefficient with the Dose Equiva-
lent Conversion Factor for the Inhalation Pathway 117
41. Parameters Used for the Calculation of Radioiodine Intakes.. 121
42. Pathway Transfer Coefficient, Pathway: Vegetables 122
43. Pathway Transfer Coefficient, Pathway: Milk 123
44. Product of the Pathway Transfer Coefficient with the Dose
Equivalent Conversion Factor (Milk Pathway and Vegetable
Pathway) 126
45. Radioiodine Population-Weighted Dose Conversion and Dose-
to-Risk Conversion Factors: Iodine-131 127
46. Radioiodine Population-Weighted Dose Conversion and Dose-
to-Risk Conversion Factors: Iodine-133 128
47. Dose Equivalents from Tritium Releases 130
48. Summary Table: Liquid Radioactive Waste Systems, Estimated
Costs, and Dose Equivalents 131
49. Summary Table: PWR Noble Gas Dose Equivalents and Health
Effects 135
50. Summary Table: BWR Noble Gas Dose Equivalents and Health
Effects 137
51. Summary Table: PWR Radioiodine Dose Equivalents and Health
Effects 141
52. Summary Table: BWR Radioiodine Dose Equivalents and Health
Effects 142
53. Worldwide Health Risk Contributions from Reactor Tritium
Releases 145
54. Worldwide Health Risk from Reactor Krypton-85 Releases 146
55. BWR Noble Gas Systems: Cost-Effectiveness 151
ix
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Page
56. PWR Noble Gas Systems: Cost-Effectiveness 152
57. BWR Radloiodine Systems: Cost-Effectiveness (Elemental
Form) 153
58. BWR Radioiodine Systems: Cost-Effectiveness (Organic Form). 154
59. PWR Radioiodine Systems: Cost-Effectiveness (Elemental
Form) 155
60. PWR Radioiodine Systems: Cost-Effectiveness (Organic Form). 156
61. Liquid Systems: Cost-Effectiveness 157
62. Cost of Producing Electric Power 158
63. Typical Charges for Electric Power Consumption 159
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PART II. NUCLEAR POWER REACTORS
INTRODUCTION
Present estimates of electrical power growth indicate a substantial
increase in the growth of nuclear powered generating stations. By the
year 2000 approximately 65 per cent of the U.S. electrical generation
is expected to come from nuclear energy. In order to meet this
projected demand, approximately 1200 nuclear reactors with a capacity
of one-gigawatt each (1 GWe = 1,000,000 kilowatts) will be required.
Projections of future technology indicate that the Liquid-Metal Fast
Breeder Reactor (LMFBR) is expected to account for a substantial
portion of this forecasted capacity. Based upon these projections
only about 500 GWe will be from light-water-cooled reactors. However,
for the purposes of this analysis, all nuclear power stations installed
through the year 2000 are assumed to employ light-water reactors.
The capacity of individual reactors has increased from 50-200 MWe
in the early 1960's to 1100-1200 MWe (1.1-1.2 GWe) for advanced
reactors presently being ordered by utilities. Problems associated
with emergency core cooling methods have led to a reduction in the
permissible operating power density and, consequently, 1000 MWe has
been assumed as a reference power level for this analysis. If these
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core-cooling problems are resolved, it is possible that the trend
toward larger units will continue so that reactors installed in the
latter part of this century might be considerably larger than the
1000 MWe size assumed in this study.
There are two basic types of light-water reactors: the pressurized-
water direct cycle plant (PWR) and the indirect cycle boiling water reactor
(BWR). The method of operation and the differences between these two
types will be discussed in the following section. At present there
are three domestic manufacturers of the pressurized-water type: the
Westinghouse Electric Corporation, the Babcock & Wilcox Company, and
the Combustion Engineering Corporation. There is only one domestic
manufacturer of the boiling-water reactor, the General Electric
Company. At the present time, pressurized-water reactors comprise
approximately two-thirds of the light-water generating capacity
committed through 1982. This 2:1 PWR:BWR ratio has been assumed to
continue through the year 2000.
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PROCESS DESCRIPTION
A light-water-cooled nuclear power station operates on the same
principle as a conventional fossil-fueled (oil or coal) power
station except that the heat generation is by nuclear fission rather
than combustion. The heat liberated in either process is used to
convert water into steam. The steam enters a three-stage turbine
consisting of one high - pressure stage and two low-pressure stages.
The turbine consists of a common central shaft attached to a circular
array of curved blades. The steam impacts on these blades turning
the rotor at high speeds. The turbine shaft is connected to a wire
wound armature in the generator. This armature rotates in an applied
magnetic field producing alternating electric current.
After passing through the turbine the low pressure steam passes
through a condenser where the steam transfers its remaining- heat to
the condenser cooling water and is condensed back into water and is
recycled into the boiler. The heated condenser cooling water may be
released directly to the environment in a single-pass open-cycle
cooling system. However, this heated water may have an adverse im-
pact on aquatic organisms and the use of open-cycle systems is
decreasing in favor of augmented cooling systems. This is particularly
important for nuclear power plants as they have a lower thermal
efficiency (32%) than fossil-fueled plants (40%) and, consequently,
discharge about two-thirds of their heat output to the environment.
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There are several types of auxiliary cooling systems which have
been proposed for nuclear power plants. The open cycle system may be
retained with the addition of evaporation ponds, long discharge canals,
or spray canals which permit the excess heat to be transmitted to the
atmosphere prior to discharge of the condenser cooling water to the
receiving water body. The other principal alternative is to employ
a closed-cycle cooling system which transfers the heat almost completely
to the atmosphere using cooling towers and recycles the cooled water
back to the condenser.
The principle differences between the nuclear and conventional
electric generating stations are in the type and quantity of fuel
consumed and the nature of the residuals which are discharged from
the process. The fossil-fueled plant will produce sulfur oxides,
carbon monoxide, nitrogen oxides, hydrocarbons, and particulate (dust)
emissions. The nuclear power plant produces highly radioactive atoms
from the fission of the uranium atoms (fission products) and also
from the absorption of neutrons by the coolant and structural
materials (activation products). These radioactive materials are
largely contained within the reactor fuel elements. The greatest
danger from a nuclear power plant would be the release of significant
quantities of these materials as a consequence of fuel element melting
in a serious accident. Because of the enormous quantity of radio-
active material generated in a nuclear power plant,and the inherent
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hazard associated with this material, many precautions must be taken
to insure that these materials are not released to the environment.
One of the most visible precautions against the accidental release
of radioactive materials is the reactor containment which provides
for total enclosure of the reactor and most of the principal reactor
systems. This containment structure is usually cylindrical but may
be enclosed within another building. A sketch of a typical two-unit
nuclear power station is shown in Figure 1. This illustration
depicts the principal structures and features which are externally
visible and characteristic of a nuclear power plant.
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REACTOR CONTAINMENT BUILDINGS
,INE BUILDING
WATER STORAGE TANK
DISCHARGE
CANAL
-..;,,,•-..
.. /r\ > '. V./; • i;r:.
03d*
•
ELECTRICAL
SWITCHYARD
INTAKE
S'l'RUCTURJi
COOLING
WATER
INTAKE
CANAL
Figure 1. Appearance of a Typical Light-Water-Cooled Nuclear Power Station Site
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Ll^ht-Water Nuclear Reactor Types
There are two basic types of light-water-cooled nuclear
reactors: the pressurized-water reactor or indirect cycle and the
boiling water reactor which operates on a direct cycle. The
fundamental difference between these two designs is evident from a
comparison of the two reactor systems which are shown in Figures 2
and 3. In the pressurized-water reactor (Figure 2), the coolant is
maintained at a high pressure (^2250 pounds per square inch) which
inhibits boiling. Steam is produced by allowing the heated primary
coolant to transfer heat to a secondary coolant which is at a lower
pressure (^1000 psi) where boiling can occur. Because the steam
production is separated from the heat generation source, this mode
of operation is termed an indirect cycle.
The boiling water reactor operates on a direct cycle where the
process steam is generated directly in the reactor vessel. This is
possible because of a lower reactor coolant pressure (^1020 "psi)
than in the pressurized-water reactor. The steam generated in the
reactor vessel is separated from excess moisture and passes directly
to the first stage (high-pressure) turbine. The principal components
of the boiling-water reactor system are shown in Figure 3.
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CONTROL ROD
DRIVE MECHANISM
SECONDARY
LOOP
-\ ,_ _ *
R\ /gp3^v*ri^;
CONDENSER COOLING WATER
CONDEN8IN6
CYCLE
REACTOR COOLANT WATER
CONDENSATE
CONDENSER COOLING WATER
Figure 2. Pressurlzed-Water Reactor Schematic
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MOISTURE SEPARATOR
ANDREHEATER
HEATERS
CONDENSATE
PUI.iPS
DRAIN
PUMPS
Figure 3. Boiling Water Reactor Schematic
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Site Characteristics
The local radiological impact of any nuclear power plant effluent
is strongly affected by site related characteristics which govern dis-
persion, reconcentration, and other environmental transport mechanisms.
The integrated population dose is also governed by demographic char-
acteristics such as location of population centers and average popu-
lation density. In order to incorporate these considerations into the
assessment of the radiological impact, three representative sites were
considered. The selection of these sites was based upon the nature of
the body of water receiving waste discharges, population density, and
meteorological conditions. The three sites include one site from each
of three major classifications of site locations: seacoast, river and
lake. The demographic characteristics of these sites are presented in
Table 1, which is based on the enclosed population values of 50 exist-
ing sites. Annual average atmospheric dispersion factors for locations
of particular interest are also given in Table 1.
The liquid effluent dilution factors are calculated on a con-
servative basis for both the maximum individual and average individual
in the population. Values for the river site dilution for the maximum
individual and the general population is based upon the ratio of the
average condenser cooling water flow rate to the river flow rate. For
a 1,000 MWe BWR or PWR the assumed condenser flow is approximately
1,800 cubic feet per second. The assumed river flow rate was 50,000
cubic feet per second. This dilution would be less for augmented
cooling systems such as cooling towers, but at large distances from
the discharge canal, this difference does not remain appreciable.
10
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Table 1 also presents estimates of the atmospheric dispersion
factors which relate airborne concentration to the rate of release
from the reactor. These are calculated for two release heights,
ground level and a 100-meter stack. In order to assure that these
dispersion factors for the selected sites were truly representative,
additional sites were examined. These additional sites are listed in
Table 2. The number of plants examined in each site category was
chosen on the basis of the estimated mix of site locations. The values
for the three representative sites appear to be typical for the average
sites except for the river site where the dispersion conditions are more
favorable at the selected site than at the majority of river sites. The
consequences of this difference will be explored in the dose evaluation
section.
11
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Table 1
Principal Characteristics of Representative Reactor Sites
4,300
5,914
231,729
3,517,236
30,883
339,704
883,774
6,528,988
1,439
25,787
103,206
749,884
Site Location Seacoast River Lake
Enclosed Population
< 5 miles
< 10 miles
< 20 miles
< 50 miles
Distance (miles) to:
Site Boundary 0.5 0.5 0.5
Nearest Resident .75 .73 .5
Nearest Farm 4.5 3.5 2.5
Annual Average Atmospheric Dispersion Factors (x/Q» seconds/cubic meter)
10-meter vent release
Site Boundary 1.2 x 10"? 2.9 x 10~J 1.7 x l(rj>
Nearest Resident 6.0 x 10-' 1.7 x 10~2 1.7 x 10'°
Nearest Farm 3.3 x 10'8 1.5 x 10'7 1.4 x 10~7
100-meter stack release
Site Boundary 3.6 x 10~8 3.3 x 10'7 2.7 x 10~8
Nearest Resident 3.2 x 10-S 2.2 x 10~7 2.7 x 10-§
Nearest Farm 1.8 x 10~8 2.3 x TO"8 2.8 x 10'8
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Table 2
Atmospheric Dilution Factors For
Vent Release (h=0) at Site Boundaries
Site Plant
Seacoast Turkey Point
Forked River
Calvert Cliffs
Seacoast Average (14%)
River Indian Point 2
Arkansas Nuclear One
Maine Yankee
Monticello
Salem
Cooper
Hanford 2
Three Mile Island
Fort Calhoun
V. C. Summer
W. H. Zimmer
LaSalle
North Anna
Waterford
River Average (67%)
Lake Fitzpatrick
Zion
Kewaunee
Fermi 2
Lake Average (19%)
Grand Average
Annual Average
X/Q (Ci/mVCCi/sec)
3.0 x 10"6
1.0 x 10"5
2.7 x 10 -6
5.2 x 10 "6
2.6 x 10~6
4.4 x ID'6
1.2 x ID'5
7.8 x 10-°
5.0 x 10~5
8.0 x 10-5
1.2 x 10-;
9.1 x lO'6
2.2 x 10-5
3.6 x 10'6
9.6 x 10 "6
8.0 x 10"6
~2.8 x 10 -6
2.2 x ID'5
1.7 x ID"5
4.0 x 10~7
3.8 x 10-;
2.4 x 10 -6
1.2 x 10-6
2.0 x 10"6
1.2 x ID'5
Distance
& Direction
640 m NE
640 m SE
1190 m SE
1000 m S
1040 m W
610 m NNE
750 m SSE
190 m WSW
122 m NE
1950 m SE
660 m ESE
400 m ESE
1600 m E
~170 m W
190 m W
1500 m SE
320 m NNE
960 m SW
320 m N
-450 m ESE
~1000 m S
750 m
Source:
From U. S. Atomic Energy Commission Final Environmental Statements for
these respective plants.
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SOURCES OF RADIOACTIVE DISCHARGES
TO THE ENVIRONMENT
Nuclear power reactors generate radioactive materials as a
consequence of the fissioning of uranium and by neutron absorption in
the coolant and in structural materials which leads to induced radio-
activity in these components. The products of uranium fission comprise
a large number of elements and include both stable and radioactive
isotopes of these elements. Among the more important radionuclides
produced by uranium fissioning are isotopes of the noble gases krypton
and xenon, the alkali metals cesium and rubidium, the alkaline earths
barium and strontium, and the halogens iodine and bromine.
The capture of the neutrons liberated in fission by the nuclei
of stable elements often results in the production of radioactive
activation products. The coolant activation products are generally
gases such as argon-41, fluorine-18, nitrogen-13, nitrogen-16, and
oxygen-19 which have short half-lives in the range of several seconds
to a few hours. The induced activities in the structural materials
may have considerably longer half-lives and comprise a much wider
range of elements including zirconium, manganese, nickel, iron, carbon,
chromium, cobalt, and copper. These radionuclides usually remain
fixed in the structural materials but can enter the coolant as a
consequence of corrosion and erosion in the pumps and other moving
components.
15
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Nuclear power reactors are constructed with multiple barriers for
isolating these radionuclides from the environment. The principal
barriers are: (1) the fuel cladding, (2) the reactor systems, and (3)
the reactor and auxiliary buildings. Release of radioactive material
to the environment occurs principally as a consequence of the penetra-
tion of one or more of these barriers. This penetration can occur due
to the presence of structural defects, leakage from pumps or other
components, or intentionally as a consequence of the particular plant
design. The remainder of this section discusses the specific release
pathways from each barrier while providing typical estimates for the
magnitude of release from each pathway.
16
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Fuel Cladding Defects
The primary barrier for isolating radioactive fission products
from the environment is the fuel rod cladding. Within a 1000-MWe
nuclear reactor there are millions of curies of radioactive isotopes;
the iodine-131 alone can amount to over 70 megacuries. The major
fraction of these fission products is retained within the ceramic
matrix of the uranium dioxide fuel pellet. However, the more volatile
elements such as the halogens and noble gases can diffuse through
this matrix into the space between the fuel pellet and the cladding.
This diffusion process is accelerated by the high fuel temperatures
and the presence of cracks and fissures produced by thermal stresses
so that appreciable amounts of these elements accumulate in the fuel-
cladding gap. Nonvolatile elements such as strontium, barium, and
cerium also accumulate there as they are daughter products produced
by the radioactive decay of short-lived noble gas precursor radio-
nuclides. These radionuclides will be contained within the fuel rod
as long as the thin metallic cladding remains intact. In this situa-
tion, the quantity of radioactive material reaching the coolant will
be limited to fission products arising from small traces of uranium
which remain on the outer surface of the fuel as a consequence of the
manufacturing process, activation products which arise from neutron
induced reactions with water and air, and traces of metallic elements
which enter the coolant as a result of corrosion of the reactor vessel,
piping,and other structural components.
17
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The fuel cladding is approximately twenty-five thousandths of an
inch in thickness for PVTRs and approximately thirty-three thousandths
of an inch for BWEs and is subjected to thermal stresses as the
reactor power level is changed and mechanical stresses from the
high pressure and velocity of the coolant or from physical contact
with the fuel as it expands. These stresses, combined with varia-
tions in the cladding thickness or other irregularities in manufac-
ture, can result in small pin-holes or defects in the fuel cladding
which allow the volatile radionuclides in the cladding gap to
escape into the coolant. Under severe conditions large failures
could occur in the cladding which would permit the coolant to con-
tact the fuel and leach out the less volatile fission products.
However, these occurrences are not common and cladding failures of
the small pin-hole type are more usual.
The extreme conditions imposed in the reactor core on the fuel
cladding together with the difficulties of producing large quantities
of thin, near-perfect tubing for the large number of fuel rods
(approximately 40,000) make it extremely difficult to eliminate such
fuel cladding failures. As a result, nuclear reactor systems are
designed to accommodate the equivalent of 1 percent of the gap
activity (contained in all of the fuel rods) escaping to the coolant
through cladding defects. Table 3 shows the relationships between
the total core inventory, the fuel plenum (gap) inventory, and the
primary coolant inventory for a representative 1000-MWe light-water
reactor. Even with defective fuel, the primary coolant activity
18
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Table 3
Volatile Radionuclide Inventory in
a 1000 MWe Nuclear Power Plant
Parameters; 3040 MWt PWR, operating at full power for 500
days with 1% of the fuel rods having cladding
defects.
Total Activity In :
Reactor Fuel-Cladding Primary
Core Gap (mega- Coolant
Radionuclide
Iodines
1-131
1-132
1-133
1-134
1-135
Half-
life
8.05d
2.3 h
21. h
52. m
6.7 h
(mega- curies)a
curies)
74.9
114.0
171.0
206.0
158.0
0.76
0.14
0.64
0.12
0.34
(curies)
465
186
766
117
420
Kryptons
Kr-85
Kr-85m
Kr-87
Kr-88
Xenons
10.8 y
4.4 h
76. m
2.8 h
0.66
33.5
64.4
93.0
Xe-133
Xe-133m
Xe-135
Xe-135m
5.3 d
2.3 d
9.2 h
15.6 m
164.0
4.0
43.6
46.4
0.067
0.95
0.076
0.149
4.17
0.019
0.084
0.016
334
439
261
775
52,290
692
1,488
42
ll megacurie = 1,000,000 curies (106 curies)
Source and Assumptions:
Appendix D of the Final Safety Analysis for
the Kewaunee Nuclear Power Plant (converted
from 1721.4 MWt to 3040 MWt and core volume
adjusted to scale).
19
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remains a small fraction of the total inventory within the reactor.
The coolant purification systems are responsible for removing most
of the primary coolant activity so that low levels are maintained in
the circulating coolant. Typical primary coolant radionuclide con-
centrations for a reactor having 1 percent failed fuel are shown in
Table 4.
20
-------
Table 4
Estimated Reactor Coolant Specific Fission Product
and Corrosion Product Activities (at 578° F)
ssion product reactor coolant concentrations
Isotope
Noble Gas Fission
Kr-85
Kr-85m
Kr-87
Kr-88
Xe-133
Xe-133m
Xe-135m
Xe-138
Total Noble Gases
Corrosion
Mn-54
Mn-56
Co-58
Fe-59
Co- 60
uCi/cc
Products
1.11
1.46
0.87
2.58
1.74 x 102
1.97
0.14
0.36
187.3
Products
4.2 x 103
2.2 x 10~ 2
8.1 x 10~3
1.8 x 10"3
1.4 x 10~3
Total Corrosion Products 3.7 x 10" 2
corresponding to 1
Isotope
Fission Products
Br-84
Rb-88
Rb-89
Sr-89
Sr-90
Y-90
Y-91
Sr-92
Y-92
Zr-95
Nb-95
Mo-99
1-131
Te-132
1-132
1-133
Te-134
1-134
Cs-134
1-135
Cs-136
Cs-137
Cs-138
Ce-144
Pr-144
% Failed Fuel
UCi/cc
3.0 x 10~2
2.56
6.7 x 10~2
2.52 x 10'3
4.42 x ID" 5
5.37 x 10"5
4.77 x 10"4
5.63 x 10"4
5.54 x 10"4
5.04 x 10~4
4.70 x 10"4
2.11
1.55
0.17
0.62
2.55
2.2 x 10~2
0.39
7.0 x 10~2
1.4
0.33
0.43
0.48
2.3 x 10~4
2.3 x 10~4
Total
Fission Products 12.8
Source: Kewaunee Final Safety Analysis Report, Appendix D, Table D 4-1
21
-------
BWR Condenser Air Ejector Off-gas
The boiling-water reactor operates on a direct cycle and the
contaminated coolant passes directly through the turbine. Entrained
radioactive gases, air which has leaked into the condenser, and
hydrogen and oxygen which result from the radiolytic dissociation of
water are removed from the main turbine condenser by the steam jet air
ejector which is used to maintain a vacuum in the condenser. These
gases are removed at a rate of about 300 cubic feet per minute.
Approximately 230 cfm represents the dissociated hydrogen and oxygen,
5-20 cfm represents air in-leakage and the remainder in water vapor;
the radioactive gases contributing negligible volume.
In the absence of appreciable failed fuel, the principal
contributor to the radioactive emission is nitrogen-13. When there
is significant failed fuel, the noble gas fission product releases
dwarf the activation gas releases as shown in Table 5.
PWR Gaseous Radwaste System
In the operation of a PWR, boron is added to the primary coolant
to act as a neutron absorber. In the beginning of the fuel cycle its
concentration is approximately 1,000 ppm. As the reactor produces
power, less and less boron is required. In order to remove this boron
a small portion of the coolant purification flow is typically "bled" to
the boron recovery system (Figure 5). Radioactive gases evolved at the
gas stripper are routed to the waste gas system for treatment. Table 6
provides an estimate of the radioactivity releases from a waste gas
system providing 45 days of holdup for these gases.
22
-------
Table 5
Estimated Air Ejector Off-gas Release Rates
Following 30-Minute Holdup
1064-MHe BUR with 0.25X
Failed Fuel, 18.5 acfm in-leakage
Radionuclide
Nltrogen-13
Krypton-83m
Krypton-85m
Krypton-85
Krypton-87
Krypton-88
Krypton-89
Xenon-131m
Xenon-133m
Xenon-133
Xenon-13Sm
Xenon-135
Xenon-137
Xe non-138
Total
Half- Emission rate
life pCi/sec
10 Bin
1.9 hrs
4.4 hrs
10.8 yrs
76 Bin
2.8 hra
3.2 min
11.8 day •
2.3 days
5.3 daya
9.1 hrs
15.6 min
3.9 min
17.5 Din
340
2,537
5,700
7.5
15,700
17,367
262
15
188
5,100
8,000
17,367
860
26,500
100.000
Annual
Discharge
Ci/yr
8,580
64,000
143,800
189
396.000
438,000
6,610
378
4,743
128,700
202,000
438,150
21,700
668,600
2.523.000
Source: Browns Ferry.Final Safety Analysis Report.
Table 6
Representative Estimated Gaseous Releases
Associated vith Primary-to-Secondary
Leakage (20 Gallons per Day)
Annual Activity Release to the Environment
(curies per year) from
Principal
Radionuclide
Krypton-85
Krypton-87
Krypton-88
Xenon-131n
Xenon-133
Xenon-135
Xtnon-138
Iodine-131
Containment
Purge
13.0
0.04
-
10.0
1005.0
0.018
0.007
0.018
Waste Gas
Processing
System
791
-
-
63
1500
-
-
.
