&EPA
           United States
           Environmental Protectio.
           Agency
           Radiation
           Office of
           Radiation Programs
           Washington DC 20460
EPA 520/1-82-025
December 1982
Draft
Environmental
Impact Statement
for40CFR191:
           Environmental Standards
           for Management and
           Disposal of Spent Nuclear
           Fuel, High-Level and
           Transuranic Radioactive Wastes

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                                    EPA  520/1-82-025
                DRAFT




   ENVIRONMENTAL IMPACT STATEMENT
           40 CFR Part 191
       ENVIRONMENTAL STANDARDS




                 FOR




       MANAGEMENT AND DISPOSAL




                 OF




 SPENT NUCLEAR FUEL, HIGH-LEVEL AND




   TRANSURANIC RADIOACTIVE WASTES
            DECEMBER 1982
U. S. ENVIRONMENTAL PROTECTION AGENCY




     Office of Radiation Programs

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                                  SUMMARY

                   DRAFT  ENVIRONMENTAL  IMPACT  STATEMENT

40 CFR Part lyi    Environmental Standards for Management and Disposal of
                   Spent Nuclear Fuel,  High-Level and Transuranic
                   Radioactive Wastes.

                               Prepared By:
                       Office of Radiation Programs
                   U.S. Environmental Protection Agency
                               December 1982

1.  The Environmental Protection Agency is proposing environmental
standards for the management and disposal of spent nuclear reactor fuel
and high-level and transuranic radioactive wastes.  Subpart A of the
standards would limit the radiation exposure of members of the public from
management and storage of spent fuel and of waste prior to disposal.
Subpart B would establish both quantitative containment requirements for
disposal systems and qualitative requirements to assure that these
containment requirements will be met.  The containment requirements would
limit the amount of radioactivity that  may enter the environment for
10,000 years after disposal.  The assurance requirements provide seven
principles necessary for developing confidence that these long-term
containment requirements will be complied with.  These principles call for
well-designed, multiple-barrier disposal systems that would not rely upon
future generations for maintenance and  would not be located near
potentially valuable resources.  They also require that future generations
be provided information about the location and dangers of the wastes and
an option to recover the wastes if they need to.  In addition, Subpart B
contains procedural requirements to ensure that the containment
requirements are properly applied.  We  think the proposed standards will
adequately protect the public health and the environment, and we believe
they can be satisfied without major economic consequences.

2.  Copies of this Environmental Impact Statement and requests for comment
have been sent to the following Federal agencies:

     Department of Commerce
     Department of Defense
     Department of Energy
     Department of Health and Human Services
     Department of the Interior
     Department of Transportation
     National Aeronautics and Space Administration
     Nuclear Regulatory Commission

     We have also sent copies to all State Clearinghouses and to other
individuals and organizations who have  notified us of their interest.

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3.  Comments on this Environmental Impact  Statement  should  be  received  by
May 2, 1983.  Please send comments (in duplicate  if  possible)  to:

     Central Docket Section (A-130)
     Environmental Protection Agency
     Attn:  Docket No.  R-82-3
     Washington,  D.C.  20460

4.  For additional information,  please contact Dan Egan at  (703)  557-8610
or write to:

     Director, Criteria and Standards Division
     Office of Radiation Programs (ANR-460)
     Environmental Protection Agency
     Washington,  D.C.  20460
                                    11

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                                 CONTENTS

                                                                  Page

Chapter 1:  Overview                                                 1

Chapter 2:  The Proposed Action                                      7

     2.1  High Level Wastes                                          7

         2.1.1  Commercial Reactor Wastes                             8
         2.1.2  Defense Wastes                                       11
         2.1.3  Definition of High-Level Wastes  in 40 CFR 191        13

     2.2  Transuranic Wastes                                        16

     2.3  Biological Effects of Radiation                            17

     2.4  Methods of Control                                        19

         2.4.1  Storage of Spent Fuel and  High Level Wastes          20
         2.4.2  Disposal in Mined Repositories                      21
         2.4.3  Other Disposal  Methods                              21

     2.5  Authorities                                               22

     2.6  The Proposed Standards                                    24

         2.6.1  Standards for Management and Storage  (Subpart A)     24
         2.6.2  Standards for Disposal (Subpart  B)                   27

     2.7  Implementation                                            34

Chapter 3:  Issues in Setting the Disposal Standards                 37

     3.1  Risks to Future Generations                               37

         3.1.1  Perspectives on Long Term  Risks                      39
         3.1.2  Risks Associated with Natural Background Radiation  40
         3.1.3  Risks from Uranium Ore Bodies                       43
         3.1.4  Risks from Nuclear Power Generation                 44
         3.1.5  Risks from Nuclear Weapons Fallout                   45

     3.2  Assessment and Reduction of Risks                         48

         3.2.1  Assessment of Risks from Waste Disposal             48
         3.2.2  Reduction of Risks by Disposal Technology           49
                                    ill

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                                                                   Page

     3.3  Dealing with Uncertainties                                 54

         3.3.1  Technical Uncertainties                              54
         3.3.2  Uncertainties in Human Behavior                      57

     3.4  Choice of Format for Disposal Standard                      62

     3.5  Alternative Regulatory Time Periods                        64

Chapter 4:  Alternatives                                             69

     4.1  Issue No Standard                                          70

     4.2  Delay Action                                               71

     4.3  Establish Only Qualitative Requirements                    72

     4.4  Delete or Deemphasize the Qualitative Assurance
          Requirements                                               74

     4.5  Select Containment Requirements on a Different Basis        76

     4.6  Set Higher or Lower Release Limits                         78

     4.7  Set Disposal Standards in Terms of Limits on
          Maximum Individual Exposure                                81

     4.8  Set Different Limits for Releases Due to Natural Causes
          and for Those Caused by People                             83

Chapter 5:  Projected Health Effects From Disposal                   87

     5.1  Model Repositories                                         88

     5.2  Release and Transport Mechanisms                           94

     5.3  Results of Risk Assessments                                97

         5.3.1  Population Risks                                     97
         5.3.2  Risks to Individuals                                105

     5.4  Conclusions                                               107
                                    IV

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                                                                   Page

Chapter 6:  Proposed Standards for Waste Disposal                   111

     6.1  Containment Requirements                                  112

         6.1.1  The Accessible Environment                          113
         6.1.2  Radionuclide Release Limits                         115
         6.1.3  The Level of Protection                             118

     6.2  Assurance Requirements                                    120

         6.2.1  Criterion 1                                         120
         6.2.2  Criterion 2                                         121
         6.2.3  Criterion 3                                         122
         6.2.4  Criterion 4                                         123
         6.2.5  Criterion 5                                         124
         6.2.6  Criterion 6                                         125
         6.2.7  Criterion 7                                         127

     6.3  Procedural Requirements                                   128

Chapter 7:  Proposed Standards for Waste Management Operations      131

     7.1  Waste Management Operations for Various
          Disposal Options                                          131

         7.1.1  Operations for Disposal of Spent Fuel               131
         7.1.2  Operations Prior to Disposal of Processed
                High-Level Wastes                                   134
         7.1.3  Collection and Disposal of Krypton-85 and
                Iodine-129 Wastes                                   139
         7.1.4  Extraterrestrial Disposal                           141
         7.1.5  Transmutation                                       142
         7.1.6  Other Separations                                   142

     7.2  Derivation of the Standards                               143

         7.2.1  Selection of Standards Format                       144
         7.2.2  Selection of Numerical Limits                       145
         7.2.3  Implications of the Operational Standards           148

Chapter 8:  Environmental Impacts                                   151

Chapter 9:  Regulatory Impact                                       157

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                                                                   Page





Appendix A:  The Proposed Standards                                 167




Appendix B:  Risk Assessments of Geologic Repositories               183




     B.I  General Features of the Model                             186




     B.2  Risk Assessments for Populations                          194




     B.3  Dose Assessment for Individuals                           212




References                                           '               223
                                    VI

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                                  TABLES

                                                                  Page

2-1  Defense High-Level  Waste  Projections                            12

2-2  Concentrations Identifying  High-Level  Radioactive Wastes
         (Table 1 in 40  CFR Part 191)                                14

3-1  Distribution of Natural Radiation Dose Equivalents              42

3-2  Dose Commitments from Fallout                                  46

5-1  Potential Health Effects  (Fatal Cancers)  Caused  per  Curie
         Released to the Environment by Different  Modes              96

7-1  Estimated Dose Equivalents  to the Maximally Exposed
         Individual from Solidification (Rural Site)                 138

7-2  Dose Equivalents to Maximum Individual from Solidification
         (Urban Site)                                               138

7-3  Population Doses from Waste Solidification                     138

B-l  Characteristics of Radioactive Waste (Spent Fuel)               188

B-2  Solubility Limits and Retardation Factors                      189

B-3  Aquifer Characteristics                                        193

B-4  Reference Case Characteristics                                 199

B-5  Projected Population Risks Over 10,000 Years:
         Reference Cases                                            205

B-6  Projected Population Risks Over 10,000 Years:
         Various Canister Lives and Waste Form Release Rates        207

B-7  Projected Population Risks Over 10,000 Years:
         Various Retardation Factors and Solubility Limits          209

B-8  Dose Equivalent Rates  (rem/yr) from Drinking Grouadwater
         Contaminated by Normal Groundwater Flow Through a
         Basalt Repository                                          219
                                    Vll

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                                  FIGURES

                                                                   Page

5-1  Reference Repository in Bedded Salt                             89

5-2  Reference Repository in Granite                                 90

5-3  Consequences and Risks for Different Events
         (Bedded Salt and Granite)                                   99

5-4  Projected Health Effects Over 10,000 Years
         for Reference Repositories in Different Geologic Media     100

5-5  Projected Health Effects Over 10,000 Years
         vs. Different Waste Form Release Rates                     102

5-6  Projected Health Effects Over 10,000 Years
         with Different Assumptions About Geochetnical Factors       103

5-7  Projected Health Effects Over 10,000 Years
         vs. Different Canister Lifetimes                           104

6-1  The Level of Protection                                        119

9-1  Variations in Waste Management Cost vs.  Level of Protection
         (Engineering Barrier Costs Only)                           160

9-2  Variations in Waste Management Cost vs.  Level of Protection
         (Engineering Barrier Costs and Site  Selection Costs)       162

B-l  Reference Repository in Basalt                                 192

B-2  Probability of Population Risks Over 10,000 Years              197
                                   vni

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                                 Chapter 1




                                 OVERVIEW









     Although high-level radioactive wastes are produced in small




quantities,  their proper management and disposal are important because of




the  inherent hazards of the large amounts of radioactivity they contain.




This need for careful control has been recognized since the inception of




the nuclear  age.  The Federal Government has always assumed responsibility




for  the  ultimate care and disposal of high-level wastes, whether they are




produced by  commercial or national defense activities.  Over the last




several  years,  the Federal Government has intensified its program to




develop  and  demonstrate a permanent disposal method for high-level wastes.




President Reagan's April 28,  1982, message to Congress on nuclear waste




disposal reaffirmed  this commitment and called  for a Federally owned and




operated permanent repository to be available at the earliest practicable




date.  The environmental protection standards described in  this Draft




• Environmental. Impact Statement will provide the basic framework for  the




long-term control of these wastes.









     The Agency is proposing  these generally  applicable environmental




standards under authorities established by the  Atomic Energy Act and




transferred  to  the EPA.by Reorganization Plan No.  3  of  1970.  This Draft




Environmental Impact Statement  includes detailed discussions  of major




decisions and extensive  summaries of  our technical analyses.  On




November 15, 1978, the Agency proposed  Federal  radiation  protection




guidance for the disposal of  all types  of  radioactive wastes  (43 FR  53262),

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After  reviewing  the public comments  on  that  proposal,  we  decided  that  the




characteristics  of different kinds of radioactive waste are  not




sufficiently  similar  for generally applicable criteria to be  appropriate.




Therefore, we stopped developing this Federal radiation protection




guidance  (46  FR  17567).  However, several of the principles  included in




this earlier  proposal have been incorporated as integral  parts of these




environmental standards.









