EPA-520/9-73-003-C
 ENVIRONMENTAL ANALYSIS
  OF THE URANIUM FUEL CYCLE
  i
  S-SSSS
  1
PART II - Nuclear Power Reactors
                  l*"""^
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
                                  vii

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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.
                                                                        13

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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

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                                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

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 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

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                               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

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                             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

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 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

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  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) (369)                         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 HlduP 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

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                                                              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 Cl21,22,38).




 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

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       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

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     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

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 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(xtyz)  =  airborne concentration at point x.y.z
          -   (x/Q')Q
                                                                        107

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      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

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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

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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

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 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

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 (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

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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

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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

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                                        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

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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

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 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

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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

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        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

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                       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)

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                                      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

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                                     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

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   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

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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

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                                      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

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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
luHul>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 Rdio*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)



  = "i 

SoufoTtrm



niysteri Holdup <15Dt
ftiyifcd Hctdup (30OI
ntysfcil HoUup (460)
Pliyjicri HoHup WOO?
Ctya. MM. or M. Abm. IKE DF-1000)
Ova. DM. orM.AbiR.  3
SJAE Ctar. Ad. h> Ctan 4

OMB
UVMM



- 1A-15
1A-3O
- 1A-46
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IB-IS
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. IMS
1MO
M
2*
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4
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58
K
SO

EctiNMHd
CvHri
CMt


3ftO.OOO
540,000

flOOjOOO
BOOjOOO
000*000
700,000
800X100
SlfBOOtOOO
tijooooo
tlJBOOOO
$2jOOOXX)0
fi^asoxm
d^OO^MO
(2.780000
S3jBOOjOOO
.
EMinuMrf
Amwri
CM


$ 6O.OOO
f 90AOO
$120,000
$1BOjOOO
$160X100
$19000
$2304)00
(290,000
SBOO^OOO
$600A
S690AOO
tfS*),000
$210,000
$790000
$10000
$730.000



tnco.it
1.3
O.14
O.04
O.02
O.02
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0.04
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0.02
0.02
0.02
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0.14
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6.1
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0.10
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MfKtary
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15.4
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0.4
0.3
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1.7
0.4
0.3
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0.2
0.2
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1.7
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126.
16.6
3.8
2.2
2.0
15.4
34
2.2
2.0
2.1
1.
1.8
0.6
15.3
14
1.6
0.4
puUtionl
nt Rat*
ram/yr)

10.2
1.2
0.3
0.2
0.2
1.2
0.3
0.2
0.2
0.2
0.2
0.2
0.1
1.2
0.1
0.1
<0.1
torn


Anraf.
SIM*
89.1
18.3
10.8
94
9.7
18.2
104
94
9.7
1.6
1.6
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18.2
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1.3
03




0.0108
04012
0.0003
0.0002
0.0002
0.0012
O.OOO3
0.0002
0.0002
0.0002
0.0002
0.0002
0.0001
0.0012
0.0002
0.0001

-------
       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
SOUK* 0 mm 3V~
Ctoan Staam Phia Ctaai 1 <. '
Claan Saum Ptu dm 7-10
Own SWam Phu ClM 240
CtoanStaamlttHCIaBlA
CtaaSaa.TiP1u.Ciaai4
Ctoaa
liyaMm
4ivi0Mtionl
o
..1
2-1O
2-20
2-40
2-o
3A
38
4
5A
68
SC
so
SE
EittauMd
Capital
Coat
S3.2OOVOOO
J8.500.000
SS.8QO.OOO
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S7.0OO.OOO
$8,000.000
$7,000,000
tijaooMo
12,6OOWO
7J0,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 lndiidul 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
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0.1
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323.
13
11.8
43
43
4.2
6.0
4.4
4.2
61
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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
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1.2
1.1
1.1
1.5
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1.1
1.4
1.9
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0.1
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214.
S.I
7.2
2.7
2.4
24
35
2.5
2.4
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5.0
0.3
1.4
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222.
5.3
11.6
4.8
4.3
4.3.
6.0
4.4
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4.0
7.9
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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
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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 A40f *
UpgrK* "> DwP Bd Cturcon AdtOttMr *.> Bkff *
CtraflStMMi V*lMt>2.in Duwn.mr
Ttirton* Gttd $Ml Cten SMWn
Nan*
ConUwWkBnt Krtlntv
A* CfKiw Cht>co*y AAorto*
Awiifwrv iW|. Ch.vc.Ml A4MrlMrb
Up9rd to OM0 8-Jd ClMwCOfl AdMrter , AuKdwy Bwtdinf1'
Cttw. SlMffi: Vl. >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* Eu*>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 ad 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.