Steam
Generator
Leakage
2.0
3.0
10.0
3.0
682.0
3.0
2.0
0.62
Source: (Table III-3) of the AZC Draft Environmental Impact
Statement for Indian Ft 1
23
-------
Liquid Radioactive Waste Treatment Systems
In BWRs and PWRs, various sources of liquid waste are handled by
liquid waste treatment systems. Each reactor type provides for the
purification of the reactor coolant. In BWRs this system is simply
referred to as the reactor water cleanup system (RWCS) and is shown
in Figure 4. In PWRs, coolant purification (and chemical adjustment)
is provided by the chemical and volume control system (CVCS) which it-
self may be classified into two subsystems, the reactor coolant cleanup
subsystem and the boron recovery subsystem (Figure 5). The boron
recovery subsystem of the CVCS in a PWR, and RWCS in a BWR, may con-
tribute radioactive liquids to the respective liquid radioactive waste
treatment systems in each type of reactor. These liquid radioactive
waste treatment systems handle the miscellaneous radioactive liquids
generated by plant operation as well as those liquids from the coolant
purification systems. Figures 7 and 11 (pages 43 and 53) illustrate
liquid radwaste systems representative of BWRs and PWRs presently oper-
ating. Table 7 (page 28) compares the magnitude of PWR liquid radio-
activity releases from the CVCS, the liquid radwaste system, and steam
generator blowdown during a postulated 20 gallon per day primary-to-
secondary leak.
Primary-to-Secondary Leakage in PWRs
In pressurized water reactors, the secondary coolant system is
isolated from the primary reactor coolant by vtrture of the tubing in
the steam generators. If these tubes remain intact, the secondary
system would be free of radioactive material. The number of these
tubes can be about 4000 per steam generator, depending upon the
reactor make and plant design power level, and the number of
24
-------
Feedwater
Return
From
Reactor
Recirculatlon
System
Main
Condenser
Radvaste.
System
(Startup)f
Cleanup
Filter-
Demi, nerall
',ers
Figure 4. BWR Reactor Water Cleanup System
10
Ul
-------
Reactor Coolant Cleanup Subsystem
Reactor
Coolant
Letdown
Charging
Line
from
—Boron Recovery
Subsystem
Boron Recovery
Subsystem
Boron Recovery Subsystem
Reactor
Coolant
Letdown
From
C Icanup
Subsystem
S &
i I
— 5 —
Ion-X
Filter
Gas
Strippi
i
, .} ...
CVCS Holdup
Tanks (3)
Cleanup
Subsystem
(Boric Actd
Tanks)
Liquid Waste Disposal Evaporator
Figure 5. PWR Chemical and Volume Control System (CVCS)
26
-------
steam generators per plant may range between 2 and 4, depending on the
power rating and reactor vendor. Due to the large number of tubes
present in a plant, the possibility of defects in the tubing either due
to manufacturing errors, or operating conditions (corrosion, burn-out,
or stress) is enhanced. If holes develop in the steam generator tubing,
the primary reactor coolant water will leak into the steam generator as
a consequence of the higher primary system pressure. This water will
contain radioactive materials at the concentration present in the pri-
mary coolant and consequently contaminate the secondary coolant system.
Volatile radionuclides which enter the secondary coolant system
as a result of primary-to-secondary leakage may be discharged to the
environment via two pathways: (1) the condenser air ejector and
(2) the steam generator blowdown flash tank. The condenser air ejector
removes entrained gases from the secondary coolant and will extract
radioactive noble gases, gaseous activation products, and some of the
halogens (iodines). In many designs, the steam is generated from
evaporation of water in the steam generator. Solids build up in the
steam generator and may impair heat transfer. To counteract this
buildup, typically 5-15 gallons of the steam generator bottom liquid
are withdrawn per minute. This hot liquid is bled into a flash tank
at a lower pressure than the secondary coolant system. The low
pressure and high temperature cause the rapid evaporation (flashing)
of 30-40 percent of the liquid into steam which, in older plant
designs, was released to the atmosphere. The volatile radionuclides
would also be carried over in the steam and, consequently, released.
27
-------
The relative magnitude of these gaseous releases from the steam
generator leakage are shown in Table 6. The major contribution from
these releases is the additional iodiiie-131 which can be considerably
greater than that from other sources.
In older plants, the blowdown liquid remaining in the flash
tank would be discharged to the condenser coolant without any cleanup.
Any radionuclides which were in the steam generator blowdown as a
result of primary-to-secondary leakage would be released to the
environment. If appreciable (20 gallons per day) pritnary-to-secondary
leakage is present, concurrent with discernible fuel cladding perforations,
the unprocessed steam generator blowdown could be the major source of
liquid radionuclide discharges as illustrated in Table 7.
Operating experience has shown that pressurized water reactors
eventually develop some primary-to-secondary leakage. Generally, with
only a few tubes having defects, this leakage may amount to only a few
gallons per day. However, several plants have experienced long periods
of operation with leakage rates of 50 gallons per day or more (Table 8).
Because of the additional solids contributed from the boric acid in the
primary coolant, high leakage rates cannot be tolerated for more than a
few days. These high leakage rates require corrective action which
usually means plugging up the defective tubes by sealing them off with
plugs and small explosive charges or by welding.
28
-------
Table T
Comparison of Estimated Kadionuclide Liquid Discharges Co
the Environment with Appreciable Priraary-to-Secondary
Leakage (20 Gallons per Day Leakage and 10 Gallons per
Minute Slowdown)
Activity (curies per year) Discharged from
u . v-nemiuai ana
naior „ —
Radionucltdes Volume Contro1
System
Molybdenum-99
Technetiuitt-99m
Iodine-130
Iodine-131
lodlne-1^2
Iodine-133
Cesium-134
Iodine-135
Cesiud-136
Ceslun-137
0.005
0.004
0.002
0.59
0.056
0.56
0.004
0.14
0.001
0.003
waste
Treatment
System
0.018
0.016
0.006
Z.06
0.19
1.92
--
0.45
0.005
0.012
a team
Generator
Blovdovn
5.51
0.61
0.009
8.1
0.12
3.46
7.1
0.62
2.05
6.06
Totals
M.4
Source: Table Ill-It of the AEC Draft Environmental Statement for the
Indian Point Nuclear Station Unit 2.
Table 8
Primary-to-Secondary System Leakage Experience
in Pressurized Water Reactors
Plant
H. B. Robinson
Unit 2
Point Beach
Unit l
Connecticut Yankee
(Huddam Neck)
San Onofrt
Unit 1
Yankee Rowe
Average Leakage
Rate (gallons)
per day)
55
14,400
up to 50
1,500
up to 15
up to 95
up to 1,200
Duration
7 months
<1 day
several months
several days
several months
several weeks
several months
Source: Operating reports for these respective plants.
29
-------
System Leakage to Building Atmosphere
Release of radioactivity from reactor and waste treatment systems can
occur via leakage directly from system components. Much of this leak-
age is from coolant pump or valve seals and is generally returned
directly to the reactor coolant system. Other leakage paths include
smaller valve seals and releases associated with chemical and radio-
logical analysis. Most of the liquid released will be collected by
plant drain systems and be processed by the waste treatment system.
The volatile elements, including the noble gases and halogens, will be
released to plant buildings such as the containment building where they
are available for leakage or discharge to the environment. In pres-
surized water reactors, this leakage may average between 0.2 - 0.3
gallons per minute and account for 0.4 - 1.0% of the coolant volume per
day. Similar leakage rates may be expected at boiling water reactor
plants. Thus appreciable quantities of the volatile elements may
accumulate in the reactor containment building and auxiliary building
atmospheres.
Containment Purging
Radioactive halogens and noble gases which escape to the reactor
containment from the reactor system may be discharged to the environ-
ment during containment venting or purging. The containment atmosphere
is vented or purged in order to test the containment isolation system
on a periodic basis, reduce containment temperature and activity levels
Environmental Report, H. B. Robinson Unit 2, Supplement No. 1. Answer
3.5f.
Operating Reports 3 and 4, Point Beach Nuclear Power Station.
Kahn, et al., "Radiological Surveillance Studies at a Pressurized Water
Nuclear Power Reactor" op. cit.
30
-------
prior to and during maintenance involving entry into the containment,
and also to reduce containment pressure if excessive system leakage
exists. The purges for testing are typically of one to several minutes
in duration and may occur on a monthly schedule. The other purging
intervals may last several hours or more and may occur 1-10 times per
year. Estimated releases of gaseous radionuclides via containment
purging are shown in Tables 6 and 9 for the two types of light-water
reactors.
Gland Seal Leakage
Equipment with external moving parts such as valves and the
coolant pumps contain a soft packing to retard the loss of fluid and
steam from the reactor system. This packing does not provide total
isolation and is a major source of the coolant system leakage described
previously.
2
In a boiling water reactor , a similar condition exists with regard
to the turbine generator shaft. As the steam passing through the
turbine was generated in the reactor vessel, it contains volatile radio-
active fission and activation products such as the noble gases and iodines.
In order to reduce the loss of these volatile nuclides from the turbine,
process steam is bled into the outer portions of the turbine seals and
removed via a gland steam condenser. The non-volatile radionuclides are
condensed and the volatile radionuclides pass through a 2-minute delay
line to the gaseous release discharge point. Small quantities of
2
A somewhat similar condition would exist in a pressurized-water reactor
which is operating with appreciable primary-to-secondary leakage and
fuel cladding defects.
31
-------
Table 9
Estimated Gaseous Releases of Principal Radionuclides from
Miscellaneous Effluents at a BWR Station
(Ci/yr per unit)
Nuclide
Krypton -83m
Krypton —85m
Krypton -85
Krypton -87
Krypton -88
Krypton -89
Xenon -131m
Xenon -133m
Xenon -133
Xenon -135m
Xenon -135
Xenon -137
Xenon -138
Iodine -131
Iodine -133
Turbine
Gland
Seal
41
69
-
200
220
490
-
4
120
320
350
900
1,020
0.041
0.214
Turbine
Building
10
16
-
49
53
17
-
1
29
82
84
290
260
0.547
2.54
Reactor Mechanical
Building Vacuum
& Containment Pump
-
-
'
-
_
-
- -
1445
-
215
• - -
. •
0.012
0.041
Source: Table 3.7 of AEC Final Environmental Statement for the Duane Arnold
Energy Center
32
-------
the steam from the seal together with the entrained volatile radio-
nuclides also escape to the turbine building atmosphere and are released
to the environment unprocessed via the turbine building exhaust.
Estimates of the total discharge of volatile radionuclides from the
gland seal system are shown in Table 9.
Other Sources of Leakage
Leakage from the coolant purification and waste treatment systems
will be collected by the auxiliary building floor drains. Volatile
radionuclides which remain entrained in the liquids will be released
to the auxiliary building ventilation system and roof vents on the
reactor building to the environment. In addition, gases will M
released from radiochemical fume hoods during sample analysis and from
tank venting and purging operations. These releases are highly
variable and depend on system design parameters, construction tech-
niques, maintenance, sampling and venting frequencies, etc. Estimates
of the releases from these sources are shown in Table 6 for the PWR
and Table 9 for the BWR.
During reactor startup, it is necessary to initially depressurize the
cooling-water condenser. As a vacuum is drawn, the coolant present in
the condenser will be partially degassed and the noble gases and a fraction
of the halogens will be released. The number of startups and the
Intervals between shutdown and startup (which represents a decay period)
are highly variable. Estimates of this frequency are about 2-10 cold
startups per year. The estimated gaseous releases for 2-3 startups per
year are shown in table 9 under the column "mechanical vacuum pump."
33
-------
Atmospheric Steam Dumps from PWRs
In order to relieve high pressures in the secondary system from
various abnormal operations (e.g., load rejection), PWR designs in-
clude a provision for relieving steam directly to the atmosphere
through atmospheric steam dump valves. Of particular interest is the
steam release which would accompany runback operations (i.e., rapid
reduction of reactor power from 100% to a level at least high enough
to supply the unit auxiliary load) via the main steam relief valves.
The magnitude of this release can be on the order of many tens of
thousands of pounds of steam in one minute. With primary-to-secondary
leakage and failed fuel, radioactivity as well as steam would be
released directly to the atmosphere. Although noble gases and some
particulates would be released, the main concern would focus upon the
release of radioiodine. Table 10 shows the estimated releases result-
ing from the actuation of the main steam valves for a one-minute
period.
Another direct atmospheric pathway for secondary system steam
exists via the feedwater heater relief valve discharge. Radio-
activity release is again predicted on the concurrent presence of
failed fuel and primary-to-secondary leakage. Table 10 also shows
the estimated releases resulting from the actuation of the feedwater
heater relief valves for a one minute period. On the order of ten
thousand pounds of steam may be released during this procedure.
34
-------
Table 10
PWR Radioactivity Releases via Atmospheric
Steam Dumps (Ci/yr for 20 gpd Primary to
Secondary Leakage and 0.2% Failed Fuel in a
3358 MWt PWR)
Radioisotopfe
Noble Gases
Kr-85
Kr-88
Xe-133m
Xe-133
Xe-135m
Xe-135
Radioiodine
1-131
1-132
1-133
1-134
1-135
Particulates
Mo-99
Tc-99m
Te-132
Cs-134
Cs-137
All Others
Total
Curies discharged per year
Main Steam
Relief Valves
(1 minute release)
9.06 (-5f
1.30 (-4)
1.09
9.78
(-4)
(-3)
3.18 (-4)
2.44 (-4)
6.60 (-4)
8.66 (-5)
6.44 (-4)
9.22 (-6)
1.83 (-4)
1.18
7.32
6.02
(-4)
(-5)
(-6)
6.78 (-6)
3.54 (-5)
3.0 (-4)
1.27 (-2)
Feedwater Heater
Relief Valves
(1 minute release)
1.28 (-4)
1.68 (-5)
1.25 (-4)
1.78 (-6)
3.54 (-5)
2.28 (-5)
1.41 (-5)
1.16 (-6)
1.31 (-7)
6.64 (-6)
4.0 (-7)
3.56 (-4)
ai.e., 9.06 (-5) is equivalent to 9.06 x 10~5
Source: Trojan Nuclear Plant Final Safety Analysis Report,
35
-------
DISCHARGE CONTROL OPTION CONSIDERATIONS
A variety of discharge control options were explored for both
the boiling-water and pressurized-water reactor plants. The effec-
tiveness of these options and their associated costs vary significantly.
The reasons for this variation are:
(1) the differences between the two reactor types and the
applicability of individual control techniques to each type of plant;
(2) the presence of multiple release pathways for the same
radionuclide and lack of detailed information on the magnitude of
secondary pathway releases;
(3) uncertainties in the effluent composition and chemical
form;
(4) variability in the estimates of effectiveness and in
the available cost data for a particular control option; and
(5) the selection and ranking of components and the order
in which they are added to the baseline system.
Item 3 is particularly significant for the radioiodine releases
as there is considerable uncertainty as to the chemical form of the
effluent. As discussed below, differences in the chemical form of
the radioiodine emissions can determine the effectiveness of control
options and also greatly affect the choice of the critical exposure
pathway.
Item 4 is also a major source of uncertainty in the analysis
presented here as there is a scarcity of available information on
system effectiveness,and inconsistencies exist in the available cost
37
-------
data. In particular, it is difficult to determine the cost components
(operating and maintenance costs, process equipment capital costs, piping
and installation costs, building and structural costs) associated with
specific system cost estimates provided in the literature (1,12-15,19-24).
This is especially true as much of the available utility data pertains
to installing additional systems in existing plants and the additional
cost for this retrofitting is not always immediately discernible. This
uncertainty would lead to overestimates in the cost of installing simi-
lar systems in new plants at the design stage.
Many operating plants and those in the construction or design stages
have specified treatment systems and the associated cost of these systems
for attaining "lowest practicable effluent discharges" as required by the
proposed Atomic Energy Commission Rulemaking, 10 CFR 50 Appendix I. This
information has been extensively used (in preparing this analysis), to-
gether with information provided by the AEC (1)• However, due to the
number of architect-engineering firms and reactor vendors and their indi-
vidual engineering and design preferences, there is a wide variety, not
only in the type of control system for a particular effluent pathway, but
also in the way various control options are combined in individual plant
systems. This multiplicity further complicates the selection of the most
cost-effective systems. In the present analysis, the attempt was made to
add on components in a logical sequence. It Is quite conceivable that,
based on different system costs or assumptions, or actual operational
experience with these systems, the order of addition of systems could
differ greatly from the present analysis.
38
-------
This analysis is also predicated on the occurence of certain
failures and departures from optimal designed operation which con-
ceivably may not take place during the lifetime of a particular plant.
In part, these assumptions have been based on operating experience
with the existing light-water reactor plants. It should be emphasized,
however, that current or past experience is not sufficiently documented
nor widely distributed in all cases to permit an "average" value to be
adopted with confidence.
A number of parameters must be specified in order to estimate the
magnitude and composition of reactor effluents. These parameters include
process flow rates, leakage rates, partition factors involving various
phase changes, and internal reactor system cleanup decontamination factors
3
(DF) . Tables 11 and 12 present the individual plant characteristics
assumed to provide the basis for source term calculations and radioactive
waste treatment system sizing.
Because of differences.in design and a lack of information with.
respect to long term operation of the larger commercial light water
reactors, no plant presently operating fits exactly the operating pa-
rameters in Tables 11 and 12. However, where possible operating
experience has been factored into these estimates along with generally
accepted values for various parameters (1,2,6,9,24,26-28,34,51,52).
3
The decontamination factor of a process is defined as:
Up „, concentration in entering stream
concentration in effluent stream '
39
-------
Table 11
BWR Plant Parameters Used in Source Term Calculations
(One Unit)
Reactor power
Capacity factor
Fraction of fission products passing through:
Condensate Denineralizer
H-3, Y, Mo
Cs, Rb
Others
Clean-up Demineralizer
H-3, Y, Mo
Cs, Rb
Others
Partition factor (iodine in vapor/water)
Reactor (steam/water)
Reactor Building (cold water)
Turbine Building
Radwaste Building (hot water)
Radwaste Building (cold water)
Gland Seal
Air Ejector
Partition factor (other fission products in
vapor/water, except H-3)
Fraction of iodine passing through:
Condensate Demineralizer
Clean-up Demineralizer
Raactor Building Filter (HEPA)
Turbine Building Filter (HEPA)
Radwaste Building Filter (HEPA)
Gland Seal Condenser
Leaks:
Reactor Building (cold water)
Turbine Building (steam)
Radwaste Building (hot water)
Radwaste Building (cold water)
Gland seal steam flow:
JLOOO MWe
80%
1.0
0.1
0.001
1.0
0.5
0.1
Elemental
U.Ulw!
0.001
1.0
0.1
0.001
0.1
0.005
Methvl
25° C 100° C
0.012 0.012
1.0
].0
1.0
1.0
0.1
0.5
0.001
0.001
0.1
1.0
1.0
1.0
0.01
1.0 gpm
2400 Ib/hr
1 gpd
19 gpd
0.1% steara flow
rate
-------
Table 12
PWR Plant Parameters Used in Source Term
Calculations (One Unit)
Reactor power
Capacity factor
Number of steam generators
Number of cold shutdowns per year
Reactor containment volume
Number of containment purges per year
Slowdown rate
Fraction of power from failed fuel
Escape rate coefficients (sec"1)
Xe and Kr
I, Br, Rb, Cs
Mo
Te
Sr, Ba
Others
Fraction of fission products passing through
primary coolant demineralizer (except H-3,
Y, Mo, Cs, Rb)
Cs, Rb
H3, Y, Mo
Partition factor (iodine in vapor/water)
Steam generator
Slowdown vent
Condenser air ejector
Containment (hot water)
Auxiliary building (hot water)
Auxiliary building (cold water)
Turbine building (steam)
Gland seal
Fraction of iodine passing through
Containment filter (HEPA)
Turbine building filter (HEPA)
Auxiliary building filter (HEPA)
Gland seal condenser
Leak rate of primary coolant:
Reactor building (hot water)
Auxiliary building (hot water)
Auxiliary building (cold water)
Steam generator
Leak rate of turbine steam
Gland seal steam flow
1000 MWe
80%
3
2 ,
2.0x10° cu ft
4
15 gpm
0.25%
6.5xlO~8
,3x10
,0x10
, 0x10
,0x10
-8
,-9
-9
,-H
1.6xlO"12
U.UJ.
0.05
0.0005
0.1
0.1
0.0001
1.0
0.1
1.0
0.1
40 gpd
1 gpd
19 gpd
20 gpd
2400 Ib/hr
0.1% steam
flow
41
-------
LIQUID DISCHARGE CONTROL OPTIONS
BWR Liquid Radwaste Systems
The liquid radioactive waste treatment system at a BWR power station
is responsible for decontaminating a wide variety of waste liquids.
These liquids may be divided into four general classes as shown in Table
13.
In general, four BWR liquid radioactive waste treatment systems were
constructed to illustrate both the spectrum of treatment options
available as well as the development that has taken place in such
systems to reduce radioactivity releases. These systems are sized for a
two-unit BWR station (1000 MWe per unit) and are shown in Figures 6
through 9. It should be noted that deep-bed condensate demineralizers
are assumed for cleanup.
Minimum treatment is afforded b> liquid radwaste system BWR-1,
namely, a three day holdup of all liquid waste steams. Of the estimated
3500 Ci/year discharge from this system, about 45% originates from
"clean" liquids and roughly 54% is derived from chemical wastes. The annual
cost for this system is estimated at about $180,000 not including the
cost of structures (Figure 6).
BWR-2 (Figure 7) represents a formerly typical liquid radioactive
waste treatment system in design but is sized for a two-unit (1000 MWe
each) BWR power station. This system is used in BWRs now operating but
many of these systems are being upgraded. Clean liquids are filtered
and demineralized, allowing a 90% recycle of such liquids. Dirty
liquids and laundry wastes are filtered prior to discharge. Chemical
wastes are evaporated prior to discharge. Of the estimated 30 Cl/year
Annual costs Include amortization and operating costs. 43
-------
Table 13
Classes of BWR Liquid Radioactive
Wastes
(2 units, 1000 MWe each)
(1) Clean Liquids (reactor grade water)
Drywell equipment drains
Reactor building equipment drains
Turbine building equipment drains
Condensate demineralizer backwash
Cleanup filter-demineralizer backwash
(2) Dirty Liquids (non-reactor grade water)
Drywell floor drain sumps
Reactor building floor drain sumps
Radwaste building floor drain sumps
Turbine building floor drain sumps
(3) Chemical Wastes
Condensate demineralizer regeneration
Decontamination drains
Laboratory drains
Shop decontamination solutions
(4) Laundry Waste
Personnel decontamination (showers)
Regulated shop drains
Laundry drains
Cask cleaning drains
44
-------
Clean Liquids (30,000 gpd)
4 tanks
50,000 gal each
(3 day holdup)
Non-tritium
Radioactivity
Release (Ci/yr)
-» 1600.
Dirty Liquids (15,000 gpd)
4 tanks
20,000 gal each
{3 day holdup) ^.
18.
Chemical Wastes ( 1.200 end)
4 tanks
2,500 gal
each
(3 day holdup)
-> 1900.
Laundry Wastes
(1.000 gpd)
4 tanks
2,000 gal
each
(3 dav holdup) v
r
0.04
3518
I/I
Figure 6. Liquid Case BWR-1: Source Term
-------
Condensate
Storage ^- — -
£ Tanks
Clean Liquids
Dirty Liquids
_ 1 L90%_ Recycle),
Precoat
*\* Filter
Surge Tank
75,000 gal
Collector Tank
25,000 gal
Floor Drains
Tank
40,000 gal
40C
Pr
Fi
J
g
ec
It
•-J^
|
D«
pm
oat
er
150 gpm
Chemical Wastes
Chemical
Waste Tank
25,000 gal
E
mineral]
400 gpn
Jx^4
raporato:
10 gpm
1 Non- tritium
Radioactivity
, Release (Ci/yr)
, Release (10%) 5,Q
Sample Tanks
2@ 25,000 ga
Recycle (90%)
Sample Tanks
2@ 20,000 ga! ^ 10-
Samp le v ^ ^
Tanks > 6'9
2@ 5,00) gal
Cartridge ^^-^
Filter
Laundry Wastes
Laundry
Waste Tanks
2@ 5,000 gal
|
-^ U.U^-J
30.
Figure 7. Liquid Case BWR-2: Presently Operating
-------
Condensate
Storage 4- —
Tanks
Su
75
Clean Liquids
Co
25
Dirty Liquids
i
J/
rge Tank
,000 gal
Hector Tank
,000 gal
Floor Drains
Tank
40,000 gal
Pre<
Fil
Pre
Pi]
I
:o
|
&f
^
ic
|
1
Re eye
at
r
De
le>
ninerali
400 gpm
oat ..
er ^^ ^X.