     About 500 million curies of radionuclides with half-lives greater




tnan 20 years exist in the wastes from  reprocessing reactor  fuel  for




national  defense activities.  These wastes are stored  in  various  liquid




and solid forms on three Federal reservations in Idaho, Washington, and




South Carolina, respectively.  Relatively small additions are being made




from ongoing  defense programs.









     The  total amount of spent fuel removed from commercial nuclear power




reactors contains about 800 million curies of radionuclides with




half-lives greater than 20 years.   Over the next few years, this inventory




is expected to grow at a rate of about 200 million curies per year from




reactors currently licensed to operate.   Virtually all of this spent fuel




is stored at  reactor sites.  At some reactor sites,  spent-fuel storage




capacity is almost used up.  Electrical utilities,  the operators of




commercial reactors,  are pursuing a variety of techniques to  increase




storage capacities.

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     These proposed standards apply to spent reactor fuel, high-level




wastes derived from reprocessing spent fuel, and wastes containing




long-lived radionuclides of elements heavier than uranium (transuranic




wastes).  High-level wastes are covered by these standards if they contain




concentrations of radioactivity greater than the limits recently set by




the Nuclear Regulatory Commission (NRG) for acceptance at low-level waste




disposal sites (47 FR 57446).  Transuranic wastes are covered if they




contain 100 nanocuries (one ten millionth of a curie) or more of




alpha-emitting transuranic radionuclides per gram of waste.  The proposed




standards do not apply to wastes that have already been disposed of.









     The objective of the proposed standards is to limit the risks to both




present and future generations and to adequately protect the public from




harm caused by management and disposal activities related to these




radioactive wastes.  Separate standards were developed for those




activities related to waste management and storage operations preparatory




to disposal (Subpart A) and for the long-term performance of disposal




systems (Subpart B).  The standards for management and storage are




intended to protect exposed individuals while these operations are in




progress.  The standards for disposal are intended to assure long-term




isolation of the hazardous wastes from the biosphere and protection of the




public health.









     The rationale for the standards for management and storage (Subpart A)




is that these operations should not be permitted to substantially increase




the risk to people beyond that now accepted for the normal operations of

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 the uranium fuel  cycle.   We conclude that these added risks can be very




 small and that the exposure limits we have already established for the




 fuel cycle (40 CFR 190)  can be extended to cover these additional




 operations.   Therefore,  the effect of Subpart  A of the proposed standards




 is  to apply  the same requirements  to the normal operations of all parts of




 the uranium fuel  cycle,  except transportation.   Similar considerations




 apply to  the management  and storage of high-level  and transuranic wastes




 from national defense activities,  which are also covered by this Subpart.









      The  disposal  standards (Subpart B)  must deal  with a yet unproven




 technology and with  the  need to extend public  health  protection far into




 the future.   As our  primary disposal standards,  we are proposing




 containment  requirements  that  place numerical  limits  on possible releases




 of  radionuclides to  the  environment for  10,000  years  after disposal.




 Although  these requirements are not specific to  any particular method of




 disposal,  we  focused  on  the use of  mined  geological repositories because




 more  information is available  on this  technique.   We  performed detailed




 technical  evaluations  of  the probabilities  and  consequences  of possible




 events that  could  disrupt such  a repository  and  cause  release  of a part of




 its contents.  Our calculations show that  the  total adverse  impact of




 releases of  radionuclides from  such a  facility  over 10,000 years can  be




 kept  very  small and should  be no greater  than  the  risks from the unmined




 ore from which  the wastes were  derived.  The proposed  containment




 requirements—which would limit  long-term  risks  to these  low levels—apply




 to any method of disposal.  Thus, any  other  method would  have  to provide




at least as much protection as  geologic disposal.

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     Because of the uncertainties inherent in projecting disposal system




performance for 10,000 years, Subpart B also contains seven qualitative




criteria that are needed to develop appropriate confidence that our




containment requirements will be met.  These assurance requirements set




forth a cautious approach to disposal of these wastes, and they provide




the context necessary for application of our containment requirements.




The assurance requirements call for well-designed, multiple-barrier




disposal systems that would not rely upon perpetual maintenance by future




generations and that would be located where it is unlikely that they would




be disturbed by natural forces or human activities.  In addition, our




disposal standards contain procedural requirements to ensure that the




containment requirements are properly implemented.









     We evaluated the effects of setting our containment requirements at




different levels of protection.   We found that the increased costs of




setting these requirements at the proposed level could range from zero to




50 million (1981) dollars per year when compared to the costs of choosing




a level more than 10 times less stringent (release limits 10 times greater




than our proposed limits).  This potential increase is much less than the




uncertainty in the total costs for waste management and disposal, since




these projected costs range from about 700 million to almost 1.5 billion




(1981) dollars per year.  For comparison, electrical utility revenues were




about 100 billion dollars in 1980.  We estimate that the potential




economic impact of choosing the more stringent level of protection could

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be about a 0.2-percent  increase  in  the costs  of generating electricity




from nuclear power plants and a  much smaller  increase  (about 0.05 percent)




in average electricity  rates.









     The standards for  waste management and storage (Subpart A) will be




implemented by the NRC  for commercial nuclear power activities and by the




Department of Energy (DOE) for national defense facilities.  Implementation




procedures for Subpart A will be similar to those for our Uranium Fuel




Cycle standards (40 CFR 190).  Our standards for disposal (Subpart B) will




be implemented by NRC for all high-level wastes, whether the wastes come




from commercial or military activities.   NRC will develop the necessary




regulations (primarily  10 CFR 60) and will issue appropriate construction




and operating licenses to DOE.   DOE will select, design, and build all




disposal facilities for high-level wastes.  Under current law,  disposal of




transuranic wastes from military activities is not regulated by NRC;




therefore,  DOE  will apply our standards  and guides to disposal  of these




transuranic wastes.

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                                Chapter 2




                           THE PROPOSED ACTION









     This Draft Environmental Impact Statement supports our proposed




environmental standards for the management and disposal of spent nuclear




fuel and high-level and transuranic radioactive wastes.  It explains how




we developed the proposed standards, examines alternative actions we could




have taken, and discusses the related environmental impacts.









     In this chapter we describe our proposed action.  We identify the




radioactive materials covered by these standards, how they are produced,




their hazards, and how we can control them.  We also explain the legal




authorities under which these standards were developed.  Finally, we




summarize the various requirements we are proposing to protect public




health and the environment.  The complete text of our proposed




environmental standards is contained in Appendix A.









2.1  HIGH-LEVEL WASTES




     These proposed standards apply to the highly radioactive wastes




resulting from reprocessing  irradiated (spent) nuclear reactor  fuel—and




to the spent fuel itself if  it is disposed of without  reprocessing or if




it is stored pending a determination on disposal.  These  wastes  are




products of the fission and  activation processes associated with nuclear




reactors.  The reactors are  used for commercial power  production,




research, and national defense activities.  Other wastes  containing  large




quantities or high concentrations of radioactivity,  which may represent a

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 significant danger  to  the public health  if  not  properly  handled,  are  not




 specifically  included  in these  standards, but they may require  similar









 environmental standards for disposal of other categories of radioactive



 waste.









      High-level radioactive wastes from reprocessing have been commonly




 designated as the material  remaining after  recovery of uranium and




 plutonium from irradiated  (spent)  reactor fuel.  These wastes include




 residual fissionable materials, fission  products, and activation




 products.   They may be in  liquid or solid form, may  be soluble or




 insoluble  in  water,  and may emit any of  several types of radiation over a



 wide range of energy.









      Both  spent  fuel and high-level  radioactive wastes from reprocessing




 are  intensely radioactive and generate substantial quantities  of  heat.




 Radioactivity and heat  production continue  for  long  periods of  time,




 because  the wastes contain  a number  of long-lived radionuclides.   The




 transuranium  elements,  especially, have  long radiological half-lives  and




 present  a  possible hazard to people  for  tens of  thousands of years.









2.1.1  Commercial Reactor Wastes




     The fuel used in the present generation of  commercial  light-water




reactors consists of a mixture of uranium-238 and uranium-235 dioxides




encased in zircalloy or stainless steel  tubes.  During reactor operation,




fission of the uranium-235  produces energy,  neutrons, and fission products.






                                    8

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The neutrons produce further fission reactions and thus sustain the chain




reaction.  The neutrons also convert some of the uranium-238 into




plutonium-239, which can fission like uranium-235.  In time, the fissile




uranium-235, which originally constituted some 3 percent of the enriched




fuel, is depleted to such a low level that power production becomes




inefficient.  Once this occurs, the fuel bundles are deemed "spent" and




are removed from the reactor (an annual removal rate of 30 tons per




reactor is typical).









     Reprocessing of commercial spent fuel has been proposed with a view




towards recovering the unfissioned uranium-235 and the plutonium for reuse




as a fuel resource.  However, spent fuel generated by commercial power




reactors is not currently being reprocessed.   The radioactive materials




in spent fuel fall into two major categories:  fission products and




actinide elements.  Typically, fresh spent fuel contains more than




100 radioactive nuclides as fission products.  Fission products of




particular importance, because of the quantities produced or their




biological hazard, are strontium-90, technetium-99, iodine-129 and -131,




the cesium isotopes 134, 135, and 137, tin-126, and krypton-85 and other




noble gas isotopes.  The actinides consist of uranium isotopes,




transuranic elements (i.e., isotopes with atomic number greater than 92,




including plutonium-239, americium-241 and -243, and neptunium-237) formed




by neutron capture, and their decay products.  Spent fuel also contains




tritium (hydrogen-3), carbon-14, and other radioactive isotopes created by




neutron activation.  The exact composition of radionuclides in any given

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 spent-fuel sample depends on the reactor type, the initial fuel




 composition,  the length of time the fuel was irradiated, and the elapsed




 time since its removal from the reactor core.









      If fuel  is reprocessed,  the resulting waste is an acidic aqueous




 solution containing most  of the actinides and the nonvolatile fission




 products present in the fuel.   About 0.5 percent of the original uranium




 and  plutonium would be left in  the  waste liquid.   Current Federal




 regulations require that  commercial high-level liquid waste be converted




 to a solid within  5 years.  Calcination is  the best available




 solidification process.   The calcined waste should be converted to  a more




 stable  form for ultimate  disposal.








      The inventory  of  spent fuel  through  1980  was  about  7000  metric  tons




 of heavy metal  (DOE  81).  The activity  of this  spent  fuel  depends heavily




 on its  age, since  fresh spent fuel  contains large  quantities  of




 short-lived fission  products.   By calculating waste activities  as of




 10 years after  removal  from the reactor, one can  largely eliminate  the




 variability due  to short-lived  fission  products.  On  this  basis,  the




 activity of the  7000 metric tons of  spent fuel corresponds  to over




 2 billion curies.  A small amount of commercial fuel  has been reprocessed,




producing about 600,000 gallons of  liquid waste (EPA  80) containing  about




30 million curies (BNWL 76).
                                   10

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     Predictions of high-level waste production from commercial generation




of electrical power depend on estimates of the growth of the nuclear power




industry.  The Interagency Review Group (IRG 79)  estimated a range of




cumulative totals by the year 2000 of about 70,000 to 100,000 metric tons




of spent fuel, corresponding to 148 and 380 gigawatt-electric (GWe) of




installed capacity, respectively.  Arthur D. Little (ADL 79) developed an




upper estimate of 700 GWe of installed capacity by the year 2010, for




500,000 metric tons of spent fuel containing about 200 billion curies.









2.1.2  Defense Wastes




     Weapons program reactors are operated to produce plutoniutn;




reprocessing to recover the plutonium is an integral part of the weapons




program operations.  Naval propulsion reactor fuel elements are  also




reprocessed to recover the highly enriched uranium they contain.









     The present inventory of defense wastes  in the U.S.  is about  290,000




cubic meters (10 million cubic feet or 70 million gallons), stored at




three Federal reservations as liquids, sludges, and  solids  (IRG  79).