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                          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.

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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.

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                                                                       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

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                              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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                            REFERENCES

1,   Draft  Environmental Statement Concerning Proposed Rule Making
    Action:   Numerical Guides for Design Objectives and Limiting
    Conditions  for Operation to Meet the Criterion "As Low as
    Practicable"  for Radioactive Material in Light-Water-Cooled Nuclear
    Power  Reactor Effluents, Prepared by the Directorate of Regulatory
    Standards,  U. S. Atomic Energy Commission, January, 1973.

2.   Portland General Electric Company, Trojan Nuclear Plant Final
    Safety Analysis Report. USAEC Docket No. 50-344, Public Document
    Room,  Washington, D.  C.

3.   Metropolitan Edison Company  and  Jersey  Central Power and Light
    Company, Final  Safety Analysis Report Three Mile Island Nuclear
    Station - Unit  1.  USAEC Docket No. 50-289, Public Document  Room,
    Washington, D.  C.

4.  Northern States Power Company, NSP Prairie Island Nuclear Generating
    Plant Units  1 and 2 Environmental Report Supplement Number  1. USAEC
    Docket Nos. 50-282 and 50-306, Public Document  Room, Washington,  D.C.

5.  "Decontamination Test Data for AMF Beaird Radioactive  Waste Evaporator
    Installed  at Rochester Gas & Electric Robert Ginna Nuclear Generating
    Station,"  AMF Bulletin RW-5, AMF Beaird, Inc.,  Shreveport,  Louisiana.

6.  Carolina Power and Light Company, H. B. Robinson Unit No.  2 Routine
    Operating  Report No. A for Period April 1, 1972 through September 30.
     1972, USAEC  Docket No. 50-261, Public Document Room, Washington, D. C.

7.  Baltimore  Gas and Electric Company, Calvert Cliffs Nuclear Power
    Plant Units  1 and 2  Final Safety Analysis Report. USAEC Docket
    Nos.  50-317  and 50-318, Public Document Room, Washington, D. C

8.   Carolina Power and Light Company, H. B. Robinson Steam Electric
     Plant Unit No. 2 Environmental  Report, USAEC Docket No. 50-261,
     Public Document Room,  Washington, D. C.

 9.   Kahn, Bernd, et.  al.,  Radiological  Surveillance Studies at a Boiling
     Water Nuclear Power Reactor. BRH/DER 70-1, Environmental Protection
     Agency, Radiation Office,  Washington,  D.  C.

10.   The Projected Radiological Impact of Primary to Secondary Leakage in
     a Pressurized Water Reactor (unpublished draft) Environmental Pro-
     tection Agency,  Office of Radiation Programs,  Technology Assessment
     Division.

11.  Eliason,  L. and C. Ross (Northern States Power Company),  M.Lagache
     and J. Vance (Bechtel-Corporation). Performance Evaluation of a BWR
     Plant  Liquid and Solid Radwaste System.
                                                                     161

-------
     12.   Virginia  Electric  and Power Company, North Anna Power Station
          Units  3 and  4 Preliminary  Safety Analysis Report. USAEC Docket- Nos
          50-404 and 50-405, Public  Document Room, Washington, D. C.