_ _ _ _ _ Non-tritium
1 Radioactivity
I Release (Ci/yr)
, Release (10%) q n
~f~
!er sample Tanks
2@ 25,000 gal
\
Recycle (90%)
Deminerallzer Sample Tanks
150 gpn
2@ 20,000 gal '
Release (50%)
35
Chemical Wastes
Laundry Wastes
Chemical
Waste Tank
25,000 gal
Laundry
Waste Tanks
2@ 5,000 gal
Cat
Fil
f ^1 2@ 5,000 gal
E\aporator
20 gpm
k^ ^
^^ . — - Sample
•tridge [ | Tanks
Lter \^^\, 2@ 2,50C
V sT ^^
^ F^s?inrtT*atQT"] .-. . i
^
% 10 gpm
^
Clean Liquids
Collector Tank
(507,)
gal
15 gpm
Figure 8. Liquid Case BWR-3: Improved Design
-------
Condensate
00
(100% Recycle)
Tanks
Clean Liquids
I
1
Surge Tanks
2@ 75,000 gal
Collector Tanks
2@ 25,000 gal
p-rt
Fi
/.r\{
a p
fa
I
1
1
j-jat"
er
Dei
^X~\^
lineralis
400 gpm
\ S
er
i
\
\
Sample Tanks
3@ 25,000 ga
(No Release)
Non-tritium
Rad ioa c t ivi ty
Release (Ci/yr)
0.0
Dirty Liquids
Floor Drains
Tank
40,000 gal
Pr«
Fi!
;c
1
oat
er
E\
r> ^
><
aporator
20 gpm
^ \
mineral!
25 gpm
ser
Sample Tanks
2@ 20,000 ga
150 gpm
Chemical Wastes
Chemical
Waste Tank
25,000 gal
E
P
l/aporal
20 gi
De
\
toi
jm
ainerali:
25 gpm
;er
Sample Tanks
2@ 5,000 gal
C
C
Laundry Wastes
Laundry
Waste Tanks
2@ 5,000 gal
Cart
Fi
ra
It
I
1
Ldge
:er
E\
De:
r\
r^
aporator
10 gpm
ninerali:
25 gpm
:er
Sample
Tanks
2@ 2,50(
gal
15 gpm
Figure 9. Liquid Case BWR-4: Maximum Treatment
0.0048
Release (507=,) 0.35
Clean Liquids
Collector Tank
(50%)
0.000047
0.35
-------
release, almost 61% comes from the release of dirty liquids. Annual
cost for this system, not including structures, is estimated at
$401,000.
An example of an improved BWR liquid radwaste system is shown as
system BWR-3, Figure 8. This technology is planned, for plants now-being
built. Clean and dirty liquids are filtered and demineralized. For the
purposes of determining radioactivity discharges it is assumed that 10%
of the clean liquids and 100% of the dirty liquids are discharged,
although it may be possible to recycle a portion of the treated dirty
liquids. Chemical wastes and laundry wastes are evaporated but 50% of
the chemical wastes are recycled after treatment. Of the estimated 9.1
Ci/year release, 55% originates from clean liquids and almost 39% from
chemical wastes. Annual cost for this system without structures is
estimated at $560,000.
"Maximum" treatment is afforded each waste stream in system BWR-4,
Figure 9. Clean liquids are filtered and demineralized but additional
tankage is added to assure complete recycling of these liquids. Dirty
liquids, chemical wastes, and laundry wastes are evaporated and
demineralized; 50% of the chemical wastes are recycled. Virtually all
of the estimated 0.35 Ci/year release originates from chemical wastes.
Estimated annual cost for this system is $788,000.
In order to calculate radioactivity releases, the plant parameters
detailed in Table 11 were utilized to derive the source term, system
BWR-1 in Table 16. Table 14 was constructed to show the total
decontamination factor (DF) a
-------
Table 14
DFs for BWR Liquid Systems
Treatment
System
BWR-1
Clean
Dirty
Chemical
Laundry
BWR- 2
Clean
Dirty
Chemical
Laundry
BWR-3
Clean
Dirty
Chemical
Laundry
BWR-4
r*1 a*ar»
i/j.cciii
Dirty
Chemical
Laundry
I
1
1
1
1
102
1
102
1
2
10
102
1033
IO3
103
Cs.Rb
1
1
1
1
10
1.
104
1
10
10
io4
102
5
IO5
IO3
Total DF
Moa
1
1
1
1
IO2
1,
IO6
1
io2
IO2
10°
IO4
Recycled
10
IO4
Ya
1
1
1
1 .
10
1
IO5
1
10
10
105
IO3
10*
10
IO3
Others
1
1
1
1
IO2
1
IO3
1
2
IO2
io3
10*
10
IO3
Q O
Includes an additional DF of 10 for Mo and 10 for Y to account
for plateout, filtration, and demineralization, where applicable,
50
-------
Table 15
Liquid Radwaste System Component DFs
Components Decontamination Factors (DFs)
I Cs.Rb Y Mo Others
DEMINERALIZERS
- PWR -
Mixed Bed (LioBO, form, CVCS)^1'2^ 10 2 1 1 10
Mixed Bed (steam generator blow-
down) (3»6»9) 10 2 1 1 50
- BWR -
Mixed Bed (H-OH form, clean 2
waste) Uill) 102 10 1 1 10
- PWR & BWR -
Mixed bed in evaporator
condensate^ » ' 10 10 11 10
EVAPORATORS (PWR & BWR)
(5 71 2444'
Waste^'/ 10 10, 1(T 10, 10;
LaundryUJ 10 10 10Z 10Z lO1*
OTHER
Removal of Mo and Y by plating
out, filtration, demineraliza-
tion, etc. ' 10 10
51
-------
Table 16
Radionuclide
Mn-54
Fe-55
Fe-59
Co-58
Co-60
Sr-89
Sr-90
Y-91
Zr-95
Nb-95
Ru-103
Ru-106
Cs-134
Cs-137
Ce-141
Ce-144
1-131
1-133
Releases of Long-Lived Radionuclides for
BWR Liquid Radwaste Systems
Annual Release3 (Ci/yr)
Treatment Option
Half-Life BWR-1 BWR-2 BWR-3
303.0 d
2.6 y
45.0 d
71.3 d
5.26y
52.0 d
28.1 y
58.8 d
65.0 d
35.0 d
39.6 d
367.0 d
2.05y
30.0 y
33.0 d
284.0 d
8.0 d
21.0 h
Total non-tritium
releases ,
Total Tritium
7(0)
8(2)
8(0)
9(2)
8(1)
4.4(2)
9.9(0)
0(2)
2(0)
6.5(0)
2.7(0)
6(0)
4(1)
6.3(1)
9.1(0)
5(0)
9(2)
2.
5.
1.
7.
4.
5.
2.1(2)
3500.0
200.0
2.4
30.0
200.0
9.8(-2)
9.1
200.0
BWR-4
6.2(-l)
4.9(-l)
1.7 (-2)
1.0(0)
1.3(-1)
1.4(0)
l.O(-l)
4.2(0)
1.6(-2)
1.7 (-2)
1.0(-2)
3.9(-3)
2.0(-1)
1.6(-1)
4.5<-2)
1.1 (-2)
6.0
8.1 (-3)
l.l(-l)
2. 6 (-3)
1.8(-1)
2.8(-2)
2.8C-1)
2.7(-2)
2.2(0)
3. 2 (-3)
3.8C-3)
l,8(-3)
9.3(-4)
1.0 (-1)
8.5(-2)
6. 7 (-3)
2.7C-3)
2.7(0)
1.9 (-4)
7.5(-3)
1-3 (-4)
l.K-2)
2.K-3)
1.7(-2)
2.K-3)
2.9(-5)
2.1 (-4)
2.8(-4)
l.OC-4)
7.0(-5)
3.3C-4)
2.8(-4)
2.8(-4)
2.0(-4)
2.6(-l)
9.7(-4)
0.35
130.0
aFor a two unit, 1000 MWe each, BWR Power Station. Releases are
written in exponential notation, i.e., 8.8(-l) = 8.8 x 10"1 = 0.88,
A conservative estimate based partly on operating experience
52
-------
Table 15. After taking into account the total DFs afforded each waste
stream, and allowing appropriate credit for recycle where applicable,
total radioactivity releases were determined for each of the four
systems. Table 16 shows the estimated releases by system for the
longer-lived or more significant radionuclides as well as total radio-
activity releases for each system. Tritium releases were estimated to
be about 200 Ci/year for all systems except the maximum treatment
system, where a complete recycle of clean liquids reduces tritium
discharges to about 130 Ci/year.
PWR Liquid Radwaste Systems
The liquid radioactive waste treatment system in a PWR power station
is also responsible for decontaminating a wide variety of waste liquids.
These liquids may be divided into five general classes, each of which
has one or more components as shown in Table 17.
In general, four PWR liquid radioactive waste treatment systems were
constructed to illustrate both the spectrum of treatment options
available as well as the development that has taken place in such
systems to reduce radioactivity releases. These systems are sized for a
two unit (1000 MWe per unit) PWR power station and are shown in Figures
10 through 13.
Minimum treatment is provided by liquid radwaste system PWR-1,
Figure 10. Clean and dirty liquids are released after a three day holdup
while laundry wastes are afforded a 30-day holdup prior to release.
Steam generator blowdown and turbine building drains liquids are
released without any holdup. Of the estimated 3600 Ci/year discharge,
about 70 percent originates from clean liquids and nearly 30% comes
53
-------
Table 17
Classes of PWR Liquid Radioactive
Wastes
(2 units, 1000 MWe each)
(1) Clean Liquids (reactor grade water)
Reactor coolant pump seal leakage
Equipment leakage
Valve leakoffs
Reactor vessel flange leakoffs
Resin flush
Filter changes
Heat exchanger, pump, and tank maintenance
CVCS letdown
(2) Dirty Liquids (non-reactor grade water)
Auxiliary building floor drains
Equipment leakage
Containment sump
Fuel building sump
Chemical laboratory drains
Decontamination area drains
(3) Steam Generator Slowdown
Steam generator blowdown
(4) Turbine Building Drains
Secondary system leakage to turbine building drains
(5) Laundry Wastes
Hot shower drains
Laundry
54
-------
Clean Liquids
(5,500 gpd)
4 Tanks
10,000 gal each
Non-tritium
Radioactivity
Release (Ci/yr)
2500.
Dirty Liquids
C1.200 gpd)
4 Tanks
2,500 gal each
1000.
Steam Generator Slowdown (43,200 gpd)
120.
Turbine Building Drains (14,400
0.051
,Laundry Wastes (50 epd)
4 Tanks
2,000 gal ea
s 0.017
ch 3600.
Figure 10. Liquid Case PWR-1: Source Term
-------
Ui
Wastes
Clean Liquids
^Dirty Liquids
Waste Holdup
Tanks
2(? 25,000 gal
F:
1C
LI
I
^
)
ier3
ET
epm
raporatoi
5 gpm
Laundry
Waste Tanks
2(? 1,000 gal
Filter
Sample Tanks
2@ 1,000 gal
Non-tritium
Radioactivity
Release (Ci/yr)
14.0
Steam Generator Slowdown
120.
Turbine Building Drains
Cartridge Filter
0.051
134
Figure 11. Liquid Case PWR-2: Presently Operating
-------
Uondensate
Storage ^ ~
Tanks
Clean Liquids
fHrf-y T.iqin'^s
Laundry Wastes
Wast
Tank
2@ 2
-*
e Holdup
s
5,000 gal
Waste Holdup
Tanlrc
2@ 10,000 ga
(90% Recycle)
~ ____--j
i
^^\ 1 R
PDenineralizer Sample Tanks
25 gpir 2(§ 5,000 gal
ff L J
^ 20 eon
— *7 ^ » YMYI ^^^^x. j^^"^
1
FiUer3
Laundry ^ Sample
Waste j; ranks
Tauks ^i 2(a 1 00 )
2C 1,000 gal | ^L i>UUJ
Release (10%)
Recycle (90%)
Non-tritium
Radioactivity
Release (Ci/yr)
0.54
0.017
5 gal
Steam Generator Slowdown
Holdup Tanks!
i
2@ 30,000 gall
\
I
D
jmineral
(Cation
50 gpra^
.zer
De
^ ^
mineral!
(Anion)
50 gpm
ser
Sample Tanks
2(? 10,000 ga!
»
4.4
^Turbine Building Drains
Cartridge Filter
0.051
5.0
Figure 12. Liquid Case PWR-3: Improved Design
-------
Condensate
Storage ^- ______
ui Tanks
°° i
t
Waste Holdup
Clean Liquids Tanks
3@ 25,000 gal
Waste Holdup
Dirtv Liquids Tanks
3@ 10,000 ga:
Laundry
Latmdrv Wastes Waste
" Tanks
2@1,000
_ (90%_R
Filt
^
esycle) _ Non- tritium
i Radioactivity
^x ' Release (Ci/yr)
X \ i Release (1070) Q ,,
/^-\ Denineraliuer ^araPle '
( \ . ranks
e*a^X'~Xvv-s, 35 com ^ 5,000
Evaporaro: ^^^ Re^cle (90%)
^ 25 gpm
— 25 gpm ^x^/ ^-""v\
Filtet
^ B
gal ^
10 gp
Fil
-^ „ „, , Holdup Tanks
Steam Generator Slowdown ^
2(3 30,000 gsl
^ X,
ADiimineralizer £anJP''-e
lanks \ n nnnm i
10 gpij 2(3 500 gal
10 gpm ^.
ra^-^^ . — . Denineraliuer Sample
r^J Tanks — > 0.055
lPr& ^f^^L 5° Spm ^@ 20,000 gal
>• Ei7aporato"s ^ J L . . .
| 2@ 25 g],m
^ ^
100 eom \ ^ ^ \
Turbine Building Drains i H°lduP Tankfj |
/ \ D( mineralize r .n,anif e
/ \ j lanks v o onnci^f.
as^ ^ 25 gpm j 2(d 10,000 gal Q 6
Evaporate:1 x. /*
2@ 20,000 gal ^ SPm
^Cartridge Filter
25 gpm
Figure 13. Liquid Case PWR-4: Maximum Treatment
-------
from dirty liquids. Annual cost for this system is estimated at $52,000
without structures.
System PWR-2 (shown in Figure 11) is typical of many PWRs now
operating. Clean and dirty liquids are evaporated prior to release,
while laundry wastes are only filtered and released. Steam generator
blowdown and turbine building drains are released without treatment. Of
the estimated 134 Ci/year which are discharged, almost 90% may be
attributed to steam generator blowdown. Estimated annual cost for this
system is $121,000.
An example of an improved system presently planned for PWRs being
built is shown as system PWR-3 in Figure 12. Clean and dirty liquids
are evaporated and demineralized, allowing a 90% recycle of these
liquids. Steam generator blowdown is passed through two demineralizers
in series prior to release. Laundry wastes are filtered and discharged
while turbine building drains liquids are released untreated. The
estimated annual release is reduced to about 5 Ci/year, of which
almost 90% is derived from steam generator blowdown. Annual cost for
this system, without structures, is about $280,000.
"Maximum" treatment of each waste stream is provided in system PWR-A,
Figure 13. All waste streams are evaporated and demineralized. Extra
tankage and processing capability is added to assure 90% recycle capa-
bility for clean and dirty liquids and to lower radioactivity concentra-
tions of recycled liquids. All other effluent streams are released after
treatment. Total release for this "maximum" treatment alternative is
estimated at 0.60 Ci/year of which 90% is derived from the discharge of
59
-------
Table 18
Treatment System
DFs for PWR Liquid Systems
Total DFa
Cs.Rb
Others
PWR-1
Clean 1 1
Dirty 1 1
S. G. Slowdown 1 1
Turbine Bldg 1 1
Laundry 1 1
PWR-2
Clean 102 10^
Dirty 102 10^
Laundry 1 1
S. G. Blowdown 1 1
Turbine Bldg 1 1
PWR-3
Clean 10^ 10^
Dirty 103 105
Laundry 1 1
S. G. Blowdown 102 4
Turbine Bldg 1 1
PWR-4
Clean 1(T W
Dirty 10:? 10:?
Laundry 10^ 10^
S. G. Blowdown 10^ 10^
Turbine Bldg 10 10
1
1
1
1
1
10C
10«
1
1
1
10
10
10
10
10*
1
1
1
1
1
10;
10-
1
1
1
10;
10-
1
10
1
10;
10;
10;
10;
10-
i
i
i
i
i
10;
io:
i
i
i
104
10*
1 ,
2.5x10''
1
10
10:
10
aExcludes DF exerted by CVCS system on liquids eventually discharged.
bIncludes an additional DF of 102 for Mo and 10 for Y to account
for plateout, filtration, and deminerallzatlon, where applicable.
60
-------
Table 19
Releases of Lang-Lived Radionuclides
For PWR Liquid Radwaste Systems
Annual Release3 (Ci/yr)
Radionuclide Half-Life
PWR-1
Treatment of Option
PWR- 2
PWR-3
PWR-4
Mn-54
Fe-55
Fe-59
Co-58
Co-60
Sr-89
Sr-90
Y-91
Zr-95
Nb-95
Ru-103
Ru-106
Cs-134
Cs-137
Ce-141
Ce-144
1-131
1-133
303.0 d
2.6 y
45.0 d
.3 d
,26y
71.
5.
52.0 d
28.1 y
58.8 d
65.0 d
35.0 d
39.6 d
367.0 d
2.05y
30.0 y
33.0 d
284.0 d
8.0 d
21 h
8.8C-1)
2.5(0)
7.2(-l)
2.4(1)
2.5(0)
1.3(0)
4. 4 (-2)
9.7(0)
2.2(-l)
2.K-D
1.6(-1)
4. 2 (-2)
2.9(2)
2.7(2)
2.4(-l)
1.4(-1)
8.3(2)
4.4(2)
3.3(-2)
9.K-2)
2. 7 (-2)
9.H-1)
9.K-2)
6.1 (-2)
2.0 (-3)
4.8(-l)
9.9(-3)
9. 9 (-3)
7.K-3)
1.8 (-3)
4.6(0)
4.0(0)
1.0 (-2)
6.2(-3)
4.1(1)
1.8(1)
1.6(-4)
4^4(-4)
1.2 (-4)
3.6(-l)
3. 6 (-2)
2. 8 (-4)
9. 5 (-6)
4. 8 (-2)
4. 6 (-5)
4. 8 (-5)
3. 3 (-5)
8. 6 (-6)
1.2(0)
1.0(0)
4.7(-5)
2.9(-5)
4.3(-l)
1.9(-1)
1.2C-5)
3.3(-5)
9.5(-6)
3.2(-4)
1.8(-6)
1.9 (-5)
6.2(-7)
1.4(-5)
3.K-6)
3.0 (-6)
2. 3 (-6)
5. 8 (-7)
3. 4 (-4)
3.1 (-4)
3. 3 (-6)
1.9(-6)
1.1 (-1)
5, 7 (-2)
Total non-tritium 3600.0 134.0 5.0 0.60
releases
Total Tritium 1200.0 1200.0 760.0 760.0
For a two unit, 1000 MWe each, PWR Power Station. Releases are
written in exponential notation, i.e., 8.8(-l) - 8.8 x KT1.
Based on operating experience of smaller plants
61
-------
treated clean and dirty liquids. Estimated annual cost for this system
is $879,000 with structures.
In order to calculate radioactivity releases, the plant parameters
detailed in Table 12 were utilized to derive the source term, system
PWR-1 in Table 19. With this source term, Table 18 was constructed to
show the total decontamination factor (DF) accorded each waste stream in
each radwaste system. Individual component DFs, which make up these
system DFs, are shown in Table 15. After taking into account the total
DFs afforded each waste stream in each system, and allowing appropriate
credit for recycle where applicable, total radioactivity releases were
determined for each of the four systems. Table 19 lists these radio-
activity releases by system for the longer-lived radionuclides as well
as total radioactivity releases for each system. Tritium releases are
estimated to be about 1,200 Ci/yr for the two-unit (1000 MWe each) PWR
power station using liquid radwaste systems PWR-1 or PWR-2. A tritium
release of 760 Ci/yr is estimated for the same PWR power station using
liquid radwaste systems PWR-3 or PWR-4.
Cost Analysis
Having selected four PWR and four BWR liquid radwaste systems, the
major components of each were broken down as shown In Tables 20 and 21.
In order to determine annual costs for each system, a fixed charge rate
of 16.6% of the capital investment (without structures) was added to
operating and maintenance costs (1). Capital costs were estimated from
reference (1) and supplemented by estimates of the nuclear industry (13,
14, 15, 19-24). Aside from being scarce, available estimates of
operating and maintenance costs are quite variable for similar equipment
62
-------
Table 20A
Equipment, Annual, and Capital Costs3 (BWR)
BWR-1 BWR-2 BWR-3 BWR-4
Equipment Items (Number of Equipment Items Required for Given Systems)
Tankage (gal)
2,000 4
2,500 4
5,000
20,000 4
25,000
40,000
50,000 4
75,000
Filters (Precoat) (gpm)
150
400
Filter (Cartridge) (gpm)
15
Demineralizers (gpm)
25
150
400
Evaporators (gpm)
10
20
Estimated Capital Cost $918,000
Estimated Annual Cost $180,000
4
2
4
1
1
1
1
1
1
1
$1,738,000
$ 401,000
2
4
2
4
1
1
1
1
1
1
1
1
1
$2,344,000
$ 560,000
2
4
2
6
1
2
1
1
1
3
1
1
2
$3,231,000
$ 788,000
aWithout structures.
63
-------
Table 2OB
Equipment, Annual, and Capital Costs (PWR)
PWR-1 PWR-2 PWR-3 PWR-4
Equipment Items
Tankage (gal)
500
1,000 4
2,000 4
2,500 4
5,000
10,000 4
20,000
25,000 2
30,000
Filters (cartridge, gpm)
10 2
25
100
Demineralizers (gpm)
10
25
35
50
Evaporators (gpm)
5 1
20
25
2
2
2
2
2
2
1
1
1
1
2
1
2
2
2
5
4
3
2
1
2
1
1
1
1
1
1
4
Estimated Capital Cost $264,000 $509,000 $1,213,000 $3,547,500
Estimated Annual Cost $52,000 $121,000 $ 280,000 $ 879,000
•a
Without structures.
64
-------
Table 21
LIQUID RADIOACTIVE WASTE SUMMARY TABLE: BWHs AND PWRa
Pressurised Water Reactors (d)
Source Term
Presently Operating
Improved
Maximum Treatment
Boiling Water Reactors (d)
Source Term
Presently Operating
Improved
Maximum Treatment
System
Designation
nm-i
FUR- 2
FUR- 3
PWR-4
BWR-1
BWR-2
BWR-3
BWR-4
Estimated
Capital
Coat
$ 264,000
$ 509,000
Si, 213,000
53,547,000
$ 918,000
$1,738,000
$2,344,000
$3,231,000
Estimated
Annual
Coat
$ 52,000
$ 121,000
$ 280,000
$ 879,000
$ 180,000
$•401,000
$ 560,000
$ 788,000
LIQUID HOH- TRITIUM RADIOACTIVITY RELEASE
(Cl/yr)
Clean
2500.
14.
0.54
0.54
1600.
5.
5.
(c)
Dirty
1000.
(a)
(a}y>
(a)
18.
18.
0.55
0,0043
Chemical
Waste
1900.
6.9
3.5
0.35
Laundry
Waste
0.017
(a)
0.017
00
0.04
0.04
0>>
0>>
Steam
Generator
Slowdown
120.
120.
4.4
0.055
Turbine
Drains
0.051
0.051
0.051
(b)
TOTAL
3600
134
5.0
0.6
3518
30
9ll
0.35
TRITIUM
(Ci/vr)
1200
1200
760
760
200
200
200
130
(a) Included wlth.clean liquids
Cb) less than 10~3 Cl/yr
(e) Mo release, 1001 recycled
(<•) Values are for two units, 1000 HWe each.
-------
items (12,13,14,15). Therefore, operating and maintenance costs were
estimated at 3% of the capital cost for tankage and 10% of the capital
cost for filters, evaporators, and demineralizers. Table 21 relates the
non-tritium activity release to annual costs for BWR and PWR liquid
radwaste systems.
Estimated costs are dependent upon the specific systems chosen.