Operation of the weapons program reactors and other  defense activities




will continue to produce defense wastes.  Table 2-1  gives projections of




both the quantity and radioactivity of high-level defense wastes (ADL 79).
                                    11

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                                 Table 2-1
                   Defense High-Level Waste Projections
                     Volumes     	Radioactivity (Ci)	

                                 Total
 Site         Cubic Meters   Fission Products    Uranium       Transuranics

Hanford(aJ         200,000       2.5 x 108          710          1.4 x 105

Savannah River(b)   83,000       3.2 x 108           48          7.4 x 105

Idaho^            ll.QQq       l.Q x 1Q8            2          1.0 x 1Q3

   TOTAL           294,000       6.7 x 108          760          8.8 x 105
(a) i99Q waste projection.
(b' 1985 waste projection.
                                   12

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2.1.3  Definition of High-Level Wastes in 40 CFR 191

     In developing its regulations for disposal of high-level wastes in

geologic repositories (10 CFR 60), the Nuclear Regulatory Commission (NRC)

defined high-level wastes as (NRC 81b):

          ". . . (1) irradiated reactor fuel, (2) liquid wastes
     resulting from the operation of the first cycle solvent
     extraction system, or equivalent, and the concentrated wastes
     from subsequent extraction cycles, or equivalent, in a facility
     for reprocessing irradiated reactor fuel, and (3) solids into
     which such liquid wastes have been converted."


     For the purposes of our environmental standards, we are proposing a

somewhat different definition of high-level wastes:

         "... any of the following that contain radionuclides in
     concentrations greater than those identified in Table 1:
     (1) liquid wastes resulting from the operation of the first
     cycle solvent extraction system, or equivalent, in a facility
     for reprocessing spent nuclear fuels; (2) the concentrated
     wastes from subsequent extraction cycles, or equivalent;
     (3) solids into which such liquid wastes have been converted;
     or (4) spent nuclear fuel if disposed of without reprocessing.


     There are two substantive differences between our definition and the

one in 10 CFR 60.  The first is that our definition does not  identify

spent nuclear fuel as a waste unless it is determined that such fuel will

be disposed of without reprocessing.  Thus, >--ovisions of our standards

that specifically apply to waste would not apply to spent fuel until such

a determination was made.
     The other major difference is that our definiton of high-level waste

excludes materials with concentrations of radioactivity below those in

Table 2-2 (which reproduces Table 1 of the proposed standards).
                                   13

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                                 Table 2-2:
          Concentrations  Identifying High-Level  Radioactive  Wastes
                       (Table  1  in 40 CFR Part  191)

      Radionuclide                                        Concentration
                                                  (curies per  gram  of waste)
     Carbon-14   	8xl06
     Cesium-135	8x  10~4
     Cesium-137	5x  10~3
     Plutonium-241	3xlO~
                                                                  _3
     Strontium-90	  7x10
     Technetium-99	3x  10~
     Tin-126	7x  10~7
     Any alpha-emitting transuranic
       radionuclide with a half-life -----------  1x10
       greater than 20 years
     Any other radionuclide with a half-life
                                                                  _3
       greater than 20 years ---------------  1x10
     NOTE:  In cases where a waste contains a mixture of radionuclides,
it shall be considered a high-level radioactive waste if the sum of the
ratios of the radionuclide concentration in the waste to the concentration
in Table 1 exceeds one.
     For example, if a waste contains radionuclides A, B, and C in
concentrations Cfl, Cb, and Cc and if the concentration limits from
Table 1 are CLg, CLb, and CLC, then the waste shall be considered
high-level radioactive waste if the following relationship exists:
                  ca       cb       cc
                  CLa      CLb      CLC    **
                                   14

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We propose to make this exclusion because some wastes derived from




reprocessing spent fuel may not require the same isolation as the wastes




containing most of the radioactivity produced in the fuel.  In particular,




processing of certain high-level wastes from national defense activities




may remove much of the radioactivity, leaving residual material that is




far less dangerous (for example, large volumes of salt cake containing




relative low concentrations of technetium-99).  Our definition would allow




such material to be disposed of without having to meet the stringent




requirements we feel are appropriate for the most highly radioactive




wastes.  The disposal requirements for materials excluded by our




definition will be addressed as we develop standards for low-level




radioactive wastes.









     The levels in Table 2-2 are equivalent to the maximum concentrations




for acceptance at shallow-land burial sites that the NRG recently




promulgated as part of 10 CFR Part 61 (47 FR 57446).  The NRG derived the




concentration limits in 10 CFR 61 so that a person intruding into a




shallow-land site—after institutional controls were no longer effective—




should not receive a radiation exposure greater than 500 millirem per year




(NRG 81a and WI 81).  We converted the units of these concentrations from




curies per unit volume to curies per unit mass by assuming a density of




1 gram per cubic centimeter.  For wastes with higher densities, the levels




in Table 2-2 would be somewhat more stringent than the levels in 10 CFR 61.
                                   15

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 2.2  TRANSURANIC WASTES




     The proposed  standards also apply  to wastes containing alpha-emitting




 transuranic nuclides with half-lives greater  than  1 year at concentrations




 greater than  100 nanocuries per gram (nCi/g).  Alpha-emitting transuranic




 nuclides represent a special type of hazard because of their long




 half-lives and high radiotoxicity.  Cohen and King (CO 78) considered the




 levels at which these materials might require handling different than that




 generally given to low-level radioactive wastes.  They concluded that the




more hazardous transuranic elements could cause excessive radiation doses




 if they are present at concentrations of about 1 microcurie per cubic




centimeter of waste.  A task group (HE 79) recommended shallow-land burial




for Department of Energy (DOE)  transuranic wastes with concentrations




below 1 to 10 nCi/g.  The task group also suggested that somewhat deeper




burial was suitable for transuranic wastes with concentrations up to




100 nCi/g.  In view of these studies, txansuranic wastes with




 concentrations aoove 100 nCi/g were included under these standards.









     There are about 1100 kilograms of transuranic elements at levels




above 10 nCi/g in defense wastes at DOE facilities (IRG 79).  These




radionuclides are contained in about 400,000 cubic meters (14 million




cubic feet) of transuranic-contaminated wastes.  Both the DOE and the EPA




have studies underway to define the amounts and characteristics of defense




transuranic wastes containing more than 100 nCi/g.
                                   16

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 2.3  BIOLOGICAL EFFECTS OF RADIATION




     Radioactive materials may emit participate  (alpha and  beta) or




 electromagnetic (gamma and X) radiation  that can  interact with matter.




 In this  interaction, energy  is transferred  from  the  radiations to  the




 matter.  Energy transferred  to matter  is called  absorbed dose.  The unit




 of dose  is the rad  (1CRU 71), which corresponds  to the absorption  of




 100 ergs per gram of material (0.01 joules  per kilogram).









     Gamma radiation, X-rays, and beta particles  are  sparsely ionizing




 radiations, with low linear  energy transfer (low-LET).  Alpha particles,




 such as  those emitted by transuranic radionuclides,  are high-LET




 radiations that exhibit a very dense ionization  pattern in  tissue.  The




 interaction of radiation with living tissue in man,  animals, or plants may




 damage the tissue.  Since the absorption of the  same  quantity of energy




 from different radiations can produce different  biological  effects  in




 living tissue, modifying factors are applied for  the  nature of the




 radiation and other considerations.  The product  of  the dose and the




modifying factors is called  the dose equivalent  (ICRU 73);  the unit of




 dose equivalent is the rem.  The total impact on  a population is the




 population dose equivalent,  the sum of the  radiation  dose equivalent




 incurred by each person in the population.









     Releases of high-level wastes to the environment may cause human




exposure from direct radiation or from breathing  or  ingesting radio-




nuclides.  These modes of exposure may arise because  of volatilization,
                                   17

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 particulate  dispersion,  and  dissolution of  the waste materials—the




 predominant  pathway(s) depend  on the mode of  release,  the  specific




 environmental  circumstances, and the isotopic  composition  of  the released




 material.








     Exposure  to  large amounts  of  radiation,  on the  order  of  tens to




 hundreds of  rems,  in a short time  (acute exposure) can produce a number  of




 effects on humans  ranging  from  barely detectable to  severe.   These  effects




 range from chromosome and  blood  cell changes  through loss  of  appetite  and




 loss of hair,  to diarrhea, vomiting,  and even  death.   The  severity  of  the




 effect is directly related to the  amount of radiation  absorbed by the




 individual.









     Releases  from a high-level waste repository, if they  occur  at  all,




 are expected to be much smaller  than  those that  would  cause acute effects.




 These continuing (chronic) exposures  to radiation, at  levels  well below




 those noted  in the preceding paragraph, have no  detectable early  effects.




 However, in  such cases, there may  be delayed effects.  The major delayed




 effects are  the development of cancer and the production of genetic




 changes.  In contrast to acute effects,  for which the  severity depends on




 the dose equivalent incurred, delayed effects develop  only in some  of  the




exposed individuals.   The rate of  incidence is a statistical  function  and




depends on the total population dose equivalent  incurred.
                                   18

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     The relationship between the radiation dose and the number of delayed




health effects has been studied extensively.  The National Academy of




Sciences (NAS 72 and 80) and the United Nations Scientific Committee on




the Effects of Atomic Radiation (UN 77) have published reviews of the




current state of knowledge on this subject.  Because of the inherent




uncertainties in our scientific evidence, and as a prudent general policy




for public health protection, we use a linear, non-threshold relationship




between radiation dose and the risk of incurring cancer or genetic




abnormalities.  This relationship assumes that any exposure can produce




some harm and that the harm is proportional to the absorbed dose.  For the




low dose rates considered here, this assumption may overestimate the risk




from low-LET radiation.  Because no method of disposing of hazardous




materials can be entirely free of risk, the development of standards




limiting the level of risk cannot be based solely on health




considerations.  The extent to which risk can reasonably be reduced,




considering social, economic, and other factors, must also be evaluated.









2.4  METHODS OF CONTROL




     Protection of humans from the hazards of radioactive wastes requires




shielding or permanent physical separation of wastes from people.




Specially designed storage facilities may provide short-term protection,




but long-term protection must rely on passive natural and engineered




barriers requiring no upkeep during their entire functional lifetime.




(Conceivably, the wastes could be removed entirely from the biosphere  by
                                   19

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 transport away  from  the earth  or  by  transmutation  to  less  harmful




materials, but  these alternative  concepts do not currently appear




practical.)









     Subpart A  of the proposed standards applies to the management,




 including storage, processing, and emplacement, of the wastes, but  it does




not cover their offsite transportation.  Subpart B applies  to the wastes




after disposal.  For a geologic repository, for example, Subpart B  would




take effect when the mine is backfilled and sealed.









2.4.1  Storage of Spent Fuel and High-Level Wastes




     There is considerable experience with storing spent fuel.  Fuel




elements are kept in racks in water basins to remove the heat and provide




shielding.   Dry, air-cooled storage of aged spent fuels, which produce




less heat than fuels recently removed from a reactor core, has been




proposed.









     There is also considerable experience with storing liquid wastes from




 the  reprocessing of  spent fuel, since most defense wastes  are of  this




nature.  Most of these acidic  liquids have been neutralized  for storage  in




 large steel tanks.  A small  fraction of these existing  liquid wastes has




 leaked  from the tanks, requiring  transfer into new tanks.  Liquid wastes




can be  converted to  solids, which are then stored  like  spent  fuel.
                                   20

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2.4.2  Disposal in Mined Repositories




     Development work on methods for disposal of spent fuel and solidified




high-level and transuranic radioactive wastes has concentrated on mined




geological repositories.  Such repositories would be constructed at depths




greater than 1000 feet by using conventional mining techniques in suitable




host media.  Suggested media include granite, basalt, volcanic tuff, and




salt.  Wastes in canisters would be placed in holes in the mine floor.




When the repository is full, it would be backfilled.  After a validation




period, during which the wastes could be retrieved, the site would be




permanently  sealed.  Protection would be provided by  a stable and




insoluble waste form, a durable canister, a  stable host medium, and  low




migration potential for radionuclides through the environment around  the




host rock.   Mined geological repositories are expected to  be  available for




use  sooner than any other disposal method.









2.4.3  Other Disposal Methods




     A number  of  other  methods  have  been  suggested  for disposal of




high-level radioactive  wastes.  These methods  include:




     — Placement of  fresh  liquid  or solid  wastes  directly into rocks by




        melting.   The heat  of  the  fresh wastes  would melt  the rock, and




         the  wastes would become incorporated as an integral component of




         that rock.