     13.   Tennessee Valley Authority, Draft Environmental Statement Supple-
          ments  and Additions. Browns Ferry Nuclear Plant Units 1, 2 and"!
          USAEC  Docket Nos.  50-259,  50-260, and 50-^96, Public Document	'
          Room,  Washington, D. C.

     14.   Virginia  Electric and Power Company, North Anna Power Station Units
          1. 2.  3.  and 4 Applicant's Environmental Report. USAEC Dockets
          50-338, 50-339,  50-404, and 50-405,  Public Document Room  Washing-
          ton, D. C.

     15.  Wisconsin Michigan Power Company,  Supplement to Applicant's
         Environmental Report, Point Beach 1,  2   USAEC Docket Nos.  50-266
         and 50-301,  Public Document Room,  U.  S.  Atomic Energy Commission
         Washington,  D.  C.                                                '

     16.  Florida Power Corporation,  Crystal River Unit 3 Nuclear Generating
         riant Final  Safety Analysis Report.  USAEC Docket No.  50-302  Public
         Document Room, Washington,  D.  C.

    17.  Omaha Public  Power District, Fort  Calhoun Station Unit  No.  1  Final
         Safety Analysis  Report.  USAEC  Docket  No.  50-285,  Public  Document
         Room,  Washington,  D.  C.

    18'  Final Environmental Statement  Related to  Operation of Surrv Power
         Station Unit  1 Virginia  Electric and  Power Company Docket No.	
         50-280  May 1972  United States  Atomic  Energy Commission Directorate
         of  Licensing.

    19.   Philadelphia Electric Company, Supplement No.  1 Applicant's Environ-
         mental  Report Operating  License^t-Tge Peach Bottom Atomic Po^r"	
         Station Units 2  & 3, TTSAFr  iw.i>f- *  sn o?7  ^  ^  o78  rubllc
         Document Room, Washington,  D.  C.

   20.   Baltimore  Gas and Electric  Company, Supplement to  Environmental
         Report  Calvert Cliffs Nuclear Power Plant. USAEG Docket Una
         50-317  and 50-318, Public Document Room, Washington, D. C.

   21.   The Toledo Edison Company, Davis-Besse Nuclear Power Station Supple-
        ment to Environmental Report. USAEC Docket No. 5Q-346; PnhUr    	
        Document Room, Washington, D. C.

   22.  Indiana and Michigan Power Company, Indiana and Michigan Electric
        System,  American Electric Power System,  Donald C. Cook Nuclear
        Plant Supplement  to Environmental Report.  USAEC Docket Nos.  5Q-315
        and  50-316, Public  Document Room,  Washington,  D. C.
162

-------
23.  Commonwealth Edison Company, Before the State of Illinois
     Pollution Control Board.  Application of Commonwealth Edison
     Company for permits for Zion Units 1 and 2 at Zion Station.
     Supplement to Environmental Reportr USAEC Docket Nos.  50-295
     and 50-304, Public Document Room,  Washington, D.C.

24  Statement of Dr.  Walton A.  Rodger  on Behalf of the Consolidated
     Utility Group; Practicability of  Meeting Proposed Appendix I
     Numerical Values, Rule Making Hearing,  Effluents from Light
     Water-Cooled Nuclear Power  Reactors, Docket No.  RM-50-2, U. S.
     Atomic Energy Commission, March 17, 1973.

25   Comments of the General Electric Company on Proposed Appendix  I
     Draft Environmental Statement. Rule Making Hearing, Effluents
     from Light Water  Cooled Nuclear Power Reactors,  Docket No.  RM-
     50-2, U. S. Atomic Energy Commission, March, 1973.

26   Final Environmental Statement Related to the Proposed William  B.
     McGuire Nuclear Station Units 1 &  2 Docket Nos.  50-369 and  50-
     370, United States Atomic Energy Commission, Directorate of
     Licensing, October 1972,  Public Document Room, Washington,  D.C.