These systems in turn are related to the mode of plant operation
assumed. For example, it was assumed above that the BWR operated with
deep bed condensate demineralizers. Also available are Powdex filter-
demineralizers. Although the on-line operating costs of the Powdex
units appear less than those of the deep bed condensate demineralizers,
operating experience with the Powdex units in BWRs is limited and the
increased cost of shipping the spent resins (Powdex units are not
regenerated) tends to increase total operating costs. In PWRs,
different reactor vendors handle secondary system cleanup in different
ways. Westinghouse prefers to blowdown the steam generators at a
controlled rate whereas Babcock and Wilcox employs full-flow condensate
demineralizers (3, 12, 16). Although B & W plants avoid a continuous
blowdown stream from the steam generators, the solution eluted from
regenerating the condensate demineralizers must be treated when
contaminated by primary-to-secondary leakage. North Anna 3 and 4 will
employ a 25 gpm evaporator for this purpose (14). Combustion
Engineering employs both steam generator blowdown (though at a lower
flow rate than Westinghouse plants) and partial-flow condensate
demineralizers (7, 17). Liquids from blowdown and condensate
demineralizer regeneration may be handled by a liquid radioactive was -e
disposal system of slightly larger capacity. The use of condensate
demineralizers in B & W and C-E PWRs necessitates cleanup by evaporation
66
-------
upon regeneration after operating with primary-to-secondary leakage.
This increases the annual costs of secondary system cleanup by almost a
factor of three but would increase total liquid radwaste system annual
costs for PWR-3, Table 21, by only about 30%.
As shown by the selection of liquid radwaste systems, no redundancy
of critical radwaste system components is assumed. However, some
inherent redundancy does exist in systems presently planned (PWR-3 and
BWR-3) for BWRs and PWRs. Recent operating reports for a few Westinghouse
PWRs show, for example, that boric acid evaporators (of the Chemical and
Volume Control System) have been used to treat steam generator blowdown although
not necessarily a desirable alternative. PWRs of Babcock and Wilcox design or
Combustion Engineering design typically provide evaporators which may be used
interchangeably for boron recovery or miscellaneous liquid waste
processing (3, 7, 16, 17). BWRs with only two evaporators (laundry
waste and chemical waste) and only two demineralizers (clean liquids and
dirty liquids) could share equipment between these respective systems to
achieve lowest practicable releases. Sizings indicated in system BWR-3,
Figure 21, would allow this sharing until repairs on the defective
evaporator and/or demineralizer could be completed. Therefore, it
appears that sufficient flexibility exists in BWR and PWR liquid
radwaste systems presently planned to maintain radioactivity releases
and dose equivalents very close to estimated levels.
Finally, estimated costs for liquid systems do not include
structures and the costs of appropriate solid waste systems.
Consideration has been given to the fact that the capital cost of
s' ructures may be on the order of a few million dollars and that solid
67
-------
waste system capacity must be increased with more sophisticated liquid
radwaste systems, resulting in perhaps one-half million dollars in
capital cost (without structures) and $200,000 - $500,000 in annual
cost0 (1). On the other hand, building space and solid waste systems
aro already planned for integration with liquid radwaste systems as
sophisticated as BWR-3 and PWR-3, respectively. Although these items do
not directly reduce radioactivity in liquid effluents, their respective
costs must also be considered.
68
-------
NOBLE GAS DISCHARGE CONTROL OPTIONS
Pressurized Water Reactors (PWRs)
Radioactive isotopes of krypton and xenon, the noble gases of
greatest concern, are generated inside the fuel rods of pressurized
water reactors. These gaseous radionuclides escape through fuel cladding
defects and enter the primary reactor coolant system. Due to
leakages of primary coolant, intentional or otherwise, the radioactive
isotopes of krypton and xenon may be released to the environment. These
release pathways may be broken down as follows:
(1) Primary Gases
Shim Bleed
Shutdown Degasification
(2) Secondary System Gases
Air Ejector Exhaust
Gland Seal Exhaust
Steam Generator Slowdown Tank (SGBT) Vent
(3) Building Ventilation
Containment Purge
Auxiliary Building
Turbine Building
Whereas the primary gas sources (shim bleed and shutdown degasification
may be routed to the waste gas treatment system, those gases resulting from
primary coolant leakages (secondary system gases and building ventilation)
usually escape to the atmosphere untreated.
In order to determine the effectiveness of various PWR noble gas
discharge control options, the source term for each of the release
pathways detailed above was calculated. Assumptions for calculating
69
-------
•vl
O
Nuclide
Kr-83m
Kr-85m
Kr-85
Kr-87
Kr-88
Kr-89
Xe-131m
Xe-133m
Xe-133
Xe-135m
Xe-135
Xe-137
Xe-138
Tota_s
Half-
life (31
1.86h
4.4h
10.76y
76m
2.8h
3.2m
11. 8d
2.26d
5.27d
15.6m
9.2h
3.9m
17.0m
PWR Nol
Primary
System
Gases
4.2(2)a
2.3(3)
1-6(3)
1.3(3)
4.0(3)
6.0(1)
1.8(3)
4.2(3)
3.2(5)
2.4(2)
6.6(3)
1.3(2)
8.3(2)
3.4(5)
*
Table 22
>le Gas Source Term: 2 Units, 1,
Secondary System Gases
Air Gland
Blowdown Ejector Seal
Vent Exhaust Exhaust
— b 3.0(0)
1.6(1)
— 1.1(1)
— 8.9(0)
2.8(1)
6.7(-l)
1.3(1)
— 3.0(1)
2.3(3) 2.0(0)
2.0(0)
— 4.8CD
— 1.4(0)
6.6(0) —
— 2.5(3) 2.0(0)
000 MWe Each (Ci/yr)
Building Ventilation
Containment Auxiliary Turbine
Purge Building Building
— 3.0(0)
1.6(1)
2.2(1) 1.1(1)
8.9(0)
l.O(-l) 2.8(1) —
6.7(-l) —
4.8(0) 1.3(1) —
2.1(0) 3.0(1)
3.8(2) 2.3(3) 1.0(0)
2.0(0)
5.8(-l) 4.8(1) —
— 1.4(0)
6.6(0)
4.1(2) 2.5(3) 1,0(0)
Total
4.2(2)
2.3(3)
1.6(3)
1.3(3)
4.0(3)
6.1(1)
1.8(3)
4.3(3)
3.2(5)
2.4(2)
6.7(3)
1.3(2)
8.4(2)
3.4(5)
4.3(4) =4.3 x 10 or 43,000
— implies less than 0.1 Ci/yr
-------
these source terms are summarized in Table 12 and are based on
operating experience where possible (2, 6, 9, 51) and/or generally
acceptable values (1-3, 7, 24, 26-28). Table 22 presents the
source term for each of these PWR noble gas release pathways for a two
unit (1000 MWe each) PWR power station. Almost 100% of the total of
these sources results from the primary gases (shim bleed and shutdown
degasification). Therefore, the majority of noble gas discharge control
options have been designed to control these gases.
The noble gas discharge control options considered fall into five
general classes in addition to the source term. The first class of options
consists of pure physical holdup of primary gases (by pressurized tanks with
or without the use of a recombiner) for 15, 30, 45, or 60 days and the delay
of primary gases on charcoal adsorption beds, providing 15, 30, 45, and 60
days delay for xenon and 1, 2, 3, and 4 days for krypton (1, 24, 29, 32,
34, 35). Flow rate of radioactive primary gases was taken as 0.5 scfm from
each unit (1.0 scfm total flow rate). Costs have been estimated from
available literature (1, 13, 21, 22, 24, 36-39). Comparison of pure
physical holdup with charcoal adsorption reveals virtually the same total
releases for xenon holdup times of 15, 30, 45, and 60 days. For shorter
holdup times, however, slightly more krypton will be released from the
charcoal adsorption beds.
The second class of PWR noble gas discharge control options
considered consists of treating primary gases with either a selective
absorption system or cryogenic distillation. In each case, the noble
gases xenon and krypton are concentrated, by means of solubility into a
71
-------
fluorocarbon solvent in the former case, and by temperature effects in
the latter case. As these options have had limited operational experi-
ence, the cost and effectiveness estimates appear less certain. Conse-
quently, two decontamination factors (DFs) for xenon, 1,000 and 10,000,
are assumed for each of these options (1,37,41,45,46), resulting in a
range of estimates for releases.
Krypton DFs are assumed to be 25% of the xenon DFs, in accordance
with assumptions presented in recent safety analysis reports. However,
once either set of noble gas DFs is applied to the primary system off
gases, the variation in total release is small as secondary sources pre-
dominate. Cost estimates show considerable variability Cl»21,22,38).
This second class of noble gas discharge control options results in a
release of noble gases similar to the first class of options providing
60 day delay or holdup. .
A third class of discharge control options may be defined by com-
bining the treatment systems afforded by the first two classes. However,
analysis shows that release of noble gases from primary gases decreases
from 51 Ci/yr to about 6.5 Ci/yr using different holdup times from 15 to
60 days, respectively; this decrease is negligible when compared to total
noble gas releases of about 5,400 Ci/yr for all such system combinations.
Therefore, the third class of discharge control options will be assumed to
be represented by a 15 day xenon delay on charcoal adsorbers and cryogenic
distillation or selective absorption which appears to be the least expen-
sive option.
A fourth discharge control option class is defined as the virtual
elimination of primary gas releases by using the cover gas recycle system.
72
-------
Volatile radioisotopes are continuously removed from the primary coolant
system by a constant purge of the volume control tank. This hydrogen-
fission gas mixture is sent to the cover gas recycle system, where it is
diluted with nitrogen, passed through a compressor and recombiner, and
stored in a gas decay tank. By using many gas decay tanks, the nitrogen
in the cover gas recycle system may be recycled indefinitely, allowing
significant decay of all noble gases (except krypton-85) before the nitro-
gen in a given decay tank must be used again. As a result, primary cool-
ant concentrations of the volatile radioisotopes are reduced. The effect,
however, is more pronounced for radioisotopes with longer half-lives (1,47)
Factor of
Reduction in
Primary Coolant
Nuclide Half-life Concentration
Xe-135m
Kr-87
Kr-85m
Xe-135
Xe-133m
Xe-133
Kr-85
15.6 m
76 m
4.4 h
9.2 h
2.26 d
5.27 d
10.76 y
1.0
1.0
1.0
1.3
5.0
7.0
10.0
Thus, cover gas recycle is unique in that the reduction of primary cool-
ant concentrations results in a decrease of the releases of volatile
radionuclides from all gaseous release pathways. Cost estimates are very
limited (1,22).
Finally, a fifth class of discharge control options is defined as
the treatment of noble gas sources other than primary gases. As shown by
the PWR Source Term, Table 22, the bulk of these releases i - ide up of
the releases from the air ejector and the auxiliary building. Because
of the high flow from the auxiliary building (100,000 cfm) (2), it
appears impractical at present to treat this effluent stream. However,
73
-------
charcoal beds may be used to delay the noble gases evolving from a PWR
air ejector (35), as Is commonly proposed for BWRs. Cost and effective-
ness, therefore, are largely based on BWR data (1, 13, 29, 32, 35, 39).
Table 23 illustrates the effectiveness of air ejector charcoal beds in
reducing these noble gases, using 1, 2, 3, and 10 days xenon delay
and a Xe/Kr delay ratio of 15 (1,24,29,35).
In order to place these PWR noble gas discharge control options in
perspective, a summary table, Table 24, was constructed to illustrate
the range of options considered, estimated costs, and activity releases.
It should be noted that the use of compressor-tankage holdup is most
typical of present PWR waste gas systems, although it appears more
expensive than charcoal adsorption. Application of some of the more
sophisticated discharge control options for primary gases results in
releases overwhelmingly dominated by the secondary source contribution.
Only the use of cover gas recycle for primary gases and the use of
charcoal delay beds on the air ejector can reduce the contribution of
these secondary release pathways of noble gases. To illustrate the
change in the technology of controlling noble gas releases, PWRs
formerly provided only a 15-45 day holdup for primary gases whereas
presently planned PWRs typically are providing 60 day holdup. At least
one PWR has proposed charcoal adsorption beds for both primary gases and
air ejector effluent (35), a few have gone to the cover gas recycle
system (4, 47), and at least two plants have proposed adding cryogenic
charcoal adsorption systems (in which the noble gases are adsorbed on low-
temperature charcoal) to existing gas decay tanks (15,57).
74
-------
Table 23
Effectiveness of Charcoal Delay
Beds on Air Ejector (PWR)
Releases (Ci/yr for 2 Units, 1000 MWe each)
Air Xe Delay: Xe Delay: Xe Delay Xe Delay
Nuclide
Kr-83m
Kr-85m
Kr-85
Kr-87
Kr-88
Kr-89
Xe-131m
Xe-133m
Xe-133
Xe-135m
Xe-135
Xe-137
Xe-138
Half-
Life(3
1.86h
4.4h
10.76y
76m
2.8h
3.2m
11. 8d
2.26d
5.27d
15.6m
9.2h
3.9m
17.0m
Ejector
Source
3.
1.
1.
8.
2.
6.
1.
3.
2.
2.
4.
1.
6.
0(0)a
6(1)
1(1)
9(0)
8(1)
7(-l)
3(1)
0(1)
3(3)
0(0)
8(1)
4(0)
6(0)
1 Day
(Kr=0.067d)
8.2(-l)
1.2(1)
1.1(1)
3.7(0)
1.9(1)
—
1.2(1)
2.2(1)
2.0(3)
—
7.9(0)
—
—
2 Days
(Kr=0.133d)
2.3(-l)
9.7(1)
1.1(1)
1.5(0)
1.3(1)
—
l.KD
1.6(1)
1.8(3)
—
1.3(0)
—
—
3 Days
(Kr=0.20d)
->
7.5(0)
l.KD
6.2<-l)
8.7(0)
—
1.0(1)
1.2(1)
1.6(3)
—
2.K-1)
—
—
10 Days
(Kr=0.67d)
—
1.3(0)
l.KD
—
5.2C-1)
—
7.2(0)
1.4(0)
6.1(2)
—
—
—
—
2.5(3) 2.1(3)
1.9(3)
1.6(3)
6.4(2)
Effectiveness based on 10 cfm air in-leakage, Xe/Kr delay ratio of
15, and following amounts of charcoal:
Xe Delay
Charcoal (tons)
Id 3.6
2d 7.2
3d 10.8
f3.0<0) - 3.0 x 10° or 3.0
«- implies less than 0.1 Ci/yr
75
-------
Table 24
SUMMARY TABLE: PWR NOBLE GAS DISCHARGE CONTROL OPTIONS
(2 UNITS, 1000 MWE EACH)
CLASS
0
1A-15
1A-30
" '5
1A-60
1B-15
1B-30
IB -45
IB-60
2A
2B
3
4
SA
5B
SC
3D
PWR NOBLE GAS DISCHARGE
CONTROL OPTION
No Treatment
Charcoal Adsorption
Charcoal Adsorption
Charcoal Adsorption
Charcoal Adsorption
Compressed Tank Holdup
(w/wo recombiner)
" "
t* if
ti ii
Cryogenic Distillation or
Selective Adsorption
n ii
1A-15 + 2A
Cover Gas Recycle
Air Ejector Charcoal
Adsorption 4- Class 1A
" + Class 2A
" 4- Class 3
.." •+ Class 4
DAYS HOLDUP
OR
PROCESS DF
None
15d Xe;ld Kr
30d Xe;2d Kr
45dJCe;3d Kr
60d Xe;4d Kr
ISd
30d
45d
60d
Xe DF - 1,000
Xe DF - 10,000
15d Xe +
Xe DF • 1,000
b
2d Xee
2d Xec
2d Xec
2d Xec
ESTIMATED
CAPITAL
COST
$* o
$360,000
$540,000
$720.000
$900,000
$850, OOO/
/$500,000
$900, OOO/
/$600,000
$950, OOO/
/ $700 ,000
$1,000, OOO/
/$800,000
$1,500,000
$1,500,000
$1,860.000
$2,000,000
$1,260,000
$2,400,000
$2,760,000
$2,900,000
ESTIMATED
ANNUAL
COST
: 0
; 60,000
> 90,000
S120.000
;iso,ooo
1270, OO/
/$164,000
$280, OOO/
/$190.000
$290, OOO/
/$225,000
$300, OOO/
/$250.000
!600,000
$600,000
$660,000
$580,000
$210,000
$750,000
$810,000
$730,000
RELEASE (Ci/yr)
Kr-85
1.6(3)
M
II
II
II
II
II
H
II
5.0(1)
4.5(1)
5.0(1)
4.4(0)
1.6(3)
5.0(1)
5.0(1)
4.4(0)
Kr-88
4.0(3)
6.7(1)
5.6(1)
II
II
II
It
tl
II
6.4(1)
5.7(1)
5.6(1)
5.6(1)
5.2U)
4. (1)
4.1(1)
4.1(1)
Xe-133
3.2(5)
4.9(4)
1.1(4)
5.8(3)
5.1(3)
4.9(4)
1.1(4)
5.8(3)
5-1(3)
5.3(3)
5.0(3)
5.0(3)
7.1(2)
4.9(4)
4.8(3)
4.5(3)
6.5(2)
Total a
3.4(5)
5.2(4)
1.3(4)
7.9(3)
7.1(3)
5.2(4)
1.3(4)
7.9(3)
7.1(3)
5.7(3)
5.4(3)
5.4(3)
9.4(2)
5.2(4)
5.1(3)
4.8(3)
7.9(2)
•From all noble gas radionuclides
"Virtually complete holdup of primary gases may be achieved; releases from secondary pathways are also
reduced by about a factor of 5.8.
C2 days xenon delay for air ejector-noble gases pins appropriate primary gas holdup or DF indicated by
dlass of treataottt.
-------
Boiling Water Reactors (BWRs)
As in the PWR, radioactive isotopes of krypton and xenon, the noble
gases of greatest concern, are generated inside the fuel rods. These
gaseous radionuclides escape through the fuel rod cladding defects and
enter the reactor coolant system. There are many leakage pathways which
may allow the radioactive isotopes of krypton and xenon to reach the en-
vironment. These release pathways may be broken down as follows:
(1) Primary Gases
Condenser Air Ejector
Turbine Gland Seal
Mechanical Vacuum Pump (at startup)
(2) Building Ventilation
Reactor Building
Radwaste Building
Turbine Building
Whereas the release through the air ejector is intimately related to
plant operation, other pathways are more or less unplanned since they
occur as leakages from plant components.
In order to determine the effectiveness of various BWR noble gas
discharge control options, the source term for each release pathway was
determined. The releases from the condenser air ejector and turbine
gland seal are based on estimates made by the nuclear industry (29,30).
**-•--
It should be noted that the air ejector source term incorporates a nomi-
nal 30-minute delay. This is typical of all previous BWR plants Cl,29).
Primary coolant concentrations corresponding to the 30-minute air ejector
source term and the assumptions specified for leakage conditions in
various buildings, Table 11, for the basis for the noble gas releases for"
77
-------
oo
Table 25
BWR Source Term: 2 Units, 1,*NX) MWe Each (Ci/yr)
Radio-
Nuclide
Kr-83m
Kr-S5m
Kr-85
Kr-87
Kr-88
Kr-89
Kr-90
Xe-1310
Xe-133m
Xe-133
Xe-135m
Xe-135
Xe-137
Xe-138
Xe-139
Half- Air -Ejector
life (31) Effluent (29.30)
1.86h
4'.4h
10.76y
76m
2.80h
3.2m
33.0s
11. 8d
2.26d
5.27d
15.6m
9.2h
3.9m
17.0m
43.0s
1.5(5?
2.8(5)
7.6(2)
7.6(5)
9.1(5)
9.1(3)
7.6(2)
1.5(4)
4.1(5)
3.5(5)
1.1(6)
3.4(4)
1.1(6)
__ _
Turbine
Gland Seal Reactor
Effluent (30) Buildina
•3.5(2) b
6.1(2)
2.4(0)
1.9(3)
2.0(3)
3.8(3)
3.1(3)
2.0(0)
3.0(1)
8.6(2)
2.4(3)
2.2(3)
1.1(4)
8.6(3)
4.7(3)
Rad waste Turbine
Build inc Building
2.9(1)
4.9(1)
4.2(-l)
1,7(2)
1.7(2)
1.0(3)
2.0(2)
1.2(0)
2.3(0)
6.6(1)
2.1(2)
1.8(2)
1.3(3)
5.9(2)
9.9(2)
Mechanical
Vacuum
Ptimp Totals
-— 1.5(5)
2.8(5)
7.6(2)
7.6(5)
9.1(5)
1.4(4)
3.3(3)
7.6(2)
1.5(4)
4.4(3) 4.1(5)
3.5(5)
6.6(2) 1.1(6)
4.6(4)
1.1(6)
5.7(3)
Totals
5.1(6)
al.S(5) - 1.5 x 105 or 150,000
b— " < 0.1
4.2(4)
5.0(3)
5.1(3)
5.1(6)
-------
building ventilation sources. Table 25 presents the source term for each
of the release pathways discussed above for a two-unit, 1000 MWe each,
BWR power station. Almost 100% of the total release is from the condenser
air ejector. As a result, all but one of the noble gas discharge control
options considered are designed to reduce this release pathway.
BWR noble gas discharge control options may be divided into five
classes; four of these reduce air ejector releases while one option has
been considered for eliminating turbine gland seal effluents. The first
class consists of a single option, physical holdup of air ejector offgases
(via recombiner-compressor-tankage) for one day; this option has been pro-
posed for at least two BWRs (19,40). These plants have stacks, however,
unlike more recent designs. A condenser air inleakage of 10 scfm is
assumed; smaller air inleakage can increase holdup time proportionately
while greater leakage would decrease holdup. Costs are based on available
data (1,21,23,38-41).
The second class of discharge control options consists of the use of
recombiners and varying amounts of charcoal (at 77° F) to achieve xenon
holdup times of 10, 20, 40, and 60 days. However, krypton delay tines of
only 1, 2, 3, and 4 days, respectively, are achieved, based on measured
and suggested values of the Xe/Kr delay ratio (1,24,29,35). Condenser air
inleakage was assumed to be 10 scfm, based on measured and suggested values
(1,29,32-34). Cost estimates appear more certain for this class of options
than any of the others and are based primarily upon nuclear industry esti-
mates (1,13,24,39,42,43). However, this second class of control options
achieves a lower release rate at costs comparable to the first class, al-
though releases are made from a plant vent rather than a 100 m stack.
79
-------
Alternatively, a third class of discharge control options may be
formed by considering the processing of the air ejector effluent through
a selective absorption system or a cryogenic distillation system. In
each case, the noble gases xenon and krypton are concentrated, by means
of solubility into a fluorocarbon solvent in the former case, and by
temperature effects in the latter case. As these options have had limited
operational experience, cost and effectiveness estimates show a wide vari-
ation. Consequently, two decontamination factors (DFs) for xenon, 1,000
and 10,000, are assumed for each of these options (1,37,41,45,46), re-
sulting in a range of estimates for releases. Krypton DFs are assumed at
25% of xenon DFs. However, since secondary release pathways dominate
total releases with either DF, the variation in total releases is small.
Cost estimates exhibit a wide variability (1,13,41). Either of these
options, selective absorption or cryogenic distillation, provides about
the same activity reduction as a 20-40 day delay of xenon on charcoal beds.
By combining the treatment options in the first three classes above,
a fourth set of discharge control options for BWR noble gas air ejector
releases may be defined. Since secondary source terms now dominate, the
variation in total release by using different combinations of classes 1,
2, and 3 is minimal^ Therefore, the combination of a ten-day xenon delay
on charcoal (0.67 day delay for krypton) and either a selective absorption
system or cryogenic distillation system will be taken to represent this
fourth class of noble gas discharge control options.
The fifth. class of BWR noble gas discharge control options is the
only option chosen to reduce noble gas emissions from'a source other than
the condenser air ejector, namely, the turbine gland seal exhaust.
80
-------
Examination of the BWR source term indicates that other than the air
ejector only the turbine gland seal, turbine building ventilation, and
operation of the mechanical vacuum pump contribute significantly to noble
gas releases. However, a source of nonradioactive steam may be used to
block the release of radioactive gases from the turbine gland seal (48),
eliminating this source of noble gas release. Cost estimates are based
on very limited data (1,24,25).
In order to place these BWR noble gas discharge control options in
perspective, a summary table, Table 26, has been constructed to illus-
trate the range of options considered, estimated costs, and activity
releases. In the case of the clean steam options, the only alternative
considered for treating a secondary source of noble gases, only those
options are shown that illustrate a large change in activity releases
when used in conjunction with a given primary gas treatment option.
Application of some of the more sophisticated discharge control options
for primary gases results in releases overwhelmingly made up of the
secondary source contribution. Although presently operating BWRs employ
essentially a Class 0 treatment, almost all are planning to retrofit
equipment to achieve Class 2-10 (charcoal adsorption) releases within
1-3 years. A few BWRs presently planned have proposed cryogenic dis-
tillation systems to treat the primary off-gas from the air ejector as
well as clean steam systems to eliminate the noble gases from the tur-
bine gland seal (58,59).