     — Placement of  containers of waste  into holes 10,000 to 30,000 feet




        deep.
                                    21

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      — Placement of containers of high-level wastes on or under the ocean

         floor.   Disposal of high-level wastes in the ocean is  prohibited

         under United States law by the Marine Protection,  Research,  and

         Sanctuaries  Act  of 1972.

      — Transport into space (extraterrestrial disposal)  or transmutation

         in  fission or fusion reactors  of  the  most  long-lived or  hazardous

         radionuclides.   Both of these  methods would  dispose of only  a

         portion  of the wastes,  leaving the  rest  for  disposal on  earth.



      These  alternative methods  of  disposal  are discussed  in more  detail in

Chapter 3.



2.5  AUTHORITIES

     Reorganization Plan No. 3  of  1970  transferred to EPA  two  functions

derived  from  the Atomic Energy Act of  1954:   (1) the responsibilities of

the former Federal Radiation Council, and (2)  the authority  to set

generally applicable environmental standards.  The authority to set

generally applicable environmental standards was transferred from the

Atomic Energy Commission:

          ".  . .  to the extent that such functions of the Commission
     consist of establishing generally applicable environmental
     standards for the protection of the general environment from
     radioactive  material.  As used herein, standards mean  limits on
     radiation exposures  or levels, or concentrations or quantities
     of radioactive material, in the general environment outside the
     boundaries of locations under the control of persons possessing
     or using radioactive material."
                                   22

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     Our generally applicable environmental radiation standards are rules

that apply outside the controlled boundaries of facilities using

radioactive materials.  They are designed to preserve the general

environment and protect the public health.  The NRC—or in some cases the

DOE—implements and enforces these standards by issuing specific

regulations and by assuring that individual and interrelated components of

the fuel cycle are constructed and operated consistent with licensing

provisions and all applicable regulations.



     The authorities inherited from the former Federal Radiation Council,

under Executive Order 10831 and the Atomic Energy Act (42 U.S.C. 2021(h)),

include the responsibility to:

          ". . . advise the President with respect to radiation
     matters, directly or indirectly affecting health, including
     guidance to Federal agencies in the  formulation of radiation
     standards ..."


     We provide advice by preparing Federal radiation protection guidance

for approval by the President as executive direction to Federal agencies.

These guides establish overall direction  for the radiation protection

programs of all Federal agencies.  Executive Order 12088 makes  the  head  of

each agency responsible for compliance with such guides, once  the

President has approved them.  In addition, the Order directs the EPA to

monitor compliance by Federal agencies.   Conflicts on implementation may

be resolved by the Director of the Office of Management and Budget.

Exemptions from the guides may be granted by the President.
                                   23

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      The environmental protection requirements described in this Draft




 Environmental Impact Statement are proposed solely under our generally




 applicable environmental standards authority.  However, we considered




 alternatives to our proposed action that would involve our Federal




 radiation protection guidance authority.  These alternatives are discussed




 in Cnapter 4.









 2.6  THE PROPOSED STANDARDS




      We are proposing generally applicable environmental standards for




 both the management and  the disposal  of spent fuel,  and high-level and




 transuranic radioactive  wastes.   When promulgated,  these standards will




 become  a new Part 191 to Title 40 of  the Code of  Federal Regulations




 (40 CFR 191).  Subpart A applies  to all operations  up to and including the




 emplacement of  the  wastes in a disposal system.   Subpart B  applies to the




 period  after disposal, except  that these standards would not apply to




 disposal directly into the  oceans or  ocean  sediments.   Both Subparts are




 summarized  below; they are  examined in  more detail  in Chapters  6  and 7 of




 this  document.









 2.6.1   Standards  for  Management and Storage (Subpart A)




     The preparation  for  disposal or  storage  of these  materials and  the




 placement of  the materials  in  a disposal  system are  included under these




 standards.  The transportation of the materials is not  covered.  These




 standards for waste management and storage operations  apply only to




anticipated normal events or releases.
                                   24

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     Most of the wastes subject to this regulation (40 CFR  191) were and




will be derived from the production of commercial nuclear power and from




activities  in national defense programs.  We have issued regulations




(40 CFR 190) covering one aspect of nuclear power production—the uranium




fuel cycle—from milling of ore through the generation of electricity at a




nuclear power plant and the reprocessing of spent uranium fuel.  The




40 CFR 191  regulations proposed here ideally should  include all operations




following those covered in 40 CFR 190 up to and  including the final




abandonment of the waste materials.  In this way, no portion of the




uranium fuel cycle subsequent to mining (which is not included under the




authority of the Atomic Energy Act of 1954, as amended) would be left




uncontrolled.  Waste management problems in national defense programs, and




in other projected or contemplated fuel cycles,  are  similar enough to




those in the uranium fuel cycle that they can also be included under




40 CFR 191.  The storage of spent fuel, whether  or not it has been




designated as a waste material, is included in this  regulation when such




storage is not already regulated under 40 CFR 190.









     We examined the radiation exposure to members of the public from




waste management and storage prior to geologic disposal and determined




that it is feasible, with available technology and proper siting of




facilities, to limit maximum individual doses from all normal operations




to those dose limits established for uranium fuel cycle operations in




40 CFR 190.  Those limits are 25 raillirem to the whole body, 75 millirem




to the thyroid, and 25 millirem to any other organ.  Therefore, we propose
                                   25

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to extend the limitations of 40 CFR 190 to the operations covered by




40 CFR 191.  Since some of these operations can be conducted with




individual doses significantly lower than these limits, Subpart A also




requires that exposures be reduced below these limits to the extent




reasonably achievable, taking into account technical, social, and economic




considerations.









     These limitations for individual exposures from waste management




operations do not include transportation of spent fuel or high-level and




transuranic wastes.   Normal exposures in transportation would be in the




direct radiation to persons near shipments;  radioactive materials could be




released only due to accidents.   Transportation of spent fuel was




considered in the development of the Uranium Fuel Cycle (40 CFR 190)




standards.   We found that the average radiation dose to individuals and




the total dose to the general public were small for transportation




activities in the uranium fuel cycle.   It is unlikely that an individual




would receive an annual exposure greater than 25 inillirem from exposure to




shipments.   We concluded, however,  that providing a regulatory guarantee




that no individual could receive annual doses greater than 25 milliretn




would be extremely difficult because of the  problems involved with




identifying and  monitoring individual exposures.  The expected risks from




transportation of high-level wastes are similarly small.  Thus, it is




inappropriate to include transportation under these standards.
                                   26

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2.6.2  Standards for Disposal (Subpart B)




     The intent of disposal is to isolate  the wastes from the environment




for a long time without the need for human intervention.  Therefore,  the




standards for a disposal system can be implemented only through general




planning and selection criteria and by designing the system to meet




projected performance requirements.








     The proposed standards for disposal apply to both normal and




accidental releases, and both types of events must be considered by those




who will design and license a repository.  The most significant releases




may result from unplanned or accidental events or processes.  These




accidental releases would occur long after the conclusion of normal




operations at the disposal site.  A release may be undetected for  long




periods of time, and there may be no way to correct the situation.  There




may be no personnel on hand to take protective actions, and  the very




concept of the site as a place under control of an organization may be




inappropriate.  Addressing the problem of unplanned releases  is,




therefore, a prime  requirement of  the proposed  standards  for disposal.









     The proposed standards  for disposal contain  numerical  containment




requirements for  the first 10,000  years after disposal,  accompanied by




qualitative requirements to assure  that these containment  requirements  are




met.  These two parts  of our  proposed  action are  complementary:  the




containment requirements set  limits on potential  releases  of radioactive




materials to the  environment—limits  that will  serve  as overall  design
                                    27

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 objectives for disposal systems; the assurance requirements provide the




 framework necessary to develop appropriate confidence in meeting these




 projected release limits in spite of the inherent uncertainties of




 predictions  over 10,000 years.  In addition,  the proposed standards




 contain procedural requirements to ensure that the containment




 requirements are properly  applied to specific disposal systems.









      Containment Requirements




      As our  primary  environmental protection  standards,  we are proposing




 numerical limits on  the projected maximum amount of radioactivity a




 disposal system  can  release to the environment.   These requirements would




 apply to both  expected  and  accidental  releases that could occur within




 10,000  years after waste disposal.   They place a limit on the  harm that  a




 disposal system  can  cause to future generations.









      For high-level  wastes,  the  release  limits are  expressed as the




 maximum amounts  of radioactivity  that can  be  released from disposal of




 tne wastes generated from 1000 metric tons of  heavy  metal  (MTHM).   For




 transuranic  wastes,  the  release  limits are expressed as  the maximum




 amounts  of radioactivity that can  be released  from  disposal of  transuranic




 wastes containing  1 million curies  of alpha-emitting transuranic




 radionuclides.  These units were chosen  so that  the  standards would




 require  alpha-emitting radioactivity from either  high-level or  transuranic




wastes to be isolated with about the same degree  of  effectiveness.
                                   28

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     To determine the potential harm to future generations from waste




disposal, we studied the projected environmental impacts of high-level




waste disposal in mined geologic repositories.  Our studies were based on




the potential releases from a generic model of a repository containing




100,000 MTHM as spent reactor fuel.  This is about as much waste as would




be generated during the 40-year operating lifetimes of  100 large reactors




of current design.  We examined the risks over the first 10,000 years




after disposal.









     As a basis for comparison, we also evaluated other sources of




radiation exposures to present and future generations.  We looked at the




radiation risks from natural background radiation, from untnined uranium




ore bodies (the initial source of these wastes), from commercial nuclear




power generation, and from fallout from previous atmospheric  testing of




nuclear weapons.









     Based on these analyses and evaluations, we decided we could choose




performance requirements that would satisfy two objectives:









     First, the requirements would limit  the  harm to future generations  to




no more than 1000 excess cancer deaths over 10,000 years  (an  average  of




one extra death every 10 years) from disposal of the wastes from  100,000




MTHM.  This risk is about the same as our smallest estimate of  the  harm




from an equivalent amount of unmined uranium  ore, and  it  is much  smaller




than the risks from the other sources of  radiation exposure that we




studied.
                                   29

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      Second,  the  requirements  appear  to  be  achievable  by well designed




 geologic  repositories  at  carefully  selected sites.  Any other method used




 would have  to provide  at  least as much protection as that projected for




 geologic  disposal.









      Once we  chose  this  level  of protection as  a basis  for  these




 standards,  we developed a set  of numerical  release  limits for the various




 radioisotopes in  the wastes.   We estimated  how  many curies  of each




 radioisotope  would  cause  1000  deaths  over 10,000 years  if it were the only




 radionuclide  released; this number  of curies determined the release limit




 for  that  radionuclide.  For releases  involving  more than one radionuclide,




 the  allowed release for each radionuclide is reduced to a fraction of its




 limit to  assure the overall limit on  harm is not exceeded.  The procedure




 for  using the release  limits is described in Table 2 of the proposed




 standards (see Appendix A of this document).









      Assurance Requirements




      Closely  associated with our numerical  containment requirements are




 seven qualitative requirements we believe are essential for developing the




 needed confidence that these long-term release  limits will  be met.  Our




 assurance requirements address and compensate for the uncertainties that




 necessarily accompany plans to isolate high-level and transuranic wastes




 from  the  environment for a very long  time.  No matter how promising




analytical projections of disposal system performance appear to be, these




wastes should be disposed of in a cautious manner that reduces the
                                   30

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likelihood of unanticipated releases.  Our assurance requirements provide




the context necessary for application of our containment requirements,  and




they should insure very good long-term protection of the environment.









     Several of the concepts incorporated in these assurance requirements




were adapted from the Federal radiation protection guidance for all types




of radioactive waste disposal that we proposed for public comment on




November 15, 1978 (43 FR 53262).  After reviewing the responses we




received, we decided that the characteristics of different kinds of




radioactive waste are not sufficiently similar for generally applicable




criteria to be appropriate.  Therefore, we stopped developing this Federal




radiation guidance (46 FR 17567).  However, we also determined that,




because of the uncertainties inherent in meeting our 10,000-year




containment requirements, several of the principles included in  this




earlier proposal needed to be incorporated as integral parts of  these




standards for disposal of high-level and transuranic wastes.