27'  Final Environmental Statement Related to Operation of Virgil C.
     Summer Nuclear Station Unit 1. Docket No. 50-395, United States
     Atomic Energy Commission, Directorate of Licensing, January 1973,
     Public Document Room, Washington,  D.C.

28  Final Environmental Statement Related to the Continuation  of
     Construction and  the Operation of  Units 1 & 2 and the Construc-
     tion of Units 3 & 4 North Anna Power Station. Docket Nos.  50-338
     and 50-339, Docket Nos. 50-404 and 50-405, United States Atomic
     Energy Commission Directorate of Licensing, April 1973, Public
     Document Room, Washington,  D.C.

29.  Michels, L. R., and N. R. Horton (General Electric Company),
     Improved BWR Offgas Systems." 12th AEG  Air Cleaning Conference

30.  Boston Edison Company, Pilgrim Nuclear  Power Station Final Safety
     Analysis Report,  Amendment  31, USAEC Docket No. 50-293,  Public
     Document Room, Washington,  D.C.

31.  Lederer, C. M., J. M. Hollander, and I. Perlman, Table of  Isotopes
     (Sixth Edition).  John Wiley & Sons, Inc., New YorkI'       

32.  Schroeder, H. J., H. Queiser, and  W. Reim, "Offgas Facility at the
     Gundremmingen Nuclear Power Plant," Kerntechnik. May 1971.
                                                                      163

-------
 33.  Supplement No. 1 to the Safety Evaluation by the Directorate of
      Licensing U.S. Atomic Energy Commission in the Matter of Boston
      Edison Company Pilgrim Nuclear Power Station, Docket No. 50-293,
      Public Document Room, Washington, D.C.   June 2, 1972.

 34.  Tennessee Valley Authority, Browns Ferry Nuclear Plant Units 1,
      2, & 3 Final Safety Analysis Report, USAEC Docket Nos. 50-259,
      50-260, & 50-296, Public Document Room, Washington,  D.C.

 35.  Draft Environmental Statement by the U.S.  Atomic Energy Commission
      Directorate of Licensing Related to the Proposed Construction of
      the Beaver Valley Power Station Unit 2  by the Cleveland Electric
      Illuminating Company, Duquesne Light Company, Ohio Edison Company,
      Pennsylvania Power Company, Toledo Edison Company, USAEC Docket
      No. 50-412, Public Document Room, Washington, D.C.  March 1973.

 36.  1000-MWe Central Station Power Plants Investment Cost Study:
      Volume I, Pressurized Water Reactor Plant,  WASH-1230 (Vol.  I),
      United Engineers and Constructors Inc., 1401 Arch Street,
      Philadelphia, Pennsylvania, 19108, Prepared for the  U.S. Atomic
      Energy Commission, Division of Reactor  Development and Technology,
      Contract No.  AT (30-1)-3032.

 37.  Martin, James E., F.  L.  Galpin, and Ted W.  Fowler, "Technology
      Assessment of Risk Reduction Effectiveness  of Waste  Treatment
      Systems for Light-Water Reactors," Radiation Data and Reports,
      January 1973, U.S. Environmental Protection Agency,  Office of
      Radiation Programs, Washington, D.C.

 38.  Tennessee Valley Authority, Draft Environmental Impact Statement,
      Sequoyah 1. _2_. USAEC Docket Nos.  50-327 and 50-328,  Public Docu-
      ment Room, Washington, D.C.

 39.  "New Cost Data for Installing Add-on Radwaste Control Equipment,"
      Nuclear Industry. May 1972.

 40.  Northern States Power Company,  Monticello Nuclear Generating Plant
      Environmental Report, USAEC Docket No.  50-263,  Public Document
      Room,  Washington, D.C.