81
-------
Table 26
SUMMARY TABLE: BWR NOBLE GAS DISCHARGE CONTROL OPTIONS
(2 UNITS, 1000 MWE EACH)
oo
K>
CLASS
0
1
2-10
2-20
2-40
2-60
3A
3B
4.
SA
SB
SC
5D
SB
DISCHARGE CONTROL
OPTION
Source Termb
Recombiner- Holdup -
Stack
Recombiner Charcoal
Adsorption
M «l
-
M It
Cryogenic Distill-
ation or Selective
Absorption
M 1*
2-10 + 3A
CLEAN STEAM -I- CLASSl
CLEAN STEAM -I- CLASS
2-10
CLEAN STEAM * CLASS
2-60
CLEAN STEAM + CLASS 3 A
CLEAN STEAM + CLASS
4
DAYS HOLDUP
OR
PROCESS DP
Id
lOd Xe
20d Xe
40d Xe
60d Xe
Xe DF - 1,000
Xe DF -10,000
lOd Xe^ Xe P?«
1,000
c
c
c
e
c
ESTIMATED
CAPITAL
COST
83,200,000
$6,500,000
$5,600,000
$6,000,000
$7,000,000
$8,000,000
$7,000,000
$7,000,000
$8,8000,000
$7,500,000
$6,600,000
$9,000,000
$8,000,000
$9,800,000
ESTIMATED
ANNUAL
COST
$ 600,000
$1,400,000
$1,200,000
$1,330,000
$1,620,000
$1,910,000
$1,400,000
$1,400,000
$2,600,000
$1,600,000
$1,400,000
$2,110,000
$1,600,000
$2,800,000
RELEASE (Ci/vr)
Kr-85
7.6(2)
ft
M
H
n
ti
7.8(0)
3.1(0)
5.8(0)
7.6(2)
7.6(2)
7.6(2)
3.4(0)
3.4(0)
Kr-08
9.1(5)
4.6(3)
1.9(4)
2.5(3)
2.2(3)
II
5.8(3)
2.5(3)
2.2(3)
2.6(3)
1.7(4)
1.7(2)
3.7(3)
1.8(2)
Xe-133
4.1(5)
3.6(5)
1.1(5)
3.5(4)
7.5(3)
5.5(3)
5.7(3)
5.4(2)
5.4(3)
3.6(5)
1.1(5)
4.6(3)
4.9(3)
4.6(3)
Xe-13B
1.1(6)
9.2(3)
n
ti
n
M
1.0(4)
9.3(3)
9.2(3)
5.9(2)
5.9(2)
5.9(2)
1.7(3)
5.9(2)
Totaia
5.1(6)
6.0(5)
2.0(5)
8.5(4)
5.4(4)
5.2(4)
6.3(4)
5.3(4)
5.2(4)
5.6(5)
1.6(5)
1.1(4)
2.1(4)
1.0(4)
°Total is sum of all noble gas radioisotope activities released.
^Illustrates effects of two 1,000 MWe BWRs operating with presently operating air ejector off-gas systems
(30 Minute delay pipe and 100 m stack)
steam may virtually eliminate noble gases and radioiodlnes from the turbine gland seal; appropriate
primary gas holdup, delay, or OF for each class must be considered for each combination.
-------
RADIOIODINE DISCHARGE CONTROL OPTIONS
Pressurized Water Reactors (PWRs)
Radioactive isotopes of iodine are generated inside the fuel
rods of pressurized water reactors. These volatile isotopes escape
through the fuel rod cladding defects and enter the primary coolant
system. Due to the leakages of primary coolant and/or various plant
operations, radioiodines may be released to the environment. These
release pathways may be broken down as follows:
(1) Primary Gases
Shutdown Degas if icat ion
Shim Bleed
(2) Secondary System Gases
Air Ejector Exhaust
Gland Seal Exhaust
Steam Generator Slowdown Tank Vent
(3) Building Ventilation
Containment Purge
Auxiliary Building
Turbine Building
83
-------
In order to determine the effectiveness of various PWR radiolodlne
control options, the source term for each, of the release pathways detailed
above was calculated and is shown in Table 27. However, radioiodine
evolving from the primary gases is not included since the noble gas
treatment options currently planned for PWR primary gases (60-day holdup,
cryogenic distillation, selective absorption, or cover gas recycle)
should effectively minimize this pathway of radioiodine release.
Assumptions for calculating the sources of radioiodine release are shown
in Table 12. Because of the uncertainty In the chemical form of
radioiodine released, two cases are chosen for consideration, namely,
that all radioiodine released is either in elemental or organic form.
Aside from the uncertainty in chemical form, uncertainty also
exists relative to the decontamination factor (DF) achieved in practice
by charcoal adsorbers. Existing test data would indicate a DF of
charcoal adsorbers for elemental radioiodine on the order of 100 - 10,000
and for methyl (organic) iodide, a DF of 4-1000 (50). As these tests
were generally performed under controlled laboratory conditions, these
DFs may not be representative of the conditions to be experienced by
charcoal adsorbers in the various types of reactor gaseous effluent
streams. Many factors (such as chemical form of Iodine, relative
humidity, atmospheric contaminants, leak-tightness of adsorber assembly,
etc.) may combine to degrade the DFs reported above. A comprehensive
investigation of the effectiveness of charcoal for removing radioiodine
has been recommended C49). As a result, DFs for charcoal adsorbers
84
-------
used on reactor plant gaseous effluents has been taken to be 10.
A DF of 100 is used for deep bed charcoal adsorbers. The use of an
internal recirculation charcoal adsorber (commonly referred to as a
"kidney") in a PWR containment, through which a fractional volume of
the building air is passed per unit time, decreases the concentration
of iodine-131 and iodine-133 in the containment atmosphere by factors
of 3 and 7, respectively. This is based on an 8,000 cfm flow rate for
16 hours of cleanup of 70% of the containment atmosphere after a 90-day
buildup. Routing of the steam generator blowdown tank vent to the main
condenser effectively minimizes this pathway of release at a PWR, partly
because of the high partition factor obtained in the condenser. The use
of clean steam on the turbine gland seal effectively eliminates this
pathway of release and has been proposed for BWRs only. In summary,
the following discharge control options were considered for a PWR:
(1) Steam Generator Blowdown Vent to Main Condenser
(2) Charcoal Kidney Adsorber inside Containment
(3) Steam Jet Air Ejector Charcoal Adsdrber
(4) Auxiliary Building Charcoal Adsorber
(5) Auxiliary Building Deep Bed Charcoal Adsorber
(6) Clean Steam: Valves > 2.5" diameter in Turbine Building
(7) Clean Steam on Turbine Gland Seal
Tables 28 and 29 detail these treatment options, generally added in
order of increasing cost per curie of iodine-131 eliminated, estimated .
85
-------
00
a\
Table 27
PWR Radioiodine Source Term: 2 Units, 1,000 MWe Each (Ci/yr)
ELEMENTAL
ORGANIC
STEAM GENERATOR SLOWDOWN TANK VENT
STEAM JET AIR EJECTOR
GLAND SEAL EFFLUENT
CONTAINMENT
AUXILIARY BUILDING
TURBINE BUILDING
PRIMARY GASES*
TOTALS
1-131
0.56
0.10
0.0002
0.70
0.11
0.04
1-133
0.34
0.062
0.0001
0.068
0.13
0.024
1-131
11.0
10.0
0.0002
7.0
13.0
0.04
1-133
6.6
6.2
0.0001
0.68
16.0
0.024
1.51
0.0031
0.624
3.0 x 10"5
41.0
0.31
29.5
0.003
aSource term given only for comparative purposes; present and future treatment systems
will reduce this source to negligible levels (i.e., less than 0.001 Ci/yr).
-------
Table 28
Annual Costs for Radioiodine (Elemental) Removal From PWR Gaseous Effluents
(2 Units, 1000 MWe each)
Upgrade to Deep Bed
Charcoal Adsorber:
Auxiliary Buildingb PGIE-6
Clean Steam: Valves
>2.5" Diameter PGIE-7
Gland Seal Clean Steam PGIE-8
Radioiodine Release
Control Option Added
None3,
Containment Kidney
Steam Generator Slowdown
vented to Condenser
Auxiliary Building
Charcoal Adsorber0
Air Ejector Charcoal
Adsorber
System
Designation
PGIE-1
PGIE-2
PGIE-3
PGIE-4
PGIE-5
Capital Cost
(Cumulative)
$
$
$
$2
$3
0
700,000
950,000
,950,000
,350,000
Annual Cost
(Cumulative)
$
$
$
$
$
0
120,000
160,000
460,000
560,000
(Ci/vr)
1-131
1.510
1.044
0.484
0.175
0.085
1-133
0.624
0.566
0.226
0.100
0.044
Total
2.134
1.610
0.710
0.275
0.129
$4,130,000 $ 860,000 0.054 0.032 0.086
$5,930,000 $1,160,000 0.022 0.013 0.035
$6,530,000 $1,360,000 0.021 0.012 0.033
a Does not include radiolodlne from primary system gases (shutdown degasification, shim bleed) as these
are effectively removed by gaseous waste treatment systems.
k Containment purge is also routed through this adsorber.
00
-------
oo
CO
Table 29
Annual Costs for Radloiodine (Organic) Removal From PWR Gaseous Effluents
(2 Units, 1000 MWe each)
Estimated
Estimated
Radioiodine Release
Control Option Added
None3
Containment Kidney
Steam Generator Slowdown
vented to Condenser
Mr Ejector Charcoal
Adsorber
Auxiliary Building Charcoal
Adsorber^
Upgrade to Deep Bed Char-
coal Adsorber: Auxiliary
Building^
Clean Steam: Valves
>2.5" Diameter
Gland Seal Clean Steam
System
Designation
PGIO-1
PGIO-2
PGIO-3
PGIO-4
PGIO-5
PGIO-6
PGIO-7
PGIO-8
Capital Cost
(Cumulative)
$
$
$
700
950
0
,000
,000
$1,530,000
$3,530,000
$4,
§5,
$6,
130,
930,
530,
000
000
000
Annual Cost
(Cumulative)
$
$
$
$
$
$
$1
$1
0
120,000
160,000
260,000
560,000
860,000
,160,000
,360,000
(Ci/vr^
1-131
41
36
25
16
2
1,
.040
.374
.374
.374
.574
.194
1.162
1.161
1-133
29
28
22
16
2
0
.504
.921
.321
.741
.254
.805
0.786
0.786
Total
70.544
65.295
47.695
33.115
4.828
1.999
1.948
1.947
aDoes not include radioiodine from primary system gases (shutdown degasification, shim bleed)
as these are effectively removed by gaseous waste treatment systems.
Containment purge is also routed through this adsorber.
-------
costs (1, 16, 24, 25), and estimated releases from a two-unit (1000 MWe
each) PWR power station. In any case, the uncertainty associated
with the costs of progressively improved treatment increases tremendously
beyond the first three or four equipment additions.
Whereas many PWRs formerly included only a charcoal adsorber on
the primary gas decay tank dishcarge and a containment kidney adsorber,
present design typically includes these as well as venting the steam
generator blowdown tank to the main condenser and auxiliary building
charcoal adsorbers. At least one PWR has planned to treat the air
ejector effluent via charcoal delay beds (35) but this is as much meant
to reduce noble gas releases as radioiodine releases. Finally, an
overall reduction by perhaps a factor of 2 can be obtained by those PWRs
using the cover gas recycle system for the control of primary gas
radioactivity release (1, 47).
Boiling Water Reactors (BWRs)
Radioactive isotopes of iodine are also generated Inside the
fuel rods of boiling water reactors and may escape to the reactor
coolant system through defects in the fuel rod cladding. Due to
leakages of reactor coolant, and/or various plant operations, radio-
iodines may be released to the environment. These release pathways
be broken down as follows:
(1) Primary Gases
Condenser Air Ejector
Turbine Gland Seal
89
-------
(2) Building Ventilation
Turbine Building
Reactor Building
Radwaste Building
The source term for each, of the release pathways detailed above
was calculated and is shown in TaBle 30. However, the noble gas
treatment options currently planned for the BWR air ejector source
term should effectively minimize this release pathway. Assumptions
for calculating the sources of radioiodine release are shown in Table 11.
Because of the uncertainty in the chemical form of radioiodine released,
two cases are chosen for consideration, namely, that all radioiodine
released is either in elemental or organic form.
Six radioiodine discharge control options were considered
for BWRs:
(1) Clean Steam: Valves>2.5" diameter (Turbine Building)
(2) Turbine Building Charcoal Adsorber
(3) Turbine Building Deep Bed Charcoal Adsorber
(4) Reactor Building Charcoal Adsorber
(5) Radwaste Building Charcoal Adsorber
(6) Turbine Gland Seal Clean Steam
Due to the uncertainty associated with, charcoal adsorber DFs, as
previously discussed, a value of 10 Has Been used. For a deep bed
charcoal adsorber, an incremental DF of 10 Is used. The use of clean
90
-------
steam on the gland seal effectively eliminates this pathway of release.
A DF of 5 is more or less^ presumed for the use of clean steam on valves
greater than 2.5 inches in diameter in the turbine building. It should
be noted that no credit is taken for the use of the standby gas treatment
system in decontaminating the reactor building releases.
Tables 31 and 32 detail these treatment options, considered as
individual successive equipment additions (in order of increasing cost
per curie of iodine-131 eliminated), estimated costs (1, 24, 25), and
estimated releases from a two-unit (1000 MWe each) BWR power station.
Uncertainty associated with the cost of systems increases rapidly with
the addition of equipment.
Whereas BWRs typically included no treatment for radioiodine
releases, BWRs of current design are incorporating combinations of
such features as supplying clean steam to valves in the turbine
building, a deep bed turbine building charcoal adsorber, and a' turbine
gland seal clean steam system.
91
-------
VO
to
Table 30
BWR Radioiodine Source Term: 2 Units, 1,000 MWe Each (Ci/yr)
GLAND SEAL
ELEMENTAL
1-131 1-133
0.0058 0.0330
ORGANIC
1-131 1-133
0.0058 0.0330
REACTOR BUILDING
RADWASTE BUILDING
TURBINE BUILDING
PRIMARY GASESa
TOTALS
0.0170
0.0014
1.0000
1.024
29.0
0.0990
0.0040
5.7000
5.836
17.2000 99.0000
0.1200 0.6800
1.0000 5.7000
28.326 105.413
170.
~150.
•850.
^Source term given only for comparative purposes; present and future treatment systems
will reduce this source to negligible levels (i.e., less than 0.001 Ci/yr)
-------
Table 31
Annual Costs for Radioiodine (Elemental) Removal From BWR Gaseous Effluents
(2 units, 1000 MWe each)
Estimated
Estimated
Radioiodine Release
Control Option Added
Presently Operating3
Noneb
Clean Steam: Valves
> 2.5" Diameter
Turbine Building
Charcoal Adsorber
Upgrade to Deep Bed
Charcoal Adsorber:
Turbine Building
Reactor Building
Charcoal Adsorber
Radwaste Building
Charcoal Adsorber
System
Designation
BGIE-1A
BGIE-1
BGIE-2
BGIE-3
BGIE-4
BGIE-5
BGIE-6
Capital Cost
(Cumulative)
$
$
$1
$4
$5
$7
$7
0
0
,800,000
,300,000
,100,000
,100,000
,350,000
Annual Cost
(Cumulative)
$
$
$
$
$1
$1
$1
0
0
300,000
750,000
,200,000
,600,000
,640,000
(Ci/vr)
1-131
(30
1
0
0
0
0
0
.0)
.02
.224
.044
.026
.017
.010
1-133
(176.)
5
1
0
0
0
0
.84
.28
.250
.147
.058
.055
Total
(206
6.
1.
0.
0.
0.
0.
.)
86
50
295
173
075
065
Turbine Gland Seal
Clean Steam
BGIE-7
$7,950,000
$1,840,000
0.004 0.022 0.026
Illustrates projected effects of two 1,000 MWe BWRs operating with presently used off-gas system
(i.e., 30 minute delay and 100 m stack for air ejector noble gases).
^Reflects source term for sources other than air ejector as "augmented" BWR noble gas treatment systems
5 (charcoal adsorption, selective absorption, or cryogenic distillation) will effectively remove air
ejector radioiodines.
-------
VO
Table 32
Annual Costs for Radioiodine (Organic) Removal From BWR Gaseous Effluents
(2 Units, 1000 MWe each)
Control Option Added
System
Designation
Estimated Estimated Radioiodine Release
Capital Cost Annual Cost . (Ci/yr)
(Cumulative) (Cumulative) 1-131 1-133 Total
$
$
0
0
$
$
0
0
$2,000,000 $ 400,000
(178.) (955.) (1133.)
135
28.3 105.
2.85 16.3
19.2
Presently Operating3 BGIO-1A
Noneb BGIO-1
Reactor Building
Charcoal Adsorber BGIO-2
Upgrade to Deep Bed
Charcoal Adsorber:
Reactor Building BGIO-3
Radwaste Building
Charcoal Adsorber BGIO-4
Clean Steam: Valves
>2.5" Diameter BGIO-5
Turbine Building
Charcoal Adsorber BG10-6
Turbine Gland Seal
Clean Steam BGIO-7
Illustrates projected effects of two 1,000 MWe BWRs operating with presently used off-gas system
(i.e., 30 minute delay and 100 m stack for air ejector noble gases).
Reflects source term for sources other than air ejector as "augmented" BWR noble gas treatment systems
(charcoal adsorption, selective absorption, or cryogenic distillation) will effectively remove air ejector
radioiodines.
$2,500,000
$2,750,000
$4,550,000
$7,050,000
$7,650,00
800,000
840,000
$1,140,000
$1,590,000
$1,790,000
1.30
1.19
0.390
0.210
0.204
7.40
6.79
2.23
1.21
1.17
8.70
7.98
2.62
1.42
1.37
-------
DETERMINATION OF POPULATION RADIATION EXPOSURE
The estimation of potential health risks associated with radio-
activity releases from nuclear power reactors requires an assessment
of the radiation exposure resulting from these releases. This dose
assessment is a difficult and complex task. The complexity of the
dose assessment results from: (a) the number of different radio-
nuclides produced and released from the nuclear reactor (there are
at least 100 major radionuclides and over 300 radionuclides of lesser
significance); (b) the multiplicity of release paths from the facility;
(c) the number of environmental vectors which can convey the radio-
nuclides to man; and (d) the number of body organs which may be
irradiated by a given radionuclide.
Detailed studies of radionuclide effluents, exposure pathways,
i
and radiation doses have indicated that this complex situation can
be simplified by consideration of the most critical pathways and
principal radionuclides which contribute significantly to the radiation
dose. Both calculational studies (53) and environmental measurements
(9,51) have indicated that the principal radionuclides which contribute
to radiation exposure from nuclear reactor effluents can be reduced to
approximately two dozen in number which interact via the exposure path-
ways shown in Table 33.
The radioiodines (principally 1-131 and 1-133) are of importance
because of the relatively large yield in uranium fission and the high
affinity of the thyroid gland for iodine. The major exposure pathways
95
-------
for radioiodine are air inhalation, ingestion of drinking water, fresh
milk, beef, and lamb (53).
Cesium isotopes (Cs-134 and Cs-137) are also produced in significant
quantities and contribute to the radiation dose received by the total
body, bone, liver, and gastrointestinal (61) tract. The principal
exposure pathways involved are drinking water and consumption of fish
and shellfish.
The noble gases krypton and xenon have many radioactive isotopes
which are formed in fission (see Table 5). These inert gases are
important because of their fission yields and half-lives. The only
important source of exposure is external whole body irradiation by
the gamma-emitting radionuclides in a cloud and the submersion skin
dose from beta emitters.
96
-------
Table 33
Principal Exposure Pathways for Radiation Exposure
from Nuclear Reactor Effluents
Radionuclide Discharge
Mode
Principal Exposure
Pathways
Critical
Organ
Radioiodine Airborne Ground deposition - external irradiation
Air inhalation
Grass-cow-milk
Leafy vegetables
Water Drinking water
Fish consumption
Shellfish
Whole body
Thyroid gland
it
it
Tritium
Noble Gases
Airborne Air inhalation and transpiration
Submersion
Water Drinking water
Food consumption
Airborne External Irradiation
Whole body
Skin
Whole body
it ii
Whole body and
Skin
Cesium
VC
Transition
metals (Fe,
Co,Ni,Zn,Mn)
Airborne Ground deposition - external irradiation
Grass-cow-milk
Grass-meat
Inhalation
Water Sediments - external irradiation
Drinking water
Fish consumption
Water Drinking water
Shellfish consumption
Fish consumption
Whole body
ii it
G.I. Tract
ii
ii
Direct Radiation
External irradiation
Whole body
-------
Estimation of Radiation.Doses from LiguicLEffluent
There are two principal pathways for radionuclides released as
liquid effluents to reach man: ingestion of drinking water and
ingestion of aquatic or marine foods (principally fish and shell-
fish). Other exposure pathways such as submersion (swimming), use
of water for irrigation, boating, etc. are generally less significant
and were not considered.
The radiation dose [equivalent] rate delivered by a given radio-
nuclide which is ingested via water or food consumption can be calcu-
lated from the following expression:
DE = GP (DICF)Q
where DE is the dose [equivalent] rate in mrem/year,
P is a pathway transfer factor relating human intake to the
radionuclide concentraton in water (pCi/year per pCi/liter),
DICF is the dose [equivalent] rate delivered per unit intake (mrem/year
per pCi/year), and
Q is the annual release rate in curies per year.
The constant G is related to the dilution afforded by the condenser
cooling water flow, V, and the conversion factor from curies to pico-
curies:
1012 pCi/Ci
b =
V liters/year
The cooling water volume, V, is calculated for a 1 GWe plant operating
for 0.8 years at a flow rate of 800,000 gallons per minute for once-
98
-------
through cooling:
° 1-27 X
-8 fe)
.2642 (gallons/liter) ;
and at 22,000 gallons per minute for the blowdown from an augmented cooling
system (cooling towers):
V = °'8 (yr) I2'2 X ]°4 STFSte5) (5'256 x 1p5 ^JtJr) = 3.47 x 1010
liter)
For once-through-cool tng 6 has the value of 0.785 and it is 28.8 for
plants with cooling towers.
The pathway transfer factor, P, for the water ingestion pathway
is defined by:
where Iw is the annual drinking water consumption rate in liters per
year, F is a factor to correct for removal of radionuclides by
conventional purification process at water intakes, DF is a factor
to account for dilution between the effluent discharge canal and the
water intake, A is the radiological decay constant of the radionuclide,
and TW is the time interval between discharge and consumption of water.
The removal factor, F, is given for various radionuclides in Table 34.
The dilution factor permits a factor of two reduction in concentration
prior to consumption of water by an individual, a factor of one hundred
reduction prior to reaching the water supp.ly for large population groups
for the lake site, and a factor of twenty reduction for the river site.
99
-------
The pathway transfer factor, P, for ingestion of fish and shellfish
has the following derivation:
where:
Ip is the food ingestion rate in kilograms/year,
R is a preparation loss factor to account for removal of radio-
nuclides during food cleaning and cooking,
DF is the dilution factor between the effluent discharge point
and the fish/shellfish,
CF is a concentration factor which relates uptake by the organism
to the concentration in water (pCi /kilogram per pCi/Hter = liter/kilo-
gram), and
Tf is the time between effluent release and consumption of fish.
The concentration factors for fish and shellfish (crabs, lobsters, clams,
oysters* etc.) vary for different radionuclides. They are also somewhat
dependent on the concentrations of chemically similar stable elements in
.'v ,
the water. Representative values are presented for marine (seacoast site)
and freshwater (river and lakesites) species in Tablt 34. The values for
the dilution factors and intake rates are presented in Table 35 along
with the values for the other parameters.
The dose [equivalent]-intake conversion factor, DICF, is given by:
100
-------
where:
f is the fraction of the intake which reaches the critical
organ from inhalation (fa) or ingestion (fw),
[F(QF)] is the product of the effective energy per disintegration
(MeV/disintegration) and the quality factor of the emitted
radiation (QF) [The quality factor is the conversion between
the dose equivalent (in rem) and the absorbed dose (in rad)
and, consequently has units of rem/rad. For beta (B) and
gamma (Y) radiation, QF = 1.0 and for alpha radiation,
QF = 10.],
m is the mass of the critical organ (grams),
t is the duration of the exposure (days), and
TE 1s the effective elimination half-time (days) for the
biological elimination from the critical organ and loss
by radioactive decay.