     We expect that the specific steps taken by the NRG  or the DOE  to




comply with each of these seven assurance requirements will be described




in the Federal environmental impact  statement—and other appropriate




decision documents—for each disposal system.  These seven requirements




are:




     1.   Wastes shall be disposed of promptly once disposal systems are




          available and the wastes have been suitably conditioned  for




          disposal.
                                   31

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2.  Disposal  systems  shall  be  selected  and  designed  to  keep  releases




    to the accessible environment as  small  as  reasonably achievable,




    taking into account  technical,  social,  and economic




    considerations.




3.  Disposal  systems shall  use several  different types  of barriers to




    isolate the wastes from the accessible  environment.  Both




    engineered and natural  barriers shall be included.  Each such




    barrier shall separately be designed to provide  substantial




    isolation.




4.  Disposal systems shall not rely upon active institutional




    controls to isolate the wastes beyond a reasonable  period o£ time




    (e.g., a few hundred years) after disposal of the wastes.




5.  Disposal systems shall be identified by the most permanent




    markers and records practicable to  indicate the  dangers of the




    wastes and their location.




6.  Disposal systems shall not be located where there has been raining




    for resources or where there is a reasonable expectation of




    exploration for scarce or easily accessible resources in the




    future.   Furthermore, disposal systems shall not be located where




    there is a significant concentration of any material which is not




    widely available from other sources.




7.  Disposal systems shall be selected so that removal of most of the




    wastes is not precluded for a reasonable period of time after




    disposal.
                              32

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     Procedural Requirements




     The containment requirements were derived with the assistance of our




performance assessments of long-term repository performance (summarized in




Chapter 5).  When these requirements are applied to a particular disposal




system, some of the procedures we used in our assessments must be retained




to ensure that the intent of our containment requirements is met.  On the




other hand, some of the assumptions we made should be replaced with the




specific information developed for each particular system.  Our three




procedural requirements provide instructions necessary for performance




assessments of specific disposal systems to properly demonstrate




compliance with our containment requirements:




     1.  The assessments shall consider realistic projections of the




         protection provided by all of the engineered and natural barriers




         of a disposal system.




     2.  The assessments shall not assume that active institutional




         controls can prevent or reduce releases to the accessible




         environment beyond a reasonable period (e.g., a  few hundred




         years) after disposal.  However, it should be assumed  that  the




         Federal Government is committed to retaining passive




         institutional control of disposal sites in perpetuity.  Such




         passive controls should be effective in deterring systematic or




         persistent exploitation of a disposal site, and  it should be




         assumed that they can keep the chance of  inadvertent human




         intrusion very small as long as the Federal Government  retains




         such passive control of disposal sites.
                                   33

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      3.  The assessments shall use information regarding the likelihood of




          human intrusion, and all other unplanned events that may cause




          releases to the accessible environment,  as determined by the




          implementing agency for each particular disposal site.








 2.7  IMPLEMENTATION




      The standards for waste management and storage (Subpart A) will be




 implemented by the NRC for  commercial nuclear power activities and by the




 DOE for national  defense facilities.   Implementation procedures for




 Subpart A will be very similar to those for the Uranium Fuel Cycle




 Standards (40  CFR 190).









      The standards  for disposal  (Subpart  B)  will  be implemented by the NRC




 for all high-level  wastes,  whether  the  wastes come  from commercial or




 national defense  activities.  The NRC will  do this  by  developing the




 necessary regulations  (primarily  10 CFR 60)  and by  issuing  appropriate




 licenses.   Under  current  law, disposal  of transuranic  wastes from national




 defense activities  is  not regulated by  the  NRC; therefore,  the  DOE will




 implement our  requirements  for disposal of  these  transuranic wastes.









     The  containment requirements in Subpart  B  will  be  applied  through




design  specifications, and  the implementing agencies will have  to  evaluate




 long-term projections  of the disposal system  performance.  As a  result, a




vital part of  implementation will be the use  of adequate models,  including




the probabilities of unplanned events, to relate  appropriate  site  and
                                   34

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engineering data to projected performance.  The NRG has made substantial




progress in developing such analytical models to predict long-term




performance of actual geologic repositories.  These models are quite




detailed, and they are capable of evaluating how important any




uncertainties in specific types of data are to the overall projections of




repository performance.  Thus, they can provide information about any




needs for obtaining better data to determine if repositories meet the




containment requirements of these standards.









     At our request, the National Academy of Sciences  studied the




difficulties in verifying compliance with long-term environmental




protection requirements for geologic disposal (NAS 79).  Our NAS panel




developed an approach that specifies the  types of information needed  and




outlines appropriate methods  for obtaining  this data at prospective




sites.  Based on this NAS study, the NRC models, our own analytical




efforts, and the confidence that should be  provided by our  assurance




requirements, we have concluded that our  containment requirements  can be




effectively implemented.
                                   35

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36

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                                 Chapter 3




                 ISSUES IN SETTING THE DISPOSAL STANDARDS








     Our standards  for disposal of high-level and  transuranic wastes are




 intended to provide  long-term protection over the  very  long periods that




 these materials present a possible hazard  to man.  This approach




 introduces questions not ordinarily considered in  regulatory decisions




 having only short-term impacts.  For disposal of high-level wastes, most




 of the potential harm would occur far  into  the future.  This chapter,




 therefore, describes how we addressed  each  of the  following questions:




     1.  How should we treat risks to  future generations?




     2.  How should we assess the capability of disposal systems  to reduce




         long-term risks?




     3.  How should we consider the uncertainties  in predicting the future?




     4.  How should we express the standards to best reflect the




         environmental protection objectives we desire?




     5.  How should we select the period of time over which we should



         regulate?









 3.L  RISKS TO FUTURE GENERATIONS




     The National Environmental Policy Act  (NEPA)  requires that we




 consider risks to future generations.  Since disposal of high-level and




 transuranic radioactive wastes create  some  risks for present and  future




generations,  regulatory actions must attempt to reduce  those risks to an




acceptable level.
                                   37

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      Some might argue  that  no  risks  to  future  generations  from disposal  of




 these wastes should be allowed,  since these  generations  receive no  direct




 benefits from present activities and since they  cannot take  part  in




 decisions to create such risks.  Others would  hold  that  the  industrial,




 medical, scientific, and military activities that produce  the  wastes may




 contribute to the knowledge and security of  future  generations.  Moreover,




 some of these activities may be alternatives to activities that might




 result in worse long-term pollution and in depletion of  resources.  These




 indirect benefits  may justify some risks to future generations  from the




 long-lived  radioactive wastes generated by these activities.









      We  have  not  tried to resolve this  question of intergenerational




 equity by attempting  to make risk versus benefit judgments.  We believe




 that  the risks  to  future  generations  should be  small—at  the very least,




 we  should not pass  on  to  future generations any radiation risks greater




 than  risks we would be willing  to assume ourselves.   We do not know of




 disposal technologies  foreseeable within this century that can eliminate




 all risks to future generations  from  disposal of these wastes.   However,




 our analysis of the risks associated  with  undisturbed uranium ore bodies




 has reinforced our decision  about the reasonableness of the residual risks




 permitted under our proposed disposal standards.  This  analysis indicates




 that  leaving equivalent amounts  of uranium  ore  unmined  presents at  least




 as great a risk to future generations as disposal of the  wastes covered by




 these standards.  Thus, there would appear  to be  no  increase  in future




risks caused by disposal of these wastes.
                                   38

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     We are not sure that this analysis provides an adequate means of




resolving the question of intergenerational risk.  It has, however, helped




to influence our decision of what is an acceptable level of residual risk




given our current scientific, technological, and fiscal capabilities.  The




following discussion provides more details on the perspectives we used to




evaluate risks to future generations.









3.1.1.  Perspectives on Long-Term Risks




     People have always been exposed to radiation.  Life has evolved in a




natural background radiation field that gives most humans doses on the




order of 100 millirem per year.  Man has added more radiation—from




medical, industrial, and national defense activities—to  this natural




background.  We believe our standards should limit the risks from




high-level and transuranic wastes after disposal to a small fraction of




today's radiation risks.  As a basis for comparison, we present estimates




of the risks from four radiation sources.  Two of these sources, natural




background radiation and uranium ore bodies, are relatively constant, and




the risks will continue far into the future.  The risks from the other  two




sources, generation of nuclear power and nuclear weapons  fallout,  are




largely short-term and will decrease with time.  The estimates of  risks




from natural background radiation, generation of nuclear  power, and




nuclear weapons fallout are much higher than those we estimate from




high-level wastes after disposal.  The estimates of risks from uranium  ore




bodies vary over a large range.  Our estimates of the risks from the




disposal of high-level wastes are slightly above the lower bound of  this




range and are substantially less than the higher part of  the range.







                                   39

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 3.1.2.   Risks  Associated with Natural  Background  Radiation



      Humans are  exposed  to natural  background  radiation  from  cosmic  rays,



 from naturally occurring radionuclides in  the  earth,  and from naturally



 occurring radionuclides  in the body, primarily potassium-40.  Cosmic



 radiation originates  primarily in galactic  and solar  processes.  Naturally



 occurring radionuclides  in the earth's crust comprise chiefly members of



 the  uraniunr-238  series,  members of  the thorium-232  series, and



 potassium-40.  There  are approximately 2.7  grams  of uranium-238, 9.6 grams



 of thorium-232,  and 20,900  grams of potassium  per metric  ton  of rock in



 the  earth's  crust (FO 71).  Rock in the earth's crust contains about


      -5                                                         -5
 1.4x10    curies  of alpha activity per  metric ton  and about 2.7x10



 curies of beta activity.







     The  United Nations  Scientific Committee on the Effects of Atomic



 Radiation (UNSCEAR) has  estimated that  the average  annual radiation dose



 from external exposure to natural sources in "normal" areas of the world



 includes  28 millirads of cosmic radiation and 32 millirads of terrestrial



 radiation (UN 77).  The UNSCEAR committee stated  that the dose from



 external  and internal radiation was 92 millirads per year to  red marrow,



 78 to gonads, and 110 to  lungs.  The corresponding dose equivalent rates,



with a quality factor of 10 for the alpha and neutron components,  are 109,



86,  and 425 millirem per year (mrem/yr), respectively.  The difference



among organs is due to differing radionuclide contents in the organs.
                                   40

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     People receive from about 50 to about 195 mrem/yr natural background




radiation dose equivalent from cosmic, terrestrial, and internal radiation.




The highest dose equivalents received in large urban areas are about




130 mrem/yr in Denver, Colorado, and about 115 mrem/yr in Albuquerque,




New Mexico.  The highest average for a state is about 115 mrem/yr in




Colorado.  Table 3-1 gives the distribution of the dose received by the




population.  It shows tnat about 60 percent of the United States




population receives from 65 to 90 rarem/yr; 20 percent more than




90 mrem/yr; and 5 percent more than 105 mrem/yr.  Natural background




radiation levels in the United States vary by a factor of three.









     Information on the variability of the dose equivalent received by




people in the United States has been published by Oakley (OA  72) and




revised by Bogen (BO 80).  Cosmic radiation dose equivalent in  the United




States varies from 29 mrem/yr at sea level to about 125 rarem/yr  at




3200 meters (10,500 feet), the altitude of Leadville, Colorado  (NCRP  75).




The terrestrial component dose equivalent from radionuclides  in  the earth




can be characterized as low (average of 23 mrem/yr) for the Atlantic  and




Gulf coastal plain, moderate (average of 46 mrem/yr)  for most of the  rest




of the United States, and high (average of 90 mrem/yr) for a  small  region




in the Colorado Front Range.