 41.  Nucleonics Week.  Volume  12, No.  20,  May 24,  1971.

 42.  Battelle Columbus Laboratories  for Commonwealth Edison' Company,
      Environmental Impact  Report;  Supplemental Information to the Quad
      Cities Environmental  Reportt  USAEC Docket Nos.  50-254 and 50-265,
      Public Document Room,  Washington,  D.C.
164

-------
43.  Commonwealth Edison Company and Battelle Columbus Laboratories,
     Environmental Impact Report;  Supplemental  Information to the
     Dresden 3 Environmental Report. USAEC Docket No. 50-249, Public
     Document Room, Washington,  B.C.

44.  Washington Public Power Supply  System, Hanford Number Two 1100 MW
     Nuclear Power Plant Environmental Report.  USAEC Docket No. 50-397,
     Public Document Room, Washington, D.C.

45.  Keilholtz, G. W., "Krypton-Xenon  Removal Systems," Nuclear Safety,
     Vol. 12, No. 6, November-rDecember, 1971.

46.  Griffith, Gary (Babcock & Wilcox), "99% Cleanup of Nuclear Gaseous
     Wastes," Power Engineering. March 1973.

47.  Carolina Power and Light Company, Shearon  Harris Nuclear Power
     Plant Units 1. 2. 3. & 4 Preliminary Safety Analysis Report,
     USAEC Docket Nos. 50-400, 50-401, 50-402,  50-403,  Public Document
     Room, Washington, D.C.

48.  Final Environmental Statement Related to Operation of  Shoreham
     Nuclear Power Station. Long Island Lighting Company. Docket  No.
     50-322, U.S. Atomic Energy Commission,Directorate of  Licensing,
     September 1972, Public Document Room, Washington,  D.C.

49.  Evaluation of Data Regarding the  Efficiency of Activated  Charcoal
     Adsorbers at Low Gas Phase Iodine Concentration WASH-1231.  U.S.
     Atomic Energy Commission, Directorate of Licensing,  May 31,  1972.

50.  Topical Report on Imgregnated Activated Carbon (AAF  Type 2701)  for
     Removal of Radioiodine Compounds  from Reactor Containment
     Atmospheres, AAF-TR-7102, R. D. Rivers,  Editor,  September 1, 1972,
     American Air Filter Company, Inc., P. 0.  Box 1100, Louisville,
     Kentucky  40201.

51.  Kahn, B., et. al., Radiological Surveillance Studies at a Pressurized
     Water Nuclear PowerJieactor. RD 71-1. U.S. Envlrcmmgntal  Pf0m-tn
     Agency, Radiation Office, Washington, D.C.

52.  Operating History U.S. Nuclear Power Reactors. WASH-1203-72,
     Prepared by Division of Reactor Development and Technology, U.S.
     Atomic Energy Commission, 1972.

53.  U.S. Atomic  Energy Commission, The Potential Radiological
     Implications of Nuclear Facilities in a Large Region in the U.S
     in  the Year  2000. USAEC Report WASH-12Q9  fnt-af-) n~+^T 1972 
                                                                      165

-------
 54.  Fletcher, J. L. and W. L. Dotson, HERMES - A Digital Computer
      Code for Estimating Regional Radiological Effects from the
      Nuclear Power industry, USAEC Report HEDL-TME-71-168 December
      1971.

 55.  Lowder, W. M., P. D. Raft, and C. V. Gogolak, Environmental Gamma
      Radiation from Nitrogen-16 Decay in the Turbines of a Large Boiling
      Water Reactor. USAEC Report HASL-TM-72-1 (February 1972).

 56.  Cooper, R. E., "EGAD - A Computer Program to Compute Dose Integrals
      from External Gamma Emitters," USAEC Report DP-1304 (1972).

 57.  Draft Environmental Statement by the Directorate of Licensing United
      States Atomic Energy Commission Related to Operation of San Onofre
      Nuclear Generating Station Unit 1.  Southern California Edison
      Company and San Diego Gas & Electric Company. Docket No.  50-206.
      Issued May 1973,  Public Document Room,  Washington,  D.C.