The constant k has the value 0.074 (gram-rad-disintegrations per MeV-
pCi-day). It is obtained as follows:
k = 1.443 (3.7 x 10'2 disintegrations/pCi-second) x (8.64 x 104 seconds/day)
x(103 mrem per rem) x (1.602 x 10'6 erg/MeV) r (100 ergs/Gram-rad);
the 1.443 factor is the inverse of the natural logarithm of 2. Values
for m, TE> f[fa or fw] and [f(QF)] were taken from reference (65) with
two exceptions: the iodine values (Table 38) were computed for the
individuals in four age groups based on the parameters shown in Table 37
t
and the DICF for tritium was computed using a quality factor of 1.0
instead of the 1.7 value used in (65).
101
-------
o
to
Table 34
Radionuclide Dependent Factors for Liquid Effluent Dose Calculations
Radionuclide
Tritium (H-3)
Mn-54
Fe-55
Fe-59
CO-58
Co-60
Sr-89
Sr-90
Y-91
Zr-95
Nb-95
Mo-99
Ru-103
Ru-106
1-131
1-133
Cs-134
Cs-137
Ce-141
Ce-144
Pr-144
Fraction Remaining
After Treatment at
Water Intake (F)(54)
1.0
0.2
.2
.2
.2
.2
.2
.2
.2
.3
.3
.8
.2
.2
.8
.8
.8
.8
.2
.2
.2
Concentration Factors (CF
Fish
1.0
3,000
1,000
1.000
100
100
1
1
30
30
100
10
3
100
500
200
30
30
30
30
30
Marine (79, 80)
Shellfish
1.0
5,000
20,000
20,000
1,000
1,000
6
6
1,000
1,000
100
10
2,000
2,000
50
50
20
20
1,000
1 ,000
1,000
in pCi/kg per
Fish
1.
25
300
300
500
500
40
40
100
100
30,000
100
100
5,000
1
1
1,000
1 ,000
100
100
100
pCi/liter)
Freshwater(SO)
Shellfish
0 1.0
40,000
3,200
3,200
1,500
1,500
700
700
1,000
1,000
100
100
2,000
2,000
25
25
1,000
1,000
1,000
1,000
1,000
-------
Table 35
Parameters Used for the Calculation of the Radiation
Dose from Liquid Effluents
Site Parameters
Seacoast River Lake Reference
Dilution Factor to Receptor (DF)
Critical Individual
Population Average
0.5
.01
0.5
.05
0.5
.01
Intake Rates (I,kilograms/year)
Critical Individual
Fish (fresh)
Shellfish (fresh)
Water (liters/year)
Population Average
Fish (fresh)
Fish (processed)
Shellfish
Water (liters/year)
18
IB, ,
—(a)
3.9
2.8
]»8t
— U)
18
440
5.1
2.6
1.2
365
18
440
2.0
2.9
0.5
365
(1)
(77)
(a) The fresh water supply at the seacoast site is not affected by plant
effluents.
Other
Reference
Dilution Flow (V liters/year)
Once-through cooling
Cooling tower blowdown
Preparation Loss Factor for Seafood (F)
Time Between Discharge and Consumption
(T, hours)
Critical Individual
Water
Fish
Shellfish
Average Individual in Population
Water
Fish (fresh)
Fish (processed)
Shellfish
1.27 x 1012
3.47 x 1010
0.8
24
24
24
36
36
30 days
30 days
(54)
103
-------
Evaluation of External Whole Body Doses from Gaseous Effluents
Radiation doses from airborne effluents were calculated using
the AIREM computer code (81). This code provides for a Gaussian or
bell-shaped concentration profile in the vertical direction and a
uniform concentration distribution in the horizontal cross wind
direction. The vertical diffusion is limited to a finite mixing
height (82) and the technique of image sources is used to account
for reflection from both the mixing layer and ground surface.
The basic diffusion equation used in AIREM is a standard sector-
averaged equation (83,84) modified to include radionuclide decay by
time of flight:
(A
g,
where:
X = ground level airborne concentration in Ci/m3,
Y = time integrated ground level concentration-exposure
in Ci-sec/m3,
Q1 = source release rate in Ci/sec,
Q = time integrated release in Ci (i.e., total release),
f = fractional wind frequency in a sector,
r = distance from the stack in meters,
h = effective stack height in meters,
n = number of sectors,
104
-------
2irr/n = sector width at distance r In meters,
az = standard deviation of the vertical distribution of an
assumed gaussian cloud, in meters,
u = average wind speed in the sector in m/s,
A = decay constant of radionuclide in sec'1.
and * = transit time from the stack to distance r, in seconds
(t=r/u).
This equation is solved repeatedly for each radionuclide and stability
class within each sector for all downwind distances of interest.
The preceding equation provides the ground level air concentration
at a distance r from the release point. This concentration is then
used to calculate the radiation dose from inhalation and transpiration
(tritium), and the deposited activity on the ground surface which
contributes to external whole body exposures and to food intake path-
ways. The inhalation and transpiration doses are computed directly
from the ground level air concentration by the following alogrithm:
D = * Q . (DCF),
Q1
Where D is the dose rate in mrem/year, (x/Q1) is the atmospheric
dispersion factor as computed above (sec/m3), Q is the annual release
rate (curies/year), and DCF is an appropriate dose conversion factor
(mrem/year per Ci-sec/m3) for the radionuclide and exposure mode of
interest. The inhalation dose conversion factors for radioiodine will
be provided in a subsequent section, for all other radlonuclides values
from (54) were used.
105
-------
The activity deposited on the ground surface was computed from
the ground level air concentration as follows:
w == x Q ud
Q1
Where w> is the deposition rate (Ci/m^ sec),-^ is the deposition
velocity (m/sec), and the other quantities are as defined above.
The deposition velocity is an empirical factor which is defined as:
d xt
Where V is the accumulated deposit (Ci/m^) and xt is the integrated
air concentration over the period of measurement. The airborn6
concentration is depleted uniformly by the deposited activity using
the continuity principal.
The accumulated deposit is given by:
u) _
Where w is the deposition rate as given above, Pis the deposited
activity (Ci/m^), xe is an effective removal rate, and t is the time
interval. For deposition onto foliage which leads to an ingestion dose,
the effective removal rate is defined as:
e 12 ,
where xe is the effective removal rate constant (days"1), X is the
radiological decay constant (days'1), and the remaining term accounts
for physical removal by wash-off, wind, and plant growth of the
106
-------
radionuclide from plant surfaces. This latter process is assumed to
have a half-time of 12 days (54). Computation of the external whole
body dose from deposited radionuclides is performed assuming a uniform
semi-infinite plane source by:
D= V (DCF') ,
where D is the annual dose rate (mrem/year), V is the accumulated
activity deposit (Ci/m2) and DCF1 is the dose conversion factor for
a semi-infinite plane source (mrem/year per Ci/m2). Values for DCF*
are taken from (54).
The gamma dose rate at the surface of a receptor at a point x ,
yr, zr in space from single energy gamma photons emitted from an
elemental volume located at point x,y,z of the radionuclide bearing
cloud is (84):
d JDR (xr,yr,zJJ =K -fif- x(x,y,z) A -p- exp f-yaR^ B dv
Where:
DR = dose rate
E = photon energy
A = gamma photon abundance (photons/disintegration)
B = buildup factor
ya = linear air attenuation coefficient - m"1
— = mass energy absorption coefficient of muscle - cm2/gm
dv = elemental volume - m3
x(xty»z) = airborne concentration at point x.y.z
- (x/Q')Q
107
-------
K = dimensional constant
R = distance between emitting volume and receptor point
222
where R= (X'Xr) + (y"yr) +(z"zr)
The x(*»y>z) is computed from the previous relationship for (x/Q1)
except that the vertical concentration profile is considered by
substituting (h-z)2 for h2 for each height z.
Integration of this equation over all space will yield the dose
rate at the receptor point due to the gamma emitters in the entire
plume. Solution of this equation yields the dose rate from mono-
energetic gamma photons at a single point on the ground surface and
for a single invariant wind speed, wind direction and dispersion
regime. The total dose rate from the entire plume material is found
by summation over all meteorological conditions and gamma energies
with appropriate weighting factors for frequency of occurrence. This
integration is performed by R.E. Cooper's EGAD code (56). The EGAD
code considers ground and inversion layer reflections and employs an
empirical third-order polynomial expression for the buildup factor B.
These relationships are more accurate than the simple one-term
uncollided flux approximation, B = 1 + yeR.
Risk calculations for external whole body photon exposure are
based on the average dose to the body allowing for both self shielding
and buildup. The dose calculation is performed in two parts. First,
the maximum dose to a differential volume of tissue is calculated using
appropriate attenuation and buildup factors for air as indicated above,
108
-------
then the average dose to the body is calculated by means of the dose
reciprocity theorem (85). Strictly speaking, this implies that the
ratio of the average to maximum dose from a finite cloud is the same
as the average to maximum dose from a semi-infinite cloud. This is a
good assumption for the cases of interest here since in either case,
we are dealing with isotopic angular flux distribution ( in 2n geometry)
and a quasi-equilibrium distribution of photon energies.
To determine the ratio of average to maximum dose from a semi-
infinite cloud, we have used an updated version of Adam's (86) solution
to this problem as updated by Russell and Galpin (87) to take advantage
of the exact photon scattering calculations published by the MIRD
committee (70,71). The latter results are tabulated in terms of
absorbed fractions 0 as a function of photon energy and body mass, a
70 kg mass being used for these calculations.
Population-integrated doses (person-rem) are computed by the
following expression:
DP - \ 1 Pj (D,., * n,} ,
where Dp is the total population-integrated dose, P- is the enclosed
population within sector j and Dj_-| and Dj are the annual individual
dose rates for the exposure mode of interest at the inner and outer
radial boundaries of sector j.
109
-------
Radioiodine Thyroid Dose Computations
The control of airborne radioiodine releases is complicated by
the diversity of chemical forms which may coexist under certain
conditions. Until recently, it has been assumed that the principal
chemical form present in reactor effluents was elemental iodine
vapor (Ip) and existing control technology was predicated on this
assumption. Recent preliminary unpublished studies conducted by the
Atomic Energy Commission have indicated that, at several facilities,
the predominant chemical form was not elemental iodine but rather a
volatile organic iodide, principally methyl iodide, CH^I. This
apparent change in chemical form may not only affect the efficacy of
control techniques and physiochemical transfer characteristics and,
consequently, the magnitude of the discharge but also may significantly
affect the critical exposure pathway leading to man. Airborne radio-
iodine discharges can result in radiation exposure to man by four
principal pathways: air inhalation, milk consumption, ingestion of
leafy vegetables and other produce, and external whole-body exposure
from activity deposited on the ground. Except for air inhalation, the
remaining pathways depend upon the transfer coefficient between air
and vegetation on the ground. This transfer coefficient is termed
the "deposition velocity." For iodine in the elemental form, the
deposition velocity generally ranges between 0.002 and about 0.1 m/sec,
110
-------
depending upon the fraction absorbed on airborne participates.
Generally, deposition velocities in the range of 0.005 to 0.015 m/sec
are considered typical for reactor effluents (60, 61). The corres-
ponding deposition velocity for the organic form, methyl iodide, has
not been well characterized, but has been shown to be several orders
of magnitude smaller than for the elemental form (61-62). This
results in negligible deposition of methyl iodide onto vegetation
(grass and leafy vegetables) and the ground surface. Because of this,
the principal exposure pathway for methyl iodide releases is likely
to be direct inhalation rather than milk ingestion. For the elemental
form, milk ingestion is likely to be the controlling pathway for
iodine-131 if there are grazing animals (dairy cattle or goats) in
the vicinity of the plant. This results from the ability of the cow
(or goat) to concentrate the radioiodine. by virtue of the large area
of grass grazed daily (20-80 m2/day). The exact ratio of the thyroid
dose from milk ingestion to air inhalation depends upon parameters
(primarily the mass of the thyroid gland, the ventilation [breathing]
rate, and the average milk consumption) associated with the age of the
individual involved. The critical; receptor is usually taken to be a
young child because of a smaller thyroid mass.and a greater daily milk
consumption than for other age groups. These parameters are presented
in Table 38 for the selected age groups.
The thyroid population doses were computed for four age groups in
order to account for varying thyroid mass, milk consumption, and
radiation sensitivity with age. The age groups used were first year infam
111
-------
(6-month old typical individual). 1-9 years (4-year old typical
individual), 10-19 years (14-year old typical individual), and
over 20 years (20-year-old typical adult). Three intake modes
were considered: air inhalation, vegetation consumption, and milk
ingestion. The external whole-body dose from deposited radioiodines ,
was also computed, but this exposure mode was negligible compared
to the inhalation and ingestion pathways. In analyzing the radiation
exposure from organic iodides, a deposition factor of 1/1000 that of
elemental iodine was used. This is an arbitrary value, as the
published values (61, 62) show considerable variation. Although
the chosen value is somewhat higher than the best available estimate
(62), it is felt that the use of the higher value is justified in
view of the uncertainty in the chemical form and the possibility of
a change in chemical form due to radiolytic or photodissociation of
the methyl iodide in the environment or its attraction to airborne
aerosols. Both of these processes could drastically affect the
chemical form and, hence, the deposition velocity.
The relationship between the dose [equivalent] rate delivered
to the critical organ by a radionuclide and the ambient air concentra-
tion of that radionuclide can be expressed as:
DE = X(DICF)P ,
where DE is the dose [equivalent] rate in mllHrem per year,
X is the air concentration of the radionuclide (pCi/m3),
P is the pathway transfer factor (pCi/year per pCi/m3) relating
the intake rate to the ambient air concentration,
112
-------
and DICF is the dose [equivalent] rate delivered per unit intake rate
(mrem/year per pCi/year).
The air concentration, x> is determined by the dispersion models dis-
cussed in the preceding section. The pathway transfer factor, P,
which relates the intake rate by an individual to the ambient air
concentration, will be discussed separately for each pathway in
subsequent sections.
The dose [equivalent] -intake conversion factor, DICF, has the
same relationship as given in the previous section dealing with water
and fish ingestion doses. The parameters for the radioiodines which
determine the DICF for ingestion and inhalation are shown in Table 37.
It is assumed that the fraction of radioiodine reaching the thyroid
gland for milk and vegetation ingestion is identical to that for
water ingestion (fw). The resulting values for DICF for intake by
ingestion and inhalation are presented in Table 38 as a function of
the age of the receptor.
The Evaluation of Thyroid Dose from Radioiodine Inhalation
The evaluation of the thyroid dose from inhalation is relatively
simple as the pathway transfer- coefficient, P, is just the breathing
rate in m3/year. These values are shown for the four age groups in
Table 39 and the products of P and the dose [equivalent]-1ntake
conversion factors, DICF, are presented in Table 40. The metabolic
properties of methyl iodide and the other organic iodine species.were
assumed to be.the sameas for the elemental forms. The critical
113
-------
Individual doses were calculated assuming no depletion in the air concen-
tration by deposition prior to inhalation. The population dose calculations,
however, did account for loss by deposition prior to inhalation.
Table 36
Population Groups Used for
Radioiodine Dose and Risk Evaluations
Age Grouo A9e of "Typical" Percent of Total
Individual Population Within Age Group
< 1 year old 6 months 1.79%
1 - 9 years old 4 years 16.47%
10-19 years old 14 years 19.57%
> 20 years standard man 62.17%
100.00%
Source: National Center for Health Statistics, DHEW.
114
-------
Table 37
Thyroid Dose Parameters
Age Group (years)
Age of Typical Individual (years)
Fraction of Iodine Reaching Thyroid
Ingestion (fw)
Inhalation (fa)
Biological Half-Time (days)
Effective Decay Energy (F, MeV)
1-131
1-133
Thyroid Mass (m, grams)
0-1
0.5
.50<64>
.38?
20b
.18(67,68)
.40°
!.9(66)
1 - 10
4
.50<54>
.38*
29b
.18(67,68)
.40C
2.7(66)
10 - 20
14
.37(64)
.28*
70*
.19(67,68)
.42C
12(66)
> 20
standard man
.30<65)
.23(65)
lOO6
.19(67,68)
.42C
20(66)
?fa was computed as 75 percent of fw as recommended in table 10, page 53 of reference (65).
Upper range values, especially for younger ages, given in reference (66).
Computed using decay scheme presented in (69) and absorbed fractions from references (70)
and (71).
-------
Table 38
Radioiodine Dose Conversion
Factors Per Unit Activity Intake
Inhalation
Age Group
(years)
< 1
1 - 9
10-19
> 20
Ingestion
< 1
1 -9
10-19
> 20
Age of "Typical"
Individual
(years)
0.5
4
14
20
0.5
4
14
20
DICF (mrem/pCi inhaled)
131j 133
1.5xlO"2 4.8xlO"3
1.2xlO~2 3.4xlO"3
2.4xlO"3 S.OxlO'4
1. 2xlO-3 S.OxlO"4
DICF (mrem/pCi ingested)
2.0xlO'2 6.3xlO-3
1.6xlO"2 4.5xlO"3
3,lxlO~3 8.0X10'4
1.6xlO"3 3.9xlO-4
116
-------
PATHWAY TRANSFER COEFFICIENT
Table 39
PATHWAY: Inhalation
Note: The transfer coefficient for the inhalation pathway is the breathing rate in
cubic meters per year.
AGE GROUPING
6 month-old
4 year-old
14 year-old
Adult
BREATHING
/ 1.15 x
3.53 x
6.44 x
7.30 x
RATE [m3/yr]
103
103
103
io3
Table 40
PRODUCT OF THE TRANSFER COEFFICIENT WITH THE DOSE EQUIVALENT CONVERSION FACTOR
FOR THE INHALATION PATHWAY
i [mrem/yr per pCi/m3 in air]
AGE GROUPING
6 month-old
4 year-old
14 year-old
Adult
Maximum Individual
(4 year-old)
1-131
..,...!.?..
42
15
8.8
42
1-133
5.5.
12
3.9
2.2
• • • . _
12
117
-------
Evaluation of Thyroid Inqestion Dose
The dose contribution to the thyroid from radioiodine via the
air-grass-milk and air-vegetable pathway was evaluated for the three
representative sites described previously. The iodine concentrations
on vegetation were calculated from annual sector-averaged air
concentrations incorporating cloud depletion. The farms and dairy
cows were assumed to be uniformly distributed with respect to distance
and direction from the source and the average ground deposition in
each radial sector was weighted by the fraction of the total area
within 50 miles of the plant it comprised. Sufficient dairy and
produce farms were assumed to be located within a 50-mile radius of
a facility to provide the total milk and produce consumption for the
total enclosed population.
The assumptions of uniform location and sufficient milk and
produce production are not strictly correct when applied to a specific
reactor site. For the representative seacoast site chosen in this
study, the nearest dairy farm is approximately 12 miles from the site
and milk production within 50 miles is not sufficient to meet the needs
of the enclosed population at that distance. Thus, milk must be
imported into the region and the population thyroid dose from that
facility alone via the milk pathway would be greatly overestimated by
the calculation model employed. However, projections of the growth
of the nuclear power industry indicate that by the year 2000, a
majority of the U. S. population will reside within tens of miles
118
-------
of at least one nuclear power station and a similar situation will
undoubtedly exist with respect to dairy farms. Thus, in the future,
milk and other produce leaving the 50-mile locality will contribute
to the population dose at another location while foodstuffs entering
the region are likely to have been produced in the vicinity of
another nuclear power plant. For the purpose of calculating the
integrated population dose, it is, therefore, reasonable to assume
uniform contamination levels in foodstuffs (providing a sufficient
decay period for short-lived radionuclides is incorporated) and
that the population dose calculated on the basis of total consumption
and production within the 50-mile radius is comparable to that calcu-
lated allowing for food transfer between regions of production and
consumption.
The assumption of a uniform distribution of the farms within the
50-mile radius was compared with the existing farm locations for
several actual sites. These locations were found to vary randomly
between sites and generally the number of farms enclosed within
different radial distances was found to increase with the enclosed
land area.
The daily milk or produce intake rates and dose conversion factors
were calculated for each of the three selected age groups and weighted
according to the fraction of the total U. S. population in each of
these age categories as shown in Table 36. These age-weighted dose
conversion factors were used together with the area-weighted Iodine
concentrations and the site-related population densities to arrive
at an integrated population dose;
119
-------
The pathway transfer factors, P, for ingestion of milk and
vegetables were computed from the following relationship:
/£CVday_\ vd__RJFI r , .n
VpCi/m^y ' Aeff [exp(-At)] ,
where: v^ is the deposition velocity (m/sec),
R is the initial retention factor,
T is a transfer factor between the deposited activity and
the radionuclide concentration in food,
F is an intake modifying factor,
Aeff is the effective removal rate constant from vegetation
(seconds-!),
A is the radioactive decay constant (days)'1!
T is the interval between production (milk) or harvesting
(vegetables) and consumption (daysX and I is the intake
rate of milk or vegetables. The values and units of these
parameters together with more complete definitions of R,T,
and F for the two pathways are presented in Table 41.
The vd/Aeff term in this expression times the average air concentration,
x, gives the saturated [equilibrium] ground deposition per unit area.
The exponential term accounts.for loss by radioactive decay between
production and consumption.
The evaluated pathway transfer coefficients, P, for radioiodine-131
and radioiodine-133 are presented in Table 42 for the vegetable ingestion
pathway and in Table 43 for the milk ingestion pathway. In both cases,
these values are presented separately for each age group. The products
of the pathway transfer factors and the dose [equivalent] conversion
120
-------
Table 41
Parameters Used for the Calculation of Radioiodine Intakes
Parameter
Deposition Velocity (vd)
elemental iodine
organic iodides
Initial Retention Factor (R)
Removal Half-time Due to Loss by
Weathering and Growth
Transfer Factor (relating intake to
deposited activity)
Intake -Modifying Factor (F)
Decay Interval Between Production
and Consumption (2)
Average individual (population)
Critical individual
Intake Rates (I)
6-month-old
4-year -old
14-year-old
adult
; critical individual
Vegetabl e
Ingestion
meters/second
0.01 c
1 x 10~5
pCi/m2[plant]
0.25
days
12
kilograms/m2(T)
1 (78)
Fraction remaining
after washing & prepa-
ration
0.4
days
7
1
kilograms/year (54,77)
0
13
20
30
13 (54,77)
Milk
Ingestion
meters/ second
0.01 r
1 x 10'5
pCi/m2[deposited]
0.25
days
12
pCi/liter per pCi/m2(RT)
1-131 0.2
1-133 0.1
Fraction of year cows
graze on pasture
0.5
days
3
1
mi 11 iHters/day
500
700
660
230
1000 (77)
Reference
(76)
(see text)
(54)
(54)
(72)
(73)
(54)
(54)
(54)
(74)
-------
Table 42
PATHWAY TRANSFER COEFFICIENT fpCi/yr ingested per pCi/nr in air]
•''~'~*- ™'•""" •' • • ------ -i - i ..-_ _ L j
PATHWAY: Vegetables
AGE GROUPING
Average
6 month-old
Average
4 year-old
Average
14 year-old
Average
Adult
Maximum individual
[4 year-old]
1-131
0
4.09 x 103
6i55 x 103
9.83 x 103
6.87 x 103
1-133
0
3.95
6.33
9.49
5.39 x 102
NOTE: Only two figures are assumed to be significant.
122
-------
Table 43
PATHWAY TRANSFER COEFFICIENT [pCi/yr ingested per jCi/m3 in air]
PATHWAY: Mllk
AGE GROUPING
Average
6 month-old
Average
1 4 year-old
Average
14 year-old
Average
Adult
Maximum individual
(6 month-old)
1-131
8.41 x 104
1.18 x 105
, 1.11 x 105
3.88 x 10*
2.01 x 105
1-133
7.66 x 102
1.07 x 103
1.01 x 103
3.52 x 102
7.89 x 103
NOTE: Only two figures are assumed to be significant.
123
-------
factors from Table 38 are presented for each age group in Table 44
Evaluation of Total Thyroid Dose and Associated Health Risks
The population-integrated dose (in person-rems) is obtained
by multiplying the average air concentration within a sector
by the number of people enclosed within that sector and by a
dose conversion factor for an "average" individual in the popu-
lation. These dose conversion factors are obtained by summing
the age-dependent products of the pathway transfer factors and
the dose [equivalent] intake conversion factors for inhalation
(Table 40) and the two ingestion pathways (Table 44) and
multiplying this value by the fraction of the total U.S. popu-
lation which is in each of the four age groups (Table 36 ). These
values for each of the four age groups are then summed to give a
factor for the "average" individual. These calculations are shown
for iodine-131 in Table 45 and for iodine-133 in Table 46.