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                                 Table 3-1



             Distribution  of  Natural  Radiation Dose Equivalents

                  (Terrestrial Plus Cosmic Plus Internal)
     Dose equivalent*                   Percentage  of U.S.  Population

         (mrera/yr)                       receiving more than  stated dose
           50                                         100

           65                                         80

           75                                         50

           90                                         20

          105                                          5

          120                                          1
*Dose equivalent includes 25 millirem per year from internal radiation,
Values have been rounded to the nearest 5 tnrem/yr.
                                   42

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     We estimate that an excess of about 200 fatal cancers per million




person-rem can be expected (NAS 72).  This means that an increase of




1 mrem/yr in natural background radiation (about 1 percent) to everyone in




the present U.S. population would result in about 40 excess fatal cancers




per year.  A 1 mrem/yr increase in radiation exposure to 5 million people




(a population that might be exposed to releases from a large high-level




waste repository) would result in  about one additional fatal cancer per




year.









     Risks associated with natural background radiation are similar to




those from radioactive wastes; both produce the same kinds of health




effects and both persist for long periods of time.  The risk from natural




background radiation is much larger than that projected for releases from




a high-level waste repository.  However, in general, the variability of




this risk does not seem to greatly influence the choice of people in




selecting a place to live.









3.1.3  Risks from Uranium Ore Bodies




     A comparison of the risks from a high-level waste repository with




those from an undisturbed uranium ore body offers another  useful




perspective.  In making our assessment (WI 80), we consider that uranium




dissolved by the groundwater eventually reaches a stream.  As  the uranium




moves along the aquifer, radiunr-226 and other daughters are formed.  We




estimated the risk from a model generic ore body containing 620,000  tons




of uranium oxide, the amount that would have to be mined to produce  the




high-level wastes contained in our model repository.  For  this estimate we






                                   43

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 used  the  same  generic  environmental  model  (SMJ  82)  that  we  used  for

 assessing the  risk  from high-level radioactive  wastes  in mined geological

 repositories (SMC 82).



     The  risks  from uranium ore  bodies cover a  wide  range,  depending on

 the ability of  uranium  and  its decay products (particularly radium-226)

 to leave  the ore body and reach  people.  To estimate a minimum impact, we

 selected  a uranium  concentration in  groundwater at  the low  end of  the

 reported  range  (FI  55).  For this assumption, we calculate  an occurrence

 of two excess fatal cancers  per  100 years  for the model  generic  ore body.

 Based on  pre-operational data for three actual  uranium mines, we calculate

 the impact of releases  of radionuclides to groundwater to cause  between

 100 to about 1000 excess cancers per year.  These estimates  from actual

 ore bodies may  be high, because  the reported concentrations  in the

 groundwater include some measurements from the  oxidizing uranium-rich

 groundwater.  Our estimate of the risks from the wastes  after disposal is,

 therefore, similar to the lowest estimates of risk from  the  unmined ore

 bodies.



 3.1.4  Risks from Nuclear Power Generation

     Ellett and Richardson (EL 77) estimated that exposure  to the radiation

 incident to the generation of 1 gigawatt-year of electrical  energy could

 result in an average of about 1.2 fatal cancers* in the  first 100 years.
*The estimated number of fatal cancers represents a probability of
incidence in the entire exposed population.
                                   44

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More recent information reduces this estimate to 0.8 fatal cancers.




Operation of the 50 gigawatts of nuclear power generation available in 1977




(ADL 79) at 60 percent of capacity could cause about 24 fatal cancers per




year.  To the extent that radionuclides such as carbon-14, thoriura-230,




and radium-226 and its daughters persist beyond 100 years, this estimate




may be low.









     Since they are limited in time, the risks from generating nuclear




electric power are not directly comparable to risks that would be




associated with disposal of high-level radioactive wastes, but they do




offer a perspective.  Our estimate of the risk from the wastes for 10,000




years after disposal is less than the current population risks from




generating the electricity.









3.1.5  Risks from Nuclear Weapons Fallout




     Fallout from atmospheric testing of nuclear weapons has added to  the




radioactivity of air, water, food, and residual radionuclides  (primarily




tritium, strontium-90, cesium-137, carbon-14, and plutonium) and  is a




source of continual radiation exposure.  UNSCEAR has estimated the dose




commitments to various organs of an average  individual  in  the  northern




temperate zone from nuclear weapons testing  through 1975  (UN 77).  The




UNSCEAR values are shown in Table 3-2.  When converted  to  millirems,  the




dose commitments are 150 millirems to gonads, 260 to bone  marrow,  300  to




bone linings, and 270  to lung.  Almost all  the dose commitment is due  to




relatively short-lived (30 years or less) nuclides, because  doses for




carbon-14 were calculated only  to the year  2000.






                                   45

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                              Table 3-2

                Dose Commitments  from Fallout  (UN 77)
External
Short-lived
137CS
Internal
14C (a)
90Sr
106Ru
137Cs
144Ce
239Pu
TOTAL ^
(Northern Temperate Zone)
tnillirad
Bone Bone
Gonads Marrow Lining

48 48 48
62 62 62

7 32 29
84 120

27 27 27

1
150 260 290
                                                               Lun
                                                                 48

                                                                 62
                                                                 41

                                                                 27

                                                                 65

                                                                  1


                                                                260
     accumulated up to the year 2000
Including nuclides not listed
                                46

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     We estimate the number of projected fatal cancers in the United




States population from fallout by multiplying the average dose committnents




in Table 3-2 by 200 million and by the appropriate risk factors.  We have




assumed that the soft tissue dose is equal to the gonadal dose and have




reduced ttie whole body risk factor of 200 per million person-rem to 110,




to avoid counting the fatal cancers from the listed organs twice.  The




expected harm from nuclear weapons testing in the United States is then




about 3300 fatal cancers from soft tissue irradiation, 2080 from bone




marrow irradiation, 600 from bone lining irradiation, and 2160  from lung




irradiation, for a total of 8140 fatal cancers over the life of the




nuclides involved.  If this number is averaged over the half-lives of




cesiutn-137 and strontium-90, the largest contributors to the dose




commitment, the result is an average of about 190 fatal cancers per year




over about 43 years.









     The risks from nuclear weapons fallout are primarily near-term risks




and are relatively uniformly distributed over the northern hemisphere.




Therefore, they are not directly comparable to the risks  from  the  disposal




of waste.  The present annual risk from fallout  is about  2000  times




greater than the projected  future annual risk from the wastes  after




disposal.
                                   47

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  3.2  ASSESSMENT AND REDUCTION  OF  RISKS




      An  important  element  in the  development  of  environmental  standards  is




  an assessment of technology to control  risks.  In  this discussion, we deal




  with the quantitative assessment  of risks,  including  the effect of




  uncertainty, and consider  some methods  suggested for  the disposal of




  high-level and transuranic radioactive  wastes.









 3.2.1  Assessment of Risks from Waste Disposal




      No high-level or transuranic radioactive wastes have ever been




 disposed of permanently.  Because there is no background of experience




 that  can be used as a basis for analysis,  we must rely on modeling




 procedures  for an assessment of risk.   To  model the characteristics and




 performance of a disposal system,  mathematical equations must be developed




 that  predict  potential  releases of waste and describe their subsequent




 transport through  the environment. These  models  predict  the consequences




 of releases over  long periods of time.   The models  can be no better than




 our understanding of all the processes  involved and,  in fact,  are  often




 simplified  to facilitate the complex computations.









     The mathematical equations describing  the  behavior of  the  system  and




 the movement  of radionuclides through the environment  to man are designed




 to be general, and  they  require appropriate  numerical  data  to produce




 useful results.  Therefore, there  are uncertainties both  in  how well the




models describe the actual physical and  chemical processes  that may occur




and also in how good the numerical input data are in terms of producing




realistic results.
                                   48

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     When attempting to model the behavior of a generic system, we must




select numbers descriptive of system characteristics.  It is customary to




make several model assessments, using several different sets of values for




the characteristics we consider important.  This practice, which is




generally called sensitivity analysis, gives us information on the extent




to which each system characteristic influences the overall estimate of




risks.









3.2.2  Reduction of Risks by Disposal Technology




     Although several methods  of isolating high-level  and transuranic




wastes are being developed, we have chosen to  assess disposal  in mined




geologic repositories.  We believe that  this  is the  only method for which




sufficient information  is available to provide the basis  for  judgments




needed to develop disposal standards.  Any other  method  for disposal  of




high-level and  transuranic wastes must result  in  risks no greater than




those we have judged would result  from disposal in carefully  selected and




well-engineered geologic  repositories.









     As part  of the development  of  these proposed standards,  EPA  asked the




MITRE Corporation  (AL  79a) to  review  the scientific  and  developmental




status of alternative  disposal concepts.  The Interagency Review  Group on




Nuclear Waste Management,  created  by  the President,  and  the Department of




Energy have also  examined alternative disposal technologies (IRG  79),




 (DOE 80).   The  Ford Foundation (NU 77)  and the American Physical  Society




 (APS 78)  have previously reviewed high-level waste disposal concepts, and
                                    49

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 several  other groups,  including an Ad Hoc Panel of Earth Scientists




 commissioned  by EPA (EPA 78a),  nave analyzed aspects of the status  of




 basic  science and  technology relevant to deep geologic disposal.  The




 following  paragraphs describe  the status of waste disposal  concepts,  based




 on  information that is  currently available.









     Emplacement in Deep Mined  Repositories




     Radioactive wastes  could be placed  in underground repositories built




 with conventional raining techniques.   The primary natural barriers  to




 migration  of  the radionuclides  would  be  the emplacement medium and  the




 surrounding geologic and hydrologic environment.   Engineered barriers




 would  include  a stable and  insoluble  waste form,  corrosion-resistant




 containers, and absorbent packing materials.   Removal of the material from




 the immediate  proximity  of  people and other living things would reduce  the




 possibility of accidents during  the several-hundred-year period during




 which  fission  products dominate  the radioactivity.   Careful attention to




 site selection and  deep  emplacement would  make  catastrophic release of  the




 waste  by severe surface  disturbances  extremely  unlikely.








     Our assessments indicate that  exploration  for resources,  in which




 holes would be drilled into  the  repository, would  be the  most  common  mode




 of inadvertent release of radionuclides  to  the  biosphere.   Drilling could




 release radionuclides to both the land surface  and  groundwater.  While




groundwater would be a pathway by which  radionuclides  from  the  waste  can




enter the biosphere, it would also  be a  protective  barrier.  Groundwater
                                   50

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velocities are slow enough, in general, that many of the radionuclides in




the waste would decay before reaching the surface.  Furthermore,




adsorption of many radionuclides by the media through which the flow




occurs can slow their movement even more and can provide considerable




added protection against exposure to radionuclides with moderately long




half-lives.








     Other Methods of Deep Continental Disposal









     Rock Melt.  Fresh high-level waste might be  emplaced  into  continental




rock formations. The heat  generated  by the  radioactivity of  the waste




would melt the  surrounding rock.  When the  rock  cooled,  the  wastes would




be  incorporated in a rock  mass  that  would be  impermeable  and in good




chemical  equilibrium with  its environment.








     Rock melting disposal methods may offer  the same protection against




catastrophic  releases  by natural forces  and human intrusion as deep-mined




geologic  repositories.   There would, however, be considerable




volatilization of  radionuclides during the melting period, after which




retention of  the  radionuclides  would depend on the extent that they were




bonded  to the rock  during melting.   There has been no evaluation of  the




extent  to which radionuclides would be immobilized in this disposal




concept,  and, therefore, we cannot make an assessment of risk.
                                    51

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     Deep  Holes.   Containers of waste could be placed in the bottoms of




holes  drilled  as  deep  as 50,000 feet.   At  such depths,  human intrusion




would  be unlikely.   The  emplacement  of wastes to this depth is not




feasible now.  Present drilling methods are believed  to be  capable  of




drilling a hole 20  centimeters  (about  8 inches)  in diameter to a depth of




about  11,000 meters  (35,000  feet)  or a hole 31 centimeters  (about 12-inch)




to a depth of  9.1 kilometers  (about  30,000  feet).   The  ability to seal




such a hole has not  yet  been  demonstrated.








     Risks  from disposal  in deep holes  would  depend on  the  ability  of the




rocks around the hole  to  dissipate the  heat and  on the  absence of




groundwater movement in any specific location.   Information for such risk




assessment is not now available.