 58.  Draft Environmental Statement by the United States  Atomic Energy
      Commission Directorate of Licensing Related to the  Susquehanna Steam
      Electric Station Units 1 and 2 of the Pennsylvania  Power  and Light
      Company Docket  Nos. 50-387 and 50-388.  Issued January 1973, Public
      Document Room,  Washington, D.C.

 59  Draft Environmental Statement by the Directorate of Licensing United
      States  Atomic Energy Commission Related to  the Limerick Generating
      Station Units 1 and 2.Philadelphia  Electric Company.Docket Nos.  50-352
      and 50-353.  Issued December 1972,Public Document Room,  Washington, D.C.

 60.  Peterson, H.T.  and J. M. Smith,  "Guides for Predicting Thyroid Doses
      from Radioiodine Releases" in Environmental Surveillance  in the
      Vicinity of Nuclear Facilities (W.  D. Reinig, ed) C. Thomas and Son
      Springfield, Illinois (1970) Chapter 23.

 61.  Zimbrick, J.D.  and P. G. Voilleque, Controlled Environmental Radioiodine
      Tests.  Progress Report Number 4. U.S.A.E.G. Report  IDO-12065 (January
      1969) [methyliodide v  < 0.05% elemental iodine v ].
                           O  '                         
 62.  Atkins, DvH.F., R. C. Chadwick and  A. C. Chamberlain,  "Deposition of
      Radioactive Methy Iodide to Vegetation", Health Physics.  13_: 41-92
      (1967)  [methyliodide - 8.6 x 10~5 cm/sec],

 63.  1000 MWe Central  Station Power Plants Investment Cost Study; Volume II,
      Boiling Water Reactor Plant. WASH-1230  (Vol II), United Engineers and
      Constructors Inc., 1401 Arch Street, Philadelphia,  Pennsylvania, 19108,
      Prepared for the  United States Atomic Energy Commission,  Division of
      Reactor Development and Technology, Contract no. AT (30-1)-3032
166

-------
64.  Age Dependence of the Biological Half-life of Iodine and Cesium
     in Man - Mary Jane Cook and W.  S.  Snyder  (ORNL annual report).

65.  ICRP Publication 2 - Report of  Committee  II on Permissible Doses
     for Internal Radiation, International  Commission  on Radiological
     Protection, Pergamon Press, New York (1959).

66.  Rohwer, P. S., K. E. Cowser,  W. S. Snyder, and E.  6, Struxness,
     "An Approach for Assessing Potential Population Exposures",
     Environmental Surveillance in the  Vicinity of Nuclear Facilities,
     Proceedings of a Symposium Sponsored by the Health Physics Society
     January 24 to 26, 1968 (ed. by  W.  E. Reinig), Charles C. Thomas  (1970).

67.  Bryant, Pamela M., Data for Assessments Concerning Controlled and
     Accidental Releases of 131i ^4 137Cs  to Atmosphere, Health
     Physics, Vol. 17, pp. 51-57  (July  1969).

68.  Hoffman, F. 0., Parameters to be Considered  when Calculating the
     Age-Dependent iBlj Dose to the thyroid, Inatitut Fur  Reaktorsicherheit,
     IRS-W-5  (April 1973).

69.  Meek, M. E. and R.  S. Gilbert,  Summary of Gamma and Beta Energy and
     Intensity Data, General Electric,  Nucleonics Laboratory, Pleasanton,
     California, NEDO-12037, January (1970).

70.  Snyder, W.  S., Mary R. Ford, G. G. Warner, and H. L.  Fisher, Jr.,
     Estimates of  Absorbed Fractions for Monoenergetic Photon Sources
     Uniformly Distributed in Various  Organs of a Heterogeneous Phanton,
     Medical  Internal Radiation Dose Committee, Journal of Nuclear
     Medicine/Supplement Number 3,  Pamphlet 5, P. 43, August (1969).

71.  Ellett,  W.  H. and  R. M. Humes, Absorbed Fractions for Small
     Volumes  Containing Photon-Emitting  Radioactivity, Medical
     Internal Radiation Dose Committee,  Journal of Nuclear
     Medicine/Supplement Number 5,  Pamphlet 8, p. 30, March  (1971).