The risk of health effects (principally thyroid cancer)
resulting from radiation exposure of the thyroid appears to be
age-dependent. For this reason, the population-integrated dose
cannot be directly multiplied by a single dose-to-risk conversion
factor. This can be accomplished, however, if the age-dependent
risk per unit dose values are weighted by the fraction of the total
population dose which is delivered to that age group. The resultant
dose-to-risk conversion factor for an "average" individual are
relatively independent of the radionuclide (1-131 or 1-133) or the
chemical form of the radioiodine (elemental or organic) as shown in
124
-------
Tables 45 and 46. For this reason a value of 56 health risks per
million person-rem was used for all integrated population doses. The
age-dependent risk values for each age group were obtained from data
presented in reference 88.
125
-------
Table 44
Product of the Pathway Transfer Coefficient with the Dose Equivalent Conversion Factor
Imrem/year per pCi/m^ in air]
MILK PATHWAY
AGE GROUPING
Average
6 month-old
Average
4 year-old
Average
14 year-old
Average
Adult
Maximum individual
1-131
1.7 x 103
1.9 x 103
3.4 x 102
6.2 x 101
4.0 x 103
1-133
4.8
4.8
8.1 x 10"1
1.4 x Hf1
5.0 x 101
VEGETABLE PATHWAY
AGE GROUPING
Average
6 month-old
Average
4 year-old
Average
14 year-old
Average
Adult
Maximum individual
1-131
0
6.5 x 101
2.0 x 101
1.6 x 101
1.1 x 102
1-133
0
1.8 x 10"1
5.1 x 10~3
3.7 x 10"3
2.4
126
-------
In Popula-
tion
Table 45
Radlolodlne Populatlon-Uelghted Dose Conversion And Dose-To-Rlsk Conversion Factors Iod1ne-I3l
Age Group
0-1
1-9
10-19
>20
Average
l«uHul>fiial
Age of Typical Percentage of
Individual Total Popula- Risk per 10°
(Tears) lion man-rem Pathway
0.5 1.79 88 Inhalation
Vegetable
Milk
Sub- total
4 16.47 50 Inhalation
Vegetable
Hilt
Sub- total
14 19.57 66 Inhalation
Vegetable
Milk
Sub- total
20 62.17 60 Inhalation
Vegetable
Milk
Sub- total
Dose Conversion Weighted
Factor Factor
(mrem/yr per pCt/m3)
17
0
1700
42
65
1900
15
20
340
8.8
16
62
elemental
.3043
0
30.43
30.734
6.917
10.705
312.93
330. 553
2.936
3.914
66.538
73.388
5.471
9.947
38.545
53.963
488.64
Dose Percentage of
Total Population
Dose
organic elemental organic
.3043
0
.0304
.3347 6.29 2.o8
6.917
0.011
0.313
7.241 67.65 44.97
2.936
0.034
0.066
3.006 15.02 18.67
5.471
.010
.038
5.519 11.04 34.28
16.10 iop.0 100.0
Weighted Risks
per 106 man-rem
to Average Popula-
tion
elemental organic
5.53 1.83
33.82 22.48
9.91 1?.32
6.62 20.57
55.90 57.21
Mote: Only two figures are assumed to be significant.
-------
S3
00
Individual
1n Popula-
tion
Table 46
Radlolodlne Population-Weighted Dose Conversion And Dose-To-Rlsk Conversion Factors Iod1ne-133
Age Group
Range
0-1
1-10
10-20
>20
Average
Age of Typical Percentage of
Individual Total Popula- Risk per 10°
(Tears) tlon man-rem Pathway
0.5 1.79 88 Inhalation
Vegetable
Milk
Sub- total
4 16.47 SO Inhalation
Vegetable
H1lk
Sub-total
14 19.57 66 Inhalation
Vegetable
Milk
Sub-total
20 62.17 60 Inhalation
Vegetable
Milk
Sub- total
Weighted
(mrem/yr
Elemental
0-0984
0
.0859
0.1844
1.9760
.0296
.7906
2.797
.7632
.0010
.1585
.9227
1.3680
.0023
.0870
1.457
5.361
Dose Factor
per pC1/m3)
Organic
0.0984
0
<.000)
0.0984
1.97CO
<.0001
.0008
1.977
.7632
<.OOOJ
.0002
.7635
1.3680
•c.OOOl
«.0001
1.369
4.207
Weighted Risks
Percentage of per 106 man-rem
Total Population to Average Popula-
Dose Hnn
elemental organic elemental organic
3.44 2.34 3.03 2.06
52.17 46.99 26.08 23.50
11.36 11.98
27.18 32.52 16.31 19.51
100.0 100.0 56.78 57.04
-------
ESTIMATED RADIATION EXPOSURE
FROM NUCLEAR REACTOR EFFLUENTS
Radiation exposure to the population within 50 miles of the reactor
site was calculated using the methods outlined in the preceding section.
These exposures were determined for each of the postulated treatment
options and for each of the three representative site classes described
in Table 1. In addition, a hypothetical "average site" was constructed
from these three sites by weighting the radiation dose at each site by
the projected distribution of sites:
D"average" = 0.25DseaCoast + 0.60Drlver + 0.15Dlake
Liquid Effluents
It is convenient to distinguish the radiation dose from tritium
releases and the dose produced from all other radionuclides discharged
in liquid effluents. These doses are shown for tritium in Table 47
and for all other radionuclides in Table *8 . Despite the relatively
large contribution to the total activity (curies) discharge due to
tritium, its biological significance in terms' of radiation exposure
to local population groups is very small. This disproportionate sig-
nificance results from the radiological parameters associated with tritium,
primarily the low average energy of its beta emission (6 keV compared to
1131 keV for the combined decay of strontium-90 and its daughter yttrium-90),
its relatively short residence time in the body (12 days) resulting
from its incorporation into body water, and its generally uniform distri-
bution throughout the body instead of being localized in one organ.
129
-------
Table 47
Dose Equivalents from Tritium Releases
System
BWR-1,2,& 3
BWR-4
PWR-1,2
PWR-3,4
Population Dose Equivalents (man-rem/yr)
Whole Body
Seacoast
.0015
.00095
.0087
.0055
River
.12
.08
.74
.47
Lake
.02
.013
.12
.076
Thyroid
Seacoast
.0015
.00095
.0087
.0055
River
.12
.08
.74
.47
Lake
.02
.013
.12
.076
Individual Dose
Equivalents (mrem/yr)
Whole Body
Seacoast a
.00012
.000079
.00074
.00047
River
.001
.00073
.0067
.0042
Lake
.001
.00073
.0067
.0042
Annual
Discharge
(Gi/vr)
200
130
1200
760
Does not include shellfish.
-------
Table 48
SUMMARY TABLE: LIQUID RADIOACTIVE WASTE SYSTEMS, ESTIMATED COSTS, AND DOSE EQUIVALENTS
System
BWR System*:
8WR-I
BWR-2
BWR -3
BWR -4
PWR Systems'
PWR-1
PWR -2
PWR -3
PWR-4
Estimated
Capital .
Co.1
918,000
1.738.000
2,344.000
3.231.000
264.000
509.000
1.213,000
i3.547.000
Estimated
Annual
Con
18O.OOO
401. OOO
660,000
788,000
52,000
121.000
280,000
879.0OO
Dose Equivalents From Non-Tritium R«dio*etiv« Releases
Individual Whole Body
Dos* Equivalents Imrem/vrl
Seacoast
2.2
0,01 S
00032
00001
60
Oil
0012
-------
For light-water nuclear power reactors there are only two important
pathways for tritium exposure: ingestion in food and drinking water. The
other principal radionuclides, however, produce radiation exposure via
several exposure modes. As the effluent composition changes with different
treatment options, the relative contributions from these different pathways
also changes. For the pressurized -water reactors, (PWRs), the only
radionuclides which produce significant radiation exposure are tritium
and the two cesium isotopes: cesium-134 and cesium-137. These latter two
radionuclides comprise over 95% of the total whole-body dose from PWR cases
1 and 2. For ingestion of drinking water and fish, the cesium-134 delivers
about 60% of the dose and the cesium-137 about 33%. Tritium becomes
significant in PWR case 3, accounting for most of the radiation exposure
from drinking water ingestion, but the cesiums are still significant via
fish consumption due to the high biological reconcentration factor exhibited
by freshwater fish. Only in PWR case 4 does tritium become the largest
source of exposure, the cesiums accounting for only 25% of the dose from
fish consumption.
For the boiling-water reactors (BWRs) other radionuclides become
significant in addition to tritium and the two cesiums. Because of the
lower amount of tritium discharged from boiling-water reactors (200 curies
versus 800 from the PWR station), it does not constitute a significant
source of radiation dose until the majority of other radionuclides are
removed in BWR treatment options 3 and 4. In BWR cases 1 and 2, the dose
contribution from fish consumption is duo to the two cesiums, with cesium-134
contributing over 60% of the dose as a consequence of the greater cesium-134
132
-------
activity released. For drinking water intake a variety of radionuclides
contribute to the radiation dose, the most important being the two cesiums
and strontium-90. These three radionuclides comprise 60% of the radiation
dose delivered via this intake pathway.
In BWR cases 3 and 4 tritium contributes significantly to the dose
from drinking water intake, accounting for about half of the dose for case
3 and essentially all of the dose in case 4 from drinking water. The
tritium contribution from fish intake in case 4 is about 15% of the total,
the two cesiums contributing the remainder of the dose.
Radiation doses from present liquid effluent control systems are
relatively small compared to other effluent paths. Present treatment systems
(BWR-2 and PWR-2) result in individual whole body dose equivalents of between
0.1 and 2.2 mrem/year and whole-body population-integrated dose equivalents
between 9 and 180 person-rem. The radioiodine discharges produce thyroid
population dose equivalents between 160 and 1100 person-rem, the
larger value associated with PWR effluents.
Table 48 shows that the dose reduction factor for liquid effluent
treatment systems can reach approximately 1,000,000 for the cases PWR-4
or BWR-4 as compared to the hypothetical zero treatment systems (BWR-1
and PWR-1). In all cases, the river site produces the maximum dose as
a consequence of the smaller dilution factor afforded at this site.
The population doses however, differ appreciably between the two reactor
types and varies by an order of magnitude or more in some cases. The dose
variation between BWRsand PWRs is due partly to the different composition
of the effluent stream.
133
-------
Noble Gases
The radiation exposure from pressurized water reactor gaseous effluents
is relatively low due to the small quantity of undecayed noble gas
radionuclides emitted. These releases are much smaller from the PWR than
from an equivalent boiling-water reactor, as indicated in Tables 49 and 50.
This is a consequence of the presence of a secondary coolant loop in the
pressurized-water reactor. In the BWR, the radiolytic hydrogen and oxygen
and other air ejector effluents constitute a significant volume. In
pressurized water reactors, the secondary coolant system isolates the
condenser air ejector from the radiolytic gases in the primary coolant and,
consequently, yields significantly smaller off-gas volumes. This permits
tank storage with the resultant decay of short-lived gaseous radionuclides.
The integrated population dose equivalents within 50 miles at the three
representative sites and an "average" site are shown in Table 49 for a
pressurized water reactor. As can be seen, the population integrated dose
equivalents are small, even for the hypothetical zero treatment case
Class-0 and the great majority of the dose results from the 5.3-day xenon-
133 for all treatment options. This is a consequence of its relatively
long half-life compared to the other noble gases. Although krypton-85
has a significantly longer half-life (10.8 years), its production in fission
is about a factor of twenty lower than xenon-133 and it emits only a low
energy beta particle. The comparable contribution of krypton-85 is about
0.006 person-rem/yr within fifty miles of the reactor. Due to the long
half-life of krypton-85, its release is not greatly affected by holdup of
the offgas as is xenon-133, and only true separative noble gas control
systems such as cryogenic distillation or selective absorption modify the
release of krypton-85.
134
-------
Table 49
SUMMARY TABLE: PWR NOBLE OAS DOSE EQUIVALENTS AND HEALTH EFFECTS (2 uniti, 1000 MW(«) •ach)
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-------
As shown in Table 49 the application of cryogenic distillation or
selective absorption (Class 2A) results in virtually the same individual
and population doses as the use of 60-day xenon holdup (Class 1A-60).
These two classes of discharge control options are very effective in
minimizing primary gas releases, so much so that secondary sources of release
now dominate the sum of all releases from all pathways. To illustrate this,
a class of discharge control optionswas devised such that 15-day xenon
holdup (Class 1A-15) was followed by selective absorption for cryogenic
distillation (Class 2A). This combination is shown as Class 3 and results
in only slightly reduced population and individual dose equivalents (see
table 37).
One last discharge control option considered for primary gases is the
cover gas recycle system. Because this option reduces the primary coolant
concentration of longer-lived (i.e., >12 hr) radionuclides, the release of
these nuclides from all pathways is reduced, resulting in a further factor
of two reduction in individual dose equivalents but a factor of four reduc-
tion in population doses.
Finally, one option for treating a secondary source of noble gases
was considered, namely, the use of charcoal adsorption beds on the air ejector
(Classes 5A through 5D). This option is only effective for those cases in
which secondary sources dominate the total release (Classes 5B, 5C, and 5D).
Individual doses and whole body population doses are reduced slightly in
each case.
As previously discussed, the differences in design between BWRs and PWRs
allows greater release rates from the BWR, especially in the case of shorter-
lived radionuclides. As shown in the summary table, Table 50, a BWR of the
136
-------
Table 50
SUMMARY TABLE: 8WR NOBLE GAS DOSE EQUIVALENTS AND HEALTH EFFECTS (2 UNITS, 1000 MW(E) EACH)
Diackanjl Control Option
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$7,000,000
tijaooMo
»12,6OOWO
»7J»0,OOO
te,«ooxMO
$9.000,000
S8.000.000
$13,800,000
Estimatod
Anmul .
COM
* 6OO.OOO
$1.400.000
$1,2OO,OOO
$1430.000
$1,62OJXX)
S1.910.000
$1,4OO^OO
\MOXOO
meOOaaO
$1*»,000
$1,400,000
$2,110^00
$1*00,000
$2,800,000
Mnimun lndi»idu«l Wkota Body OOM Equintant RIM
(mram/rrl*
SHt Boundary
IjMtOOIt
104.
2.S
2.7
1.1
1O
1X1
1.4
10
ID
IB
u
0.1
O.S
0.1
flMtr
323.
13
11.8
43
43
4.2
6.0
4.4
4.2
61
a2
0.5
2J
0.5
L««
222.
5.3
116
4.8
4.3
4.3
6.0
4.4
43
4.0
7.9
0.5
2.3
0.5
Avwagt
Sin
2S3.
6.1
9.5
3.9
3.4
3.4
4.8
3.6
3.4
4.7
6.5
O.4
1.9
0.4
N*>rM Ruidont
SWCOilt
72.8
\a
ia
1.2
1.1
1.1
1.5
1.1
1.1
1.4
1.9
0.1
06
0.1
RMr
214.
S.I
7.2
2.7
2.4
24
35
2.5
2.4
4.0
5.0
0.3
1.4
0.3
Laki
222.
5.3
11.6
4.8
4.3
4.3.
6.0
4.4
4J
4.0
7.9
OS
2.3
0.5
Avarofa
Sita
180.
4.3
6.7
2.7
24
2.4
3.4
2.4
14
3.4
4.7
0.3
1.3
OJ
Whoto Body Population
Data CaunaMm Rat.
(paraon-ratn/yr)*
5~C~M
1030.
35.8
22.5
5.1
3.8
38
7.3
4.1
38
328
193
0.6
4.1
OS
n»>
4810.
207.
148.
34.7
24.0
23.4
44.7
25.3
23.4
183.
127.
3.6
24.7
3.7
L.k.
521.
18.6
14.2
3.5
2.6
2.6
4.8
2.8
2.6
17.
12.1
0.3
2.6
0.3
A..«.
Sita
3220.
140
99.4
26.6
19.8
19.4
29.4
16.6
1S.4
131.
86.8
6.4
16.2
2.4
Annual Haalth Effects
0.72
0.0250
0.01 57
0.0036
0.0027
0.0026
0.0051
0.0029
0.0026
0.0229
0.0135
0.0004
0.0028
0.0004
Rinr
3.37
0.145
0.102
0.0243
0.0166
0.0164
0.0313
00177
0.0164
0.1350
0.0889
0.0026
0.0173
0.0026
Lafca
0.365
0X1130
0.01OO
0.0025
0.0018
0.0018
0.0034
00070
0.0018
0.0119
0.0084
0.0002
0.0018
O.OOO2
Av.»g.
Sit.»
2.25
0.0981
0.0696
00186
0.0138
0.0136
00206
00116
0.0103
0.0908
0.0608
0.0049
0.0114
0.0017
aDose equivalents and health effects from noble gases only. Radioiodines are also removed from the air ejector
effluent and contribute 1470-3710 person-rem/yr and 0.083-0.21 health effects/yr at the average site (where
the low estimate assumes organic radioiodine and the high estimate assumes elemental radioiodine) for all
discharge control options except Class 0.
Includes worldwide krypton-85 commitment.
u>
-------
1,000 MWe size with the control system typical of BWRs presently operating
. would deliver site boundary whole body dose equivalents of over one hundred
millirem. Population dose equivalents might be expected to reach over one
thousand person-rem/yr at an average site with one 1,000 MWe unit.
Furthermore, the release of radioiodine would account for site boundary
individual dose equivalents of many hundred millirem and thyroid population
dose equivalents of almost two thousand thyroid person-rem at an average
site with one unit (see Table 52, page 139).
In selecting the discharge control options, consideration was given to
those options in current use as well as to those presently committed.
Classes of options were arbitrarily defined. Unlike PWRs, dose equivalents
are not dominated as strongly by xenon-133.
The first option considered was the use of physical holdup to allow
decay of shorter-lived radionuclides. This system is similar in concept
to the typical PWR gas decay tank system. However, long decay times (such
as 60 days) cannot be routinely obtained because of the high air ejector
offgas flow rates, even after the use of a recombiner to eliminate the
condensible gases. Such a system is limited by the condenser air in-leakage
flow rate,and delay time is inversely proportional to such in-leakage.
Table 50 shows that for a nominal 24-hour delay, individual whole body
dose equivalents are lowered to less than five millirem and population whole .
body dose equivalents are reduced to less than 100 person-tern for a 1,000
MWe BWR at an average site. However, radioiodine dose equivalents are
less than one per cent of those for the source term, Class 0, discussed above.
The second discharge control option considered was the use of a
recombiner followed by ambient temperature charcoal adsorption beds. As
138
-------
shown in Table 50 for Classes 2-10 through 2-60, a choice exists as to
the number of beds and hence the delay achieved. Class 2-10 fails to
quite achieve the individual whole body dose equivalents of the 24 hour
holdup system, Class 1. This is due to the differing delay times of
xenon and krypton in the charcoal beds; although a 10-day xenon decay is
achieved, the kryptons are delayed only one-fifteenth as long, allowing
more of the radiologically significant short-lived krypton isotopes to be
released. Population dose equivalents, however, are lower than Class 1
as the xenon isotopes are delayed effectively enough to offset the increase
in population dose equivalent achieved by krypton-88, now released without
a 100 m stack. Class 2-20 is capable of reducing both the individual and
population whole body dose equivalents to well below those of Class 1.
Only slight reductions in dose equivalents are achieved by the addition of
more charcoal adsorber beds. It should be pointed out that because of
the recombiner condensers and the huge amounts of charcoal employed,
radioiodine is also removed from the air ejector effluent.
Cryogenic distillation or selective absorption, the third discharge
control option selected, appears to be about as effective as the charcoal
adsorption systems providing 20- to 40-day xenon delay. However, performance
projections are more uncertain and thus the evaluation for two different
xenon DFs, Classes 3A and 3B. Radioiodine is also effectively removed
from the air ejector offgas.
Combining a charcoal adsorption system and a cryogenic distillation
system (or selective absorption), the fourth discharge control option
selected, provides negligible reductions in dose equivalents over the use
of either alone compared to the incremental cost incurred.
139
-------
Finally, one option for treating a secondary source of noble gases
was considered, namely, the application of a clean steam source to the
turbine gland seal. This option is more effective for those cases in
which secondary sources dominate the total releases. Reduction factors
of from 1.3 to 10.0 in the individual whole body dose equivalents are
possible but whole body population dose equivalents are reduced by no
more than a factor of three.
Radioiodine
The radiation doses from PWR radioiodine releases are shown in
Table 51 for both the general population integrated dose and the dose
to the maximum individual at the site boundary. Table 51 shows that,
with the release conditions postulated in preceding sections, the PWR
site boundary thyroid dose equivalent can be reduced to 10-22 mrem/year
for radioiodine at an average site (the lower dose being due to the
smaller contribution from milk for the organic iodines). However, the
assumption of an infant spending 100% residence at the site boundary
is not realistic, as the nearest residence and dairy farm are not
usually located at the same position nor at the site boundary. Dose
equivalents at all three locations are given in Table 51 to show the
reduction that may be obtained.
The radioiodine doses from BWR effluents are presented in Table 52.
The prohibitively high site boundary doses shown for a release from
the 30-minute holdup system are purely hypothetical. All current plants
employing this system are much smaller and are planning to retrofit
augmented systems which will effectively minimize radioiodine from th
140
-------
Table 51
SUMMAtY TABU: PWR RADIOIODINE DOSE EQUIVALENTS AND HEALTH EFFECTS (2 unit., KXX> MW(.) each)
Control Option Arid.*
Strain Gwwator BloMdoiwrt Vefittd 1O Mem ContftM***'
Avft.ti.vv Bid*. Clwrco*! A4K»-l.*rta
A.H- EpKtOT ChtVCCWt A4»0f *«
UpgrK* "> DwP B«d Cturcon AdtOttMr *«.»»«•> Bkff *
CtraflStMMi V*lMt>2.»in Duwn.mr
Ttirton* Gt»td $Ml Cten SMWn
Nan*
ConUwWkBnt Krtlntv
A* CfKiw Cht>co*y AAorto*
Awiifwrv iW|. Ch.vc.Ml A4MrlMrb
Up9r«d« to OM0 8-Jd ClMwCOfl AdMrter , AuKdwy Bwtdinf1'
Cttw. SlMffi: V«l«««. >2fttn frMWMr
OUndtmOt-W *••«
—
TOIE-
PGIE-
POIt-
PCIE
fGIE
Kit
roio-
KIO
KIO
KIO
KIO
KtO •
KIO
CUM*
S 0
8 MO.OOO
1 2 950.000
tl.UO.000
84.1)0.000
t5.9K.000
86.510.000
8 0
8 700.0OO
tl.tlO.OOO
SliN.OOO
M.130.000
86.930.000
86.UO.OOO
Annul
CM
0
180.000
460.000
560.000
860.000
1.160.000
I.MO.OOO
0
120XWO
260000
MO 000
860.000
.160.000
.J80.000
nuinr
ICI/yrf
1 131
1 51
0464
0 175
0085
0054
0022
00>1
41O
384
184
2.»7
1 19
1 162
1 161
1-133
062*
0226
0 100
0044
0.032
0013
Ml
167
225
0806
0786
0786
E^iMvatanf AM* !mi«*lt/vrl;
SJuBowfidwv
S«««|
232
74
270
34
791
71 3
338
51
31
22
2}
Rww
667
182
558
81
181
174
828
126
66
54
54
Lid*
337
108
391
49
115.
103
490
76
11
33
12
A.«w
tiw
443
144
521
66
62
1U
138
651
10.0
46
43
43
Do*» E«u*»>lwi* RM« li*ic«
llitlM
116
37
115
157
161
26
1 1
1.1
1 \
Mww
175
104.
177
lit.
99.9
471
7.J
33
1.1
11
L*.
337
108
18.1
lift.
101.
490
7.6
31
32
33
vaw«
~/».l:
Siu
J75
879
11.8
96
818
843
400
61
27
26
28
Daw tvlol**H fl*w (MMm/yr):
S~-«
66
20
07
02
33
20
09
O1
01
01
01
Rmr
266
92
33
10
97
8.6
42
06
01
03
03
UIM
27.3
87
32
1.0
93
64
40
06
03
0.3
03
A>««*
8iw
22.9
73
2.6
0.8
78
70
33
05
O3
03
03
EauivBlMlt RM>:
(ptnon-ram/yr)
fa«»«
504
162
59
16
543
469
231
35
\&
15
15
RMr
296
95
34
10
319
268
136
207
92
90
69
L*.
32
102
37
11
0.5
141
309
146
22
10
10
1.0
CM
195.
62.6
22.7
110
70
29
27
210
189.