     Emplacement in Deep Ocean  Sediments




     Containers of waste or spent  fuel  could  be  placed  tens of meters or




more below the ocean floor in stable regions  where  the  sediments  are thick




and uniform.  The wastes would  be  far  from  human activity,  and the




sediments would be expected to  prevent  the  waste nuclides from reaching




the ocean water.








     Lack of knowledge of the stability and adsorptive  properties of the




sediment barrier when wastes are emplaced in  it  makes it impossible  to




assess the risks of this method of disposal.  Disposal  in the  ocean  is
                                   52

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prohibited under United States law by the Marine Protection, Research and




Sanctuaries Act of 1972.  Legislative action would be required before




seabed disposal could be implemented.








     Extraterrestrial Disposal




     High-level and transuranic radioactive wastes could be launched into




space.  This method of disposal would potentially eliminate future risk to




living things.  A solar orbit appears to be the most technically and




economically achievable option and would provide long-term  isolation.




This method would not provide for injection of all the waste  into space,




because costs would be too high.  Present concepts would require chemical




separation of the wastes and would send only the long-lived radionuclides




into space .









     Disposal in space would permanently isolate radioactive  nuclides  from




humans.  However, methods for assuring the protection of public  health in




the event of  launch failures have not yet been developed,  and separation




of wastes introduces added risks.  The overall costs of  this  method  of




disposal are  high.









     Transmutation of Waste Radionuclides




     Radionuclides in waste could be transformed  to  less hazardous




nuclides by  irradiation  in nuclear reactors  or  in  particle accelerators.




Long-lived and  highly radiotoxic actinide radionuclides,  including




neptunium-237,  plutonium-239  and -240, and  araericium-241 and  -243,  are the
                                    53

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most important candidates for transmutation.  However, other long-lived




nuclides, such as iodine-129 and technetiura-99, have also been




considered.  Materials to be transmuted would have to be chemically




separated from the irradiated fuel.  High-efficiency actinide separation




has not yet been developed.








     Transmutation of nuclides would eliminate any possibility that they




would ever affect people.  However, other risks of transmutation have not




been fully evaluated and may be high.  Separation and recycling of the




actinides and other radionuclides could increase occupational,




transportation, and routine release risks.









3.3  DEALING WITH UNCERTAINTIES




     In considering the possible harm that might be incurred in the future




by people as a result of releases from a disposal system, we must allow




for uncertainties in our expectations of the future.  In our development




of this proposed action, we did this in two ways: by using conservative




values in our assessments and by including the qualitative assurance




requirements in our proposed standards.  In the following paragraphs, we




will discuss the issues of uncertainty in disposal system performance and




of future human behavior.









3.3.1  Technical Uncertainties




     Disposal systems limit potential harm to people from the disposal of




high-level and transuranic radioactive wastes by placing barriers between




the wastes and the human environment.  Judgments of the acceptability of






                                   54

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the system depend on estimates of the effectiveness of these barriers.  We




can make such estimates by considering the performance expected from each




barrier, using conservative values, and combining these into an assessment




of the performance of the entire system.  We allowed for uncertainties in




these estimates by selecting values for the effectiveness of each barrier




that we are confident are well within the capabilities of engineering and




site characterization.  The actual effectiveness of each barrier will




probably be considerably better than our selected value.  The entire




system should perform much better than our assessment of its performance,




which we obtain by combining our values for the effectiveness of all  the




barriers.









     We also allowed for uncertainty qualitatively by providing for




redundant independent barriers.  Each barrier would function whether  or




not other barriers perform in the manner we expect.  We believe such




redundancy and independence is particularly appropriate for a system  that




must operate over long periods of time without human control or




intervention.  Accordingly, we have adopted both the conservative  system




analysis, which we used to select limits on the quantities  of




radionuclides projected to be released  into the accessible  environment,




and the qualitative concept of redundant independent barriers  included  in




our assurance requirements.









     We could also have required that the disposal  system meet our




containment requirements even if some of their barriers are assumed to




fail.  Such a requirement might be  stated as:  (L) meet the numerical






                                    55

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 requirements even if any one of the barriers  fails, or (2) meet the




 requirements if only one of the barriers works.  A repository built to




 comply with such requirements probably would  be much more expensive than




 other designs we have studied, and much of the necessary technology does




 not appear to exist yet.  Although some additional protection might be




 provided by such an approach,  we do not believe this would justify the




 extra costs,  difficult implementation, and long delays that would result.









      We must  also consider the possibility of presently unknown processes




 and phenomena that could produce much greater risks than we project from




 the system.   Development of technical and scientific knowledge has often




 revealed such unexpected processes or phenomena.   Our knowledge of




 ionizing radiation itself is  less  than a century  old.   In geology, the




 theory  of tectonic  plate movement  is  quite recent.   Therefore, to allow




 for such possible  major  technical  uncertainties,  we believe that we have




 an  obligation  to  leave future  generations  the option of removing the




 wastes  from a disposal system  found to be  excessively  hazardous.




 Accordingly, we are proposing  an assurance  requirement that requires




 disposal  systems  to be designed  so that  recovery  of most  of the  wastes is




 not  precluded for a reasonable period  after disposal.









     Some disposal methods, such as deep-hole  disposal  and  rock-melting




 concepts, may not be able to comply with this  assurance requirement.




 If we did not include this provision,   it might be possible  to  develop




disposal methods that would have smaller projected  risks  than  those we
                                   56

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project for mined geologic repositories.   However,  if we were to dispose




of wastes in a way that could not be reversed, then future generations




would not have the option to take advantage of new information or better




technologies.  We believe that mined geologic disposal methods, from which




wastes can be recovered, can keep the projected risks to future




generations small, and we judge that any additional protection is less




important than retaining future options to alter the disposal method, if




necessary.









3.3.2  Uncertainties in Human Behavior




     Human behavior in the future may affect the risks from disposal




facilities in two ways.  First, human actions may change the ability of




the system to isolate the wastes (for example, if a person drills through




a geologic repository while exploring for some mineral resource, he may




provide a pathway through which radionuclides can reach people).  Second,




human actions may affect the likelihood of human exposure to radionuclides




that have escaped from isolation (for example, exposure to radionuclides




in contaminated groundwater occurs only as a result of using the water).









     Institutional Controls




     Institutional controls are intended to assure the  integrity of  a




waste isolation repository and to eliminate the possibility  of human




damage to the isolation capacity.  The pertinent  issue  regarding




institutional controls is how long we should rely on  them in our planning




and design of disposal facilities.  Different attitudes toward
                                   57

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 institutional controls can lead to very different strategies for




 protecting the environment.  This issue was discussed extensively during




 the development and review of the Federal radiation protection guidance




 for disposal of all types of radioactive wastes that we proposed on




 November 15, 1978.   Institutional control was defined as "activities,




 devices,  and combinations thereof which involve the performance of




 functions by human  beings to limit contact between the waste and the




 humans  or the environment."   The proposed criteria stated ".  . .  Controls




 which are based on  institutional  functions should not be relied upon for




 longer  than  100 years  .  .  .".









     Public  comments with  specific recommendations about how long  we




 should  rely  on  institutional controls were divided approximately evenly




 among four positions:




     1.   That  institutional controls should  be relied upon  for only about




          20  to  30 years.




     2.   That  the 100-year period was  about  right.




     3.   That  the 100-year period should  be  extended  to 500 to 1000 years,




          or  even longer.




     4.   That we should limit reliance on controls,  but let the




          regulatory agencies select the appropriate  time  period.









     On balance, a few more commenters  felt that the  100-year period  was




too short compared to those who felt it was too  long.   Several  other




comments made a distinction between "active" controls,  such  as  restricting
                                   58

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access to disposal sites, and "passive" controls, such as warning future




generations about what we have done by creating extensive records and




markers.









     In the proposed assurance requirements that are part of these




standards, we have decided to limit reliance on "active" controls—such




as guarding a disposal site, performing maintenance operations, or




controlling or cleaning up any releases from a site—to a "reasonable"




period of time after disposal, which we believe should be no more than a




few hundred years.  However, because the Federal Government is committed




to retaining control over these disposal sites in perpetuity, we expect




that "passive" institutional measures should substantially reduce the




chance of inadvertent human intrusion well beyond this period.  Such




passive controls will include permanent markers placed at a disposal site,




public records or archives, Federal ownership or control of land use,




and other methods of preserving knowledge about the disposal  system.




These passive controls should not be assumed to prevent all possibilities




of inadvertent intrusion, because there is always a chance that the




controls will oe overlooked or misunderstood.  However, such  measures




should be effective in deterring systematic or persistent exploitation of




a disposal site.









     As one alternative to this proposal, we could have assumed that




institutional controls such as those we now use would be effective for a




very long time (thousands of years).  Inadvertant human intrusion would
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 not be a concern,  and few areas would have to be avoided because of




 resource potential.   This would permit use of monitored, engineered




 storage systems—carefully designed to reduce the chances of accidental




 releases—which would have a number of advantages over disposal systems.




 In particular,  failures of the waste containment could be quickly detected




 and corrected  before  significant contamination of the accessible




 environment  occurred.









     On the  other  hand,  we could have assumed that  society could lose  all




 knowledge  of the disposal  system and the  dangers of its contents in a




 relatively short period  of time.  Thus, not  only could people  intrude  into




 the system,  they might not  respond  to any unusual conditions they




 encountered  and might continue  to spread  the wastes throughout  the




 environment  for a  long time.   In  this case,  the  first  priority  for a




 disposal method would be to  reduce  the chance of inadvertant intrusion as




 much as possible,  even if  this  resulted in higher risks from other




 events.  This would strongly affect  the relative attractiveness  of various




 disposal methods.  Techniques such as  deep-hole  placement,  which removed




 the wastes far from man's  typical activities,  would  be  favored.   We also




might have to delete our assurance requirement that  the wastes  be




 recoverable,  to avoid ruling out many  of  these disposal  techniques.









     When there are no active institutional controls,  the  likelihood of




inadvertant human  intrusion into the disposal area may  be  reduced  if the




existence of  the disposal site  is known.  The waste disposal system can
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provide such information in the form of archival records and permanent




markers at the site.  In the public workshops leading up to our




November 15, 1978, proposal of Federal radiation protection guidance for




radioactive waste disposal, there was general acceptance of the concept of




passive communication of the disposal system (EPA 78b).  Over long periods




of time, any particular marker or record may be destroyed or lost, but




wide distribution of duplicate records would greatly reduce the likelihood




of all knowledge of the disposal location being lost.  Accordingly, we




have included an assurance requirement for markers and records.








     Resource Potential




     We must consider the possibility of human activities that would



degrade the isolation capability of a waste disposal system in the absence




of active institutional controls or of knowledge of the disposal system's




existence.  Human intrusion is more likely if the disposal site can be




expected to yield valuable mineral, energy, or other resources.  Locations




that are now known to contain actual or potential resources, such as coal



or oil shale, must be avoided.









     We must be cautious in our judgments as to what would be attractive




to future prospectors, because we cannot predict what materials will be



valuable to people in the future.  A few centuries ago oil, uranium ores,




and aluminum ores were not significant resources.  In our assurance




requirements, therefore, we state that wastes should not be placed in an




area where there is a reasonable expectation of future mining or
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 exploration.  To allow for our uncertainty about what materials may be




 considered valuable in the future, we extend this requirement to sites




 containing a substantial concentration of any material that is not widely




 available.








 3.4  CHOICE OF FORMAT FOR DISPOSAL STANDARD




      In developing the proposed standards, we considered a number of




 possible ways to limit risks  to individuals or population groups.   We also




 considered a number of possible time periods over which we should




 regulate.   This  section and the following one evaluate these alternative




 approaches to the  disposal  standards.









     Alternative 1;  Standards  expressed  in numbers  of health  effects




 projected  over some  period  of time.  This form  most  directly addresses the




 population risks and would  be preferred,  all  other  things being  equal.




 However, regulations in this form  require assumptions  on  the exact




 relationship  between dose and health effects  and  would also  require




 knowledge  of  future  population  and demography at  a site,  an  obvious




 impossibility.  We therefore rejected Alternative 1.