 72.  Garner,  R.  J., and R.  Scott  Russell,  "Isotopes of Iodine", Chap.  14
     from Radioactivity and Human Diet (1966).

 73.  Garner,  R. J. "A Mathematical  Analysis  of the Transfer  of Fission
     Products to Cow's Milk",  Health Physics  . 13;205-212  (1967).

 74.   "Consumption of Selected  Food  Items in U.S. Households, July 1962",
      Radiological Health Data. A;124-129 (March. 1963).  ;

 75.   Radiological Health Handbook.  U.S.  PHEW. U.S. PHS.  (January, 1970).
                                                                      167

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    76.  Hoffman, F. 0., "Environmental Variables Involved with the
        Estimation of the Amount of 1311 in Milk and the Subsequent
        Dose  to the Thyroid," Instltut Fur Reaktorsicherheit, IRS-W-6
        (June 1973).

    77.  "Food Consumption of Households in the United States" U.S.
        Dept. of Agriculture, Household Food Consumption Survey 1965=66,
        Report #12.

    78.  "Vegetables - Fresh Market"  1972 Annual Summary Vg 2-2(72),
        U.S.  Dept. of Agriculture, Dec., 1972.

    79.  A. M. Freke, "A Model for the Approximate Calculations of Safe
        Rates of Discharge of Radioactive Wastes into Marine Environments,"
        Health Physics. Vol. 13, pp. 743-758, (1967).

    80.  W. H. Chapman, H. L. Fisher, and ,M. W. Pratt, "Concentration Factors
        of Chemical Elements in Edible Aquatic Organisms," Lawrence Radiation
        Laboratory, University of California, UCRL-50564, (Dec.  1968).

    81.  Martin, J. A., C. B. Nelson, and P.A. Cuny.  AIREM Program Manual-
        A Computer Code for calculating doses and depositions due to
        emissions of radionuclides.  (In Press) Office of Radiation Programs,
        U.S.  Environmental Protection Agency, Washington, D.C.  20460.

    82.  Holzworth, George C.  Mixing heights, wind speeds and potential for
        urban air pollution throughout the contiguous United States.  U.S.
        Environmental Protection Agency, Office of Air Programs, Research
        Triangle Park, North Carolina 27711, (January 1972).

    83.  Turner, D. Bruce.  Workbook of atmospheric dispersion estimates,
        Office of Air Programs Publication No. AP-26, U.S. Environmental
        Protection Agency, Research Triangle Park, North Carolina 27711, (1970)

   84.  Slade, D.  H.,  editor.   Meteorology and atomic energy, 1968, TID-
        24190.  National Technical Information Service, U.S. Department
        of Commerce,  Springfield, Va.   22151 (July 1968).

   85.  Loevinger,  R.,  "Distributed Radionuclide Sources" Chapter 18 of
        Radiation  Dosimetry (F.  H.  Attix and E.  Tochilin, eds.)  New York:
        Academic Press  (1969),  pp.  51-90.

   86.  Adams, N.,  Appendix I  to U.K.A.E.A.  Report AHSB(RP)-R101,  1969.

   87.  Russell, J.L. and  F. L.  Galpin,  "Comparison of  Techniques  for
        Calculating Doses  to the Whole Body and  to the  Lungs from
        Radioactive Noble  Gases^ Paper presented at the Mid-year topical
        symposium of the Health  Physics  Society  at Richland,  Washington,
        November 1971.
168

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88.  United Nations  Scientific Committee  on the Effects of Atomic
     Radiation.   Ionizing Radiation: Levels and Effects, Vol.  II:
     Effects, United Nations Publication  E. 72. IX.18, United  Nations,
     New York  (1972) pp.87-105.
       *U.S.GOVSRNMENT PRINTING OPCe:l73 9M-JU/XT6 -3

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