894
116
6.1
6.9
59
Annual
Awn*
SIM
1 . H- 21'
3.5f- 31
1 .31- 31
6.K-4I
3.91-41
1.61-41
1 51- 41
1.21-21
1.11-21
S.OI-D
7.61- 4]
3 41- 41
3.31-41
3.3<- 41
aAn expression such as 1.1 (-2) is equivalent to 0.011.
Containment purge is also routed through this-filter.
-------
10
Table 52
SUMMAIY TABLE: BWR tADIOIODME DOSE EQUIVALENTS AND HEALTH EFFECTS (2 unto, MOO MW(.| .o2.S~ DIMMIW
TurtwM BU| OMTOMI Acborbw
TwtoNw Gwtd Seal Oan Siwm
OWfMtwn
BGIE 1A
BGIE-
BGIE
BGIE
BGIE
BGIt
BGie
BGIE
BGIO
BGIO
BGIO
BGIO
BGIO
BGIO
EttinwM*
(cwntwUtwvl
(S3, 200.0001
S 0
$1,800.000
$4, 300 .OOO
Sb 100.000
S7.35O.OOO
S 7.950.000
S2.000.000
S2.SOO.OOO
S7.7SO.OOO
S4.55O.OOO
S7.0SO.OOO
S7.6SO.OOO
CMNfttM
COTI
-------
air ejector offgas, which is the largest source of radioiodine. Thus, the
"source term" for future BWRs will consist of only secondary release
pathways, case BGIE-1 or BGIO-1, in Table 52.
The data in Table 52 show that, with the release conditions postulated
in preceeding sections, the BWR site boundary thyroid dose equivalent
can be reduced to 10-14 mrem/yr for an average site with two 1,000 MWe
units (the lower number being attributed to a 100% organic radioiodine
release). However, the assumption of the continual presence of an
individual at the site boundary is not realistic, since the nearest
residence and dairy farm are not usually located at the same position nor
at the site boundary. To illustrate the reduction in dose which may be
obtained, Table 52 lists the dose equivalents for the maximum individual
at all three locations. It appears that an order of magnitude reduction,
or more, is possible as one proceeds from the site boundary to the nearest
farm.
Depending on the assumed chemical form of the radioiodine discharged,
the critical individual will change from the 6-month-old infant, for the
case where the iodine is 100% elemental, to the 4-year-old child for the
case where the iodine is 100% organic. This is because for the latter
assumption the inhalation pathway prevails, and the combination of internal
dose factor and breathing rate for the child results in a high dose
equivalent rate for a given air concentration of radioiodine. However,
in the realistic case where some iodine is expected to be elemental, and
if the assumption is made that a milk pathway exists, the 6-month-old
infant will be the critical individual, and the individual thyroid dose
equivalent calculated for that age will be limiting.
143
-------
Worldwide Dose Contributions
In dddition to producing local radiation exposure, two long-lived
radionuclides are produced which can accumulate in the biosphere and which
are capable of migrating over large distances. The two radionuclides of
concern are the 12.3-year radioisotope of hydrogen, tritium, and the
10.8-year noble gas, krypton-85. Tritium is released from reactors via
liquid effluents as tritiated water, HTO, and in this form enters into the
water cycle. The evaporation and subsequent movement of tritiated water
vapor permits worldwide dispersion. Krypton-85, however, is a non-reactive
gas and diffuses in the atmosphere.
Although the dose contributions from these two radionuclides may be
significant in other portions of the fuel cycle, their cumulative effect
is much smaller from reactors (Tables 53 and 54) because of the smaller
release rates from power stations compared to fuel reprocessing plants.
The annual discharge of these two materials from a reactor is approximately
equivalent to the daily discharge rate from a spent-fuel reprocessing plant
and the resultant exposures are correspondingly smaller for the reactor.
144
-------
Table 53
Worldwide Health Risk Contributions from Reactor Tritium Releases
TRITIUM
TREATMENT
OPTION
BWR-1,2,3
BWR-4
PWR-1,2
PWR-3,4
ANNUAL8
DISCHARGE
(Ci/yr)
200
130
1200
760
TOTAL3
ANNUAL DOSE ANNUAL HEALTH a»d TOTAL HEALTH
COMMITMENT RISK COMMITMENT RISK COMMITMENT
(Man-Rem) WHOLE BODY THYROID 30 YR
0.18 0.126(-03)
0.117 0.819 (-04)
1.08 0.756 (-03)
0.684 0.479(-03)
.331 (-05) 0.0019
.215 .(-05) 0.0013
.199(-04) 0.0116
.126 (-04) 0.0074
TOTAL
YEAR 2000 30-YR COMMITMENT0
FOR ALL U.S. PLANTS
PERSON-REM HEALTH EFFECTS
1,080 0.78
702 0.50
12,960 9.3
8,208 5.9
Note: .479(-03) » .479 X 10~3 or .000479
a for 2 1 GWe plant site
b per GWe reactor
c 400 GWe BWR; 800 GWe PWR
d assuming worldwide age breakdown equivalent to that given for U.S.
-------
Table 54
Worldwide Health Risk From Reactor Krypton-85 Releases
System
Annual Annual 100-yr
Discharge Integrated Pop-
(Ci/yr)a ulation Dose
Commitmenta
(person-rem)
Annual Commitment from
Health 30-yr 1 GWe
Effects Plant Operation
Committed3
person- } health
rem I effects
Total Commitment from
30-yr Operation of
all U.S. Plants in
Year 2000b
person-
rem
health
effects
BWR 30-min Holdup 760.0 0.266
& All Delay Systems
BWR Cryogenic 7.8 ,0028
Distillation
(Kr DF = 250)
BWR Cryogenic 3.1 .0011
Distillation
(Kr DF = 2,500)
PWR Source Term & 1600.0 0.56
All Delay Systems
PWR Cryogenic 50.0 .0175
Distillation
(Kr DF = 250)
PWR Cryogenic 45.0 .0158
Distillation
(Kr DF = 2,500)
PWR Cover Gas 4.4 0.0015
Recycle
0.186(-03) 4.0 0.28(-02) 1596.0 1.1
.19 (-05) 0.041 .28(^04)
.110(-04)
16.5 0.011
.76 C-06) .016 .114 (-04) 6.5 .0046
,392(-03) 8.4 .588(-02) 6720.0 4.70
.123(-04) .262 ,184(-03) 210.0 .147
.236 .166(-03) 189.0 .142
.108(-05) 0.023 .162(-04) 18.5 0.013
aPer 2 unit plant.
b 400 GWe BWR, 800 GWe PWR.
-------
ECONOMIC AND ENVIRONMENTAL COSTS
Total Costs
The total cost of a treatment system was calculated from the
annualized cost as being equivalent to the present worth of the annualized
costs using a 7.5% present worth rate. The present worth of all costs is
defined as
m
P.W. « Z Pn
n=l
where "Pn" is the cost in the nth year and "r" is the applicable interest
rate. For the present analysis, r = 0.075, m - 30 years, and due to the
depreciation method assumed in the analysis, the annualized costs, Pn - P,
are constant so that this term may be removed from the sum:
30
P.W. = P I 1
n-1 (1.075)^
The present worth factor, which is the sum given in the above equation,
is equal to 11.8104 for the parameters given. Thus the total cost of a
reactor treatment system is equivalent to 11.8104 times the annualized cost
of that system. The total economic impact of a given discharge control
option may be .determined by multiplying its total cost per nominal GWe reactor
by the appropriate number of reactor plants assumed operable by the year 2000
(800 PWRs and 400 BWRs).
147
-------
The cost to the consumer, which is an important consideration in ascer-
taining the economic impact of any effluent control measure, was computed
from the annualized cost using a capacity factor of 0.8. The conversion
factor to mills per kilowatt-hour from annualized dollars per GWe plant is
1.43'x 10~7.
The environmental impact was determined by multiplying the annual
impact per GWe plant by the operating lifetime of the plant (30 years).
The total environmental impact for that type of plant was then computed by
multiplying the impact for a GWe plant by the number of such plants predicted
to be in operation by the year 2000 (800 for PWRs and 400 for PWRs). Included
in this estimate are also the worldwide impacts from radionuclides with long
half-lives, e.g., krypton-85 among the noble gases, tritium, cobalt-60,
strontium-90, cesium-134, and cesium-137 in the liquids, and iodine-131
among the gaseous radioiodines.
The economic and environmental impacts associated with selected effluent
control options for an average site are presented in Tables 55-61. These
tables illustrate the range of system effectiveness, economic commitments,
and environmental commitments that may be achieved for a nominal one GWe
plant and for all such plants estimated to exist by the year 2000. The
health risks associated with the operation of a single plant are less than
six per year, even assuming the hypothetical "zero treatment" options for
all release pathways. The annualized cost for the "maximum treatment"
options on all release pathways amounts to about $2,800,000 for a BWR and
about $1,500,000 for a FWR. The added costs of these treatment systems to
the consumer is also small (up to a few tenths of a mill per kilowatt-hour
for the most advanced systems).
148
-------
On a national scale, however, both health risk commitments and control
costs become appreciable. With the hypothetical "zero treatment" cases
(minimal control costs), total health risks would reach into the neighborhood
of 100,000, while minimal health risks (on the order of 25) may be achieved
with the expenditure of billions of dollars. '
Cost-Effectiveness and the Consumer Perspective
Cost-effectiveness for effluent control systems may be considered
from two different but complementary viewpoints: the cost per risk averted
or the cost per benefit received. Tables 55-61 detail the cost per risk
eliminated by various types of effluent control options applicable to light-
water reactors. Except for the radioiodine control options, the cost per risk
eliminated increases from about $10,000 - $100,000 (for the first option
addition) to well over one million dollars for "maximum treatment" options.
Although radioiodine discharge control options are generally more expensive
in terms of cost per risk reduced, they more effectively limit the maximum
individual dose.
Despite a national total expenditure in excess of 30 billion dollars,
the codt of achieving "maximum treatment" for all reactors, the added cost
to the consumer per benefit received would be only about 0.38 mills per
kilowatt-hour of electricity consumed. When compared to production costs
(Table 62) or typical charges for power (Table 63), this cost appears small,
making up about 57, of power production costs and eVen less for typical power
charges. For'a typical consumer using 7,700 kilowatt-hours per year
(based upon average residential comsumption of 5491 kWh in 1968), the total
annual cost of achieving even this maximum level of control would amount
to $2.92, or less than 25 cents per month.
149
-------
Although the consumer cost for installation of the maximum control
technology appears small on a mills per kilowatt-hour basis, the total
cost involved amounts to over 30 billion dollars. Analysis of the total
risks reduced after the addition of two or three discharge control options
to the various source terms ("zero treatment") generally shows relatively small
reductions. Justification for effluent control options is generally dif-
ficult to provide since other technical factors (such as maximum individual
dose equivalent), as well as political and social considerations, usually
enter into the decision-making process.
150
-------
Dollar Costs Per GWE
TABLE 55
BWR Noble Gas Systems: Cost-Effectiveness
Environmental Costs
Cost Effectiveness
System
Designation
CLASS
CLASS
CLASS
CLASS
CLASS
CLASS
Ob
2-10
2-20
2-60
4
5E
Annual
Cost
($1000)
300
600
665
955
1300
1400
Consumer
Coat
(mill/kWh)
0.043
0.086
0.095
0.14
0.19
0.20
Present
Worth
($Millions)
3
7
7
11
15
16
.043
.086
.854
.279
.354
.535
Annual Per OWE
Person-Ren Health
Risk
2910.
49.7
13.3
9.7
7.7
1.2
1.2(o)C
3.5(-2)
9. 3 (-3)
6.8(-3)
5. 4 (-3)
8.5(-4)
30-YR Total: All Reactors
Person-Rem Health
Risk
3.5(7)
6.0(5)
1.6(5)
1.2(5)
9.2(4)
1.4(4)
1.4(4)
4.2(2)
1.1(2)
8.2(1)
6.5(1)
1.0(1)
^Dollars
AHealth
Risk
1.0(5)
1.0(6)
4.6(7)
9.7(7)
8.7(6)
iHo.-il th
Rirsk
i Dollars
9. 8 (-6)
1.0(-6)
2.2(-8)
1.0 (-8)
1.2 (-7)
Collars are present worth dollars and health risks are 30 year total; each Is placed on a per GWe basis.
''includes 1300 person-rea/yr per GWe and 0.074 health Tlsks/yr per GWe due to radlolodine released at air
ejector (assuming that the radiolodine released is 502 elemental and 50Z organic). Remaining systems listed
do not release this source of radiolodine to environment.
cAn expression such as 1.2(0) is equivalent to 1.2x10 , or 1.2.
-------
Ui
to
System
Designation
CLASS 0
CLASS 1A-15
CLASS 1A-60
CLASS 2A
CLASS 3
CLASS 5C
Dollar Costs Per CUE
Annual
Cost
($1000)
0
30
75
300
330
405
Consumer
Cost
(mill/ kWh)
0.0000
0.0042
0.011
0.043
0.047
0.058
TABLE 56
FUR Noble Gas Systems: Cost-Effectiveness
Environmental Costs
Present
Worth
($Millions)
0.000
0.708
1.771
7.086
7.794
9.566
Annual Per GWE
Person-Rem
44.6
9.2
4.9
0.8
0.8
0.7
Health
Risk
3.1(-2)b
6.4(-3)
3.4(-3)
5. 5 (-4)
5.5(-4)
4.5(-4)
30-YR Total:
Person-Rem
1.1(6)
2.2(5)
1.2(5)
1.9(4)
1.9(4).
1.6(4)
:A11 Reactors
Health
Risk
7.5(2)
1.5(2)
8.2(1)
1.3(1)
1.3(1)
1.1(1)
Cost Effectiveness3
ADollars
AUealth
Risk
4.8(5)
5.9(6)
3.1(7)
c
CD
3.0(8)
AHealth
Risk
ADollars
2.K-6)
1.7(-7)
3.2(-8)
0
3. 4 (-9)
a Dollars are present worth dollars and health risks are 30 year totals; each Is placed on a per GWe basis.
b An expression such as 3.1(-2) is equivalent to 3.1x10" , or 0.031.
c Because Class 3 achieves the same level of health risk as Class 2A (but costs slightly more than Class 2A).
-------
TABLE 57
BUR Radioiodine Systems: Cost-Effectiveness
(Elemental Form)
Dollar Costs Per GWE
Environmental Costs
Coat Effectiveness3
Annual
System Cost
Designation ($1000)
BGIE-1 0
BG1E-2 150
BGIE-3 375
BCIE-4 600
BCIE-5 800
BCIE-6 820
BGIE-7 920
Consumer
Costs
(ralll/kWli)
0.00
0.021
0.054
0.086.
0.11
0.12
0.13
Present Annual Per GWE
Worth Person-Rem
($Milllons)
0.000 69.5
1.772 15.3
4.429 3.0
7.086 1.8
9.448 1.2
9.685 0.7
10.866 0.3
Dollars are present worth dollars and health risks are 30 year total; each is
An expression such as 3.9(-3) is equivalent Co 3.9x10 , or 0.0039.
H"
Ul
U)
Health
Risk
3. 9 (-3)
8.5(-4)
1.7 (-4)
1.0 (-5)
6.5(-5)
3. 8 (-5)
1.6 (-5)
placed
30-YR Total :A11 Reactors
Person-Rem Health
Risk
b 8.3(5) 4.7(1)
1.8(5) 1.0(1)
• 3.6(4) 2.0(0)
2.2(4) 1.2(0)
1.4(4) 7.8(-l)
8.4(3) 4.6(-l)
3.0(3) 1.9(-1)
on a per GWe basis.
ADollars AHealth
AHealth Risk
Risk ADollars
1.9(7) 5.2(-8)
1.3(8) 7.7(-9)
1.3(9) 7.9(-10)
2.2(9) 4.4(-10)
2.9(8) 3.4(-9)
1.7(9) 5.7(-10)
-------
u-.
Dollar Costa Per GWE
TABLE 58
BWR Radioiodine Systems: Cost-Effectiveness
(Organic Form)
Environmental Costa
Cost Effectiveness3
System
Designation
BGIO-1
BGIO-2
BGIO-3
EG 10-4
BGIO-5
BGIO-6
BCIO-7
Annual
Cost
($1000)
0
200
400
420
570
795
895
Consumer
Cost
(mill/kWh)
0.000
0.029
0.057
0.060
0.082
0.11
0.13
Present
Worth
($Mlllions)
0.000
2.362
4.724
4.960
6.732
9.389
10.570
Annual Per
Person-Rem
119.0
15.2
6.9
6.3
2.1
1.1
1.1
GWE
Health
Risk
6.5(-3)b
8. 5 (-4)
3. 9 (-4)
3.6(-4)
1.2(-4)
6. 5 (-5)
6.0 (-5)
30-YR Total:
Person-Rem
1.4(6)
1.8(6)
8.3(4)
7.6(4)
2.5(4)
1.3(4)
1.3(4)
:A11 Reactors
Health
Risk
7.8(1)
1.0(1)
4.6(0)
4.3(0)
1.4(0)
7.8(-l)
7.2(-l)
ADollars
AHealth
Risk
1.4(7)
1.7(8)
2.6(8)
2.5(8)
1.7(9)
2.3(9)
AHealth
Risk
ADollars
7.2(-8)
5. 9 (-9)
3. 8 (-9)
4.1(-9)
5. 9 (-10)
4. 3 (-10)
aDollars are present worth dollars and health risks are 30 year total; each is placed on a per GWe basis.
"An expression such as 6.5(-3) is equivalent to 6.5xlO~ or 0.0065.
-------
Dollar Coats Per GWE
TABLE 59
PWR Radioiodine Systems: Cost-Effectiveness
(Elemental Form)
Environmental Costs
Cost Effectiveness
System
Designation
PGIE-1
PGIE-2
PGIE-3
PGIE-4
PCIE-5
PGIE-6
PGIE-7
PCIE-8
Annual
Cost
($1000)
0
60
80
230
280
430
580
680
Consumer
Costs
(mills/ kWh)
0.0000
0.0085
0.011
0.033
0.040
0.061
0.083
0.097
Present
Worth
($Millions)
0.000
0.709
0.945
2.716
3.307
5.078
6.850
8.031
Annual Per
Person-Rem
97.5
67.5
31.3
11.4
5.5
3.5
1.5
1.4
GWE
Health
Risk
5.5(-3)b
3. 8 (-3)
1.8 (-3)
6. 5 (-4)
3.1 (-4)
2.0(-4)
4.0 (-5)
3. 8 (-5)
30-YR Total
Person-Rem
2.3(6)
1.6(6)
7.5(5)
2.7(5)
1.3(5)
8.4(4)
3.5(4)
3.2(4)
:A11 Reactors
Health
Risk
1.3(2)
9.1(1)
4.2(1)
1.6(1)
7.3(0)
4.7(0)
1.9(0)
1.8(0)
ADollars
AHealth
Risk
1.4(7)
3.8(6)
5.4(7)
5.7(7)
5.4(8)
5.1(8)
7.9(9)
AHealth
Risk
ADollars
7.2(-B)
2.6(-7)
1.9 (-8)
1.8 (-8)
1.9(-9)
1.9(-9)
1.3 (-10)
a Dollars are present worth dollars and health risks are 30 year total; each is placed on a per GWe basis.
H> t>An expression such as 5.5(-3) is equivalent to 5.5xlO~3, or 0.0055.
-------
K
cr>
System
Designation
PGIO-1
PGIO-2
PGIO-3
PCIO-4
PGIO-5
PGIO-6
PGIO-7
PGIO-8
Dollar
Annual
Cost
($1000)
0
60
80
130
280
430
580
680
Costs Per GWE
Consumer
Cost
(mills/ kWh)
0.0000
0.0085
0.011
0.019
0.040
0.061
0.083
0.097
TABLE 60
PWR Radioiodine Systems: Cost-Effectiveness
(Organic Form)
Environmental Costs
Present
Worth
($Millions)
0.000
0.709
0.945
1.535
3.307
5.078
6.850
8.031
Annual Per GWE
Person-Rem
105.
94.5
67.0
44.7
6.8
3.1
3.0
3.0
Health
Risk
6.0(-3)b
5. 5 (-3)
3.8(-3)
2.5(-3)
3. 8 (-4)
1.7 (-4)
1.7(-4)
1.7 (-4)
30-YR Total:
Person-Rem
2.5(6)
2.3(6)
1.6(6)
1.1(6)
1.6(5)
7.3(4)
7.1(4)
7.1(4)
:A11 Reactors
Health
Risk
1.4(2)
1.3(2)
9.0(1)
6.0(1)
9.1(0)
4.1(0)
4.0(0)
4.0(0)
Cost Effectiveness a
^Dollars
Mtealth
Risk
4.7(7)
4.5(6)
1.6(7)
2.8(7)
2.8(8)
1.2(10)
«c
AHealth
Risk
ADollars
2.1 (-8)
2.2(-7)
6.4(-8)
3.6(-8)
3.6(-9)
8. 5 (-11)
0
^Dollars are present worth dollars and health risks are 30 year total; each is placed on a per GWe basis.
bAn expression such as 6.0(r3) is equivalent to 6.0xlO~ , or 0.006.
'Because PGIO-8 achieves the same level of health risk as PGIO-7 but at greater cost.
-------
TABLE 61
Liquid Systems: Cost-Effectiveness
Environmental Costs
Dollar Costs Per GWE
System
Designation
BWR Systems
BWR-1
BWR-2
BWR-3
BWR-4
PWR Systems
PWR-1
PWR-2
PWR-3
PWR-4
Annual
Cost
($1000)
26
61
140
440
90
200
280
390
Consumer
Cost
(milla/kWh)
0.0037
0.0087
0.020
0.063
0.013
. 0.029
0.040
0.056
Present
Worth
($Mllllons)
0.307
0.720
1.65
5.19
1.06
2.36
3.30
4.60
Annual Per GWE
Whole Body
Person-Real
1,650
4.5
2.1
1.9(-2)
6,500.
90.
24.
6. 5 (-3)
Thyroid
Peraon-Rem
8,000.
80.
37.
2.9
12,500.
550.
6.0
0.9
Total
Health Risk
1.6(0)b
7.o(-3)
3. 5 (-3)
1.7 (-4)
5.3(0)
9.4C-2)
1.7(-2)
5. 5 (-5)
30 YR Total
Whole Body
Person-Rem
2.0(7)
5.4(4)
2.5(4)
2.3(2)
1.6(8)
2.2(6)
5.8(5)
1.6(2)
: All Reactors
Thyroid
Person-Rein
9.6(7)
9.6(5)
4.4(5)
3.5(4)
3.0(8)
1.3(7)
1.4(5)
2.2(4)
Total
Health Risk
1.9(4)
9.2(1)
4.4(1)
2.0(0)
1.3(5)
2.3(3)
4.1(2)
1.3(0)
Cost-Effectiveness
ADollars
AHealth
Risk
8.6(4)
7.8(6)
3.4(7)
8.3(3)
4.1(5)
2.6(6)
AHealth
Risk
ADollars
1.2(-5)
1.3(-7)
3.0(-8)
1.2(-4)
2.4(-6)
3.9(-7)
"Dollars are present worth dollars and health risks are 30 year total; each is placed on a per GWe basis.
bAn expression such as 1.6(0) is equivalent to 1.6x10 , or 1.6.
-------
Table 62
Cost of Producing Electric Power
Cost (mills/kWh)
Costs Allocated to Range U.S. Average
(Consumption)
1968 Actual
Power Production 6.81-9.28 7.75
Fuel (Included in above) 2.47-3.04 2.47
Transmission 1.56-2.26 1.98
Distribution 4.46-7.71 5.69
Total 12.71-19.25 15.42
1990 Projected (1968 dollars)
Power Production 10.02-12.09 10.83
Fuel (Included in above) 2.67-3.27 2.86
Transmission 2.11-3.96 2.99
Distribution 2.71-7.03 4.43
Total 15.00-23.08 18.25
Source: Federal Power Commission, "The 1970 National Power Survey,
Part I" Table 19.11 page 1-19-10 (December 1971).
158
-------
Table 63
Typical Charges for Electric Power
Consumption
,- a - TT Cost (mills/kWh) (1969)
lype or Use Ranee A
,, , Range Average
(based on consumption rate)
Industrial 17.2-21.2 19>2
Commercial 23.7-35.8 29.6
Residential 18.0-40.5 29.7
Source: Federal Power Commission, "Typical Electric Bills - 1969"
F.P.C. Washington, D.C. (November 1969).
159
-------
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*U.S.GOVSRNMENT PRINTING OP«Ce:l»73 9M-JU/XT6 »-3
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