     Alternative 2;  Standards  expressed  in terms of population  dose




equivalents (person-rems) over  some period of time.  This form requires  no




assumptions as to the nature of the dose  response (health effects per




person-rem) but is otherwise similar to Alternative  1.  It would require




the same knowledge of future population and demography as Alternative  1,




and therefore we rejected it also.






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     Alternative 3:  Standards expressed in dose-equivalent to individuals




over a lifetime.  This form would provide an estimate of the total risk to




an individual.   However,  this alternative might not assure adequate




protection to the total population, and it appears very difficult to




implement.  We therefore rejected this alternative.









     Alternative 4;  Standards expressed in terms of dose equivalent rates




(rem per year) to  individuals.  This form measures risk in terms of annual




increments of exposure and is more useful in regulating short-term hazards




than for controlling long-lived radionuclides.  However, this alternative




(or equivalent standards which limit radionuclide concentrations in air or




water) is a traditional form of radiation protection standard.  Therefore,




although we rejected this option,  it is further examined in Chapter 4  as




one of the major alternatives to our proposed  action.









     Alternative 5;  Standards expressed  in terms  of total  integrated




radionuclide releases  (curies) to  the  accessible environment  over  a  long




period of time.  This  choice would serve  as a  surrogate  for Alternatives  1




and 2 in estimating population risk.   Standards  in this  form  would be




easier to implement, since they only require  estimation  of  the  probable




quantities ot radionuclides  leaving the  disposal  site  and  their physical




transport to a  location accessible to  man.  For  these  reasons,  we selected




this  form for the  standards.
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      Alternative 6;  Standards expressed in terms of radionuclide release




 rates to the accessible environment (curies per year).  Our investigations




 (SMC 82) snowed that radionuclide release rates were less closely related




 to population risk than total radionuclide releases.  We rejected




 Alternative 6 for this reason.








      Alternative 7;   Standards  expressed in concentration of radionuclides




 (curies  per cubic meter)  in environmental media.   Concentrations of




 radionuclides in specific  environmental  media  are less closely related to




 population  risk than total curie  releases,  and standards in this form




 could encourage siting  where  there  are large volumes of diluting water.




Although this could  reduce doses  to exposed individuals,  the ultimate




population  dose would  be the  same.   The  calculations required  for




 implementation of  this  standard would be much  more difficult,  because they




would require detailed  analysis of  environmental  pathways.   We therefore




rejected  this alternative,  but—as  discussed for  Alternative 4 above—this




overall  approach  is  considered as one of the major alternatives to  our




proposal  in Chapter 4.








3.5  ALTERNATIVE REGULATORY TIME PERIODS




     The goal  of our disposal standards  is  to  provide  protection as long




as the wastes  present an unacceptable risk.  The  standards  require  a




reasonable expectation  that releases over a  specified  time  will be  less




than the specified limits.  Stating  a period for  regulation in the




standard does not  imply that the radioactive wastes  will  be released  after
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that time; it establishes the period over which the releases must be




assessed.  We considered several alternative time periods for assessment:




100 years; 1000 years; 10,000 years; 1,000,000 years;  and forever.  We




could also have based our assessment on the risk to any future generation.








     Alternative 1;  100 years.  The Uranium Fuel Cycle (UFC) regulations




(40 CFR 190) are based partially on 100-year projections of risk.  In




those regulations, we limited the projections to 100 years because of




uncertainties in population, demography, and use of the environment.  We




do not need precise estimates of risk for development of these proposed




high-level and transuranic radioactive waste standards, because balancing




these risks and costs is not a major basis  for the standards.  Our risk




assessments showed that few health effects  are predicted in the first




hundred years, mainly because of long groundwater transport times and  the




low probability of disrupting events in such a short time period.




Therefore, an analysis of consequences limited to 100 years after disposal




would give an unrealistically low estimate  of the total potential




impacts.  We rejected this alternative for  these reasons.









     Alternative 2:   1000 years.  Some have suggested that most risks




occur in  the first 1000 years, because strontium-90 and  cesium-137  decay




in that time.  The same considerations that argue against  the 100-year




period apply to the 1000-year period.  Few  health effects  occur during




this period; many nuclides would still be  in  transit  towards  the
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 accessible  environment;  disruptive  event  probability is  still  low;  and  a




 considerable  number  of  the  important  nuclides  would not  have undergone




 significant radioactive  decay.  We  therefore rejected Alternative 2.









     Alternative 3:   1,000,000  years.  A  period  of  1,000,000 years  has




 been advocated  on  the grounds that  the hazard  from  high-level  radioactive




 wastes  is not adequately reduced  by radioactive  decay until then.




 However, it would  be difficult  to make quantitative assessments  of  the




 potential environmental  impacts of  waste  disposal over such a  long  time,




 and prediction  of  geological changes over such a long period is  not




 reliable.  These reasons make Alternative 3 unacceptable.









     Alternative 4;  Forever.   The  arguments against  selection of the




 1,000,000-year period apply even more strongly to regulation over all




 time.  This option, too, was rejected.









     Alternative 5;  10,000 years.  We have selected  10,000 years as a




 regulatory period.   In 10,000 years, many  of the radionuclides that can




 pass through groundwater transport paths will  have  reached surface water




 and the accessible environment.   At 10,000 years, the radiological hazard




 of the wastes would be substantially reduced through  the decay of most of




 the significant fission products and of many of  the actinides.  The only




 remaining nuclides presenting a potentially significant health risk would




 be technetiura-99 and iodine-129 and the actinides plutonium-239  (about




75 percent of the original inventory), plutonium-240, and neptunium-237.
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This inventory will decrease fairly rapidly after 10,000 years by decay of




plutoniura-239 and -240; it will then decrease slowly because of the long




half-lives of technetiutn-99, iodine-129, and neptunium-237.  The




radionuclide hazard potential at 10,000 years is, therefore, fairly




descriptive of the situation for a long time thereafter.  Population and




demographic conditions 10,000 years from now are unknown and unknowable,




but they do not have a controlling effect on setting the standards.  Since




10,000 years is a short time geologically, changes in geological




conditions are expected to be small.  Only massive changes  in climate,




such as could conceivably be induced by man-made additions  to the




atmosphere (e.g., the carbon dioxide "greenhouse" effect ), could produce




large changes in the transport of radionuclides.  The adverse effect of




these climate changes would be so great that any additional impact of  the




radioactive wastes would be relatively  insignificant.









     Alternative 6;  A reasonable objective for waste disposal  is  that no




future generation should incur more than a fraction of  the  risks  that




would be acceptable to the current generation.  We have considered this




principle in formulating these standards., However, we  also recognize  that




it could be difficult to implement as stated.  The risks to each




generation for a long period of time would have  to be examined  to  find out




if they were below the appropriate level.  Our evaluations  of the




environmental effects of disposal, described in Chapter 5,  indicate  that




total risks over 10,000 years can be made small.  Limiting  total  risks to




this level ensures that no single generation within this period can  incur
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greater risks.  These risks do not appear unreasonable even if incurred by




only one generation.  Therefore, we believe it is simpler and adequate to




restrict risks over the total period of 10,000 years rather than provide a




limit on risks to a single generation.
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                                 Chapter  4




                               ALTERNATIVES









     In developing our proposed environmental standards, we considered a




number of alternatives.  This chapter examines those alternatives that




would involve major changes to our proposed action.  Additional




alternatives that would affect specific details of the standards are




discussed in other chapters of this Draft Environmental Impact Statement:




Chapter 3 examines some different approaches we could take to compensate




for the uncertainties of the future and also considers different time




periods and alternative units for the disposal standard, and Chapter 7




considers alternative dose limits for the waste management standards.




Each of our discussions of alternatives  indicates why we think our




proposed action is preferable to the other options we considered.




However, in each case we would like to receive public comments that




evaluate whether we made appropriate choices.









     Our consideration of alternatives is influenced by the  limits on our




information.  High-level radioactive wastes have never, to our knowledge,




been permanently disposed of.  Furthermore, since we believe these




disposal systems should isolate  the wastes for many thousands of years,




the relative benefits of different systems can be  judged only by




predictions of their long-term performance.  With  this  pervasive




limitation in mind, the following are the major alternatives to our




proposed action.
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4.1  ISSUE NO  STANDARD

     The decision  to prepare  these  standards was  an administrative action

taken by EPA and was not mandated by  law.  We were directed  to prepare

standards as part  of President Ford's Nuclear Waste Management Plan on

October 27, 1976.  President  Carter established an Interagency Review

Group (IRG) on Waste Management  in  March 1978 to  review existing programs

and recommend new  policies where necessary; the IRG recommended that EPA

set standards  for  nuclear waste management and disposal activities and

accelerate its programs to do so.   President Carter approved  this

recommendation as  part of his Program on Radioactive Waste Management

announced on February 12, 1980.  The NRC best described (NRC  80) the

expected goal of these standards:

         ".  .  . (EPA) standards represent a broad social consensus
     concerning the amount of radioactive materials and levels of
     radioactivity in the general environment that are compatible
     with protection of the health  and safety of  the public."


     Although satisfactory management and disposal of high-level

radioactive wastes could be attained by other regulatory means (e.g., NRC

licensing requirements), such means would not likely receive  broad public

acceptance.   By promulgating  standards, EPA directly focuses  on

environmental protection requirements without the constraints of many of

the complex engineering requirements necessary for consideration in

licensing.   Also,  because of  its broad environmental charter, EPA is in

the best position  to assure broader input into the standards  by groups

outside nuclear interests—e.g., those charged with protecting

groundwater.  Thus, not issuing standards is an unacceptable  option.
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4.2  DELAY ACTION




     We could have chosen to delay issuing standards until we developed




more information about disposal systems.  This delay could allow us to




better evaluate how well disposal systems can protect the environment and




might lessen the chance of poor judgments of how much protection is




necessary and reasonably achievable.  We might then be able to develop




information on disposal systems other than geologic repositories and to




develop quantitative comparisons of the costs and benefits of these




alternatives.  However, we do not expect enough information to become




available during the next decade to allow more comprehensive evaluations.









     We choose to propose these standards now, because we believe that




environmental standards must be developed to guide the progress of the




national waste disposal program.  DOE has decided to focus the national




program on mined geologic repositories (46 FR 26677), after evaluating




alternatives through a comprehensive environmental statement process




(DOE 80).  Ttie DOE program anticipates identification and characterization




of three sites at which to begin exploratory shaft construction in 1983.




The overall goal is to provide the first licensed, fully operational




repository within the period 1997 to 2006 (NE 81).









     Delay in proposing these standards could diminish public confidence




in the national waste management program.  Loss of confidence might  lead




to action by more state governments to restrict exploratory and research




activities.  This could reduce the options available to the Federal
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 government in exploring for repository sites.  We believe that proposing

 standards now will enable the national radioactive waste disposal program

 to proceed on an orderly schedule.  This is consistent with the synopsis

 of the public comment received by the IRG (IRG 79):

          "Comment from both the industrial sector and the
      environmental community urged the acceleration  of EPA standards
      particularly to instill confidence that proper  protection of
      the public's health and safety is being provided.  They
      expressed the concern that early standards are  essential to
      permit  the waste management program to proceed  expeditiously."


      In addition,  proposing these standards now will reduce the chances of

 harm  from long-term storage of existing wastes in surface facilities.

 Although such storage offers no significant danger under normal

 conditions,  the wastes  are more vulnerable to accidental release.   For

 example,  several  leaks  have occurred  from the high-level waste storage

 tanks  at  Hanford  (ERDA  75).  These leaks  have not caused any exposures of

 the general  population, nor have they contaminated any areas beyond the

 site  boundary.  However,  they  do represent unplanned releases that  could

 have had  serious consequences.   We believe that our  standards will

 adequately protect  public  health and  expect  that  disposal technologies

 satisfying these standards  will  be available within  the next decade.



 4.3  ESTABLISH ONLY QUALITATIVE  REQUIREMENTS

     The  large uncertainties inherent  in  predicting  the performance of

 disposal  systems over thousands  of years  present  an  argument  for  issuing

 environmental protection requirements  that contain only qualitative

criteria—such as our proposed assurance  requirements.  Application of
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such criteria would be easier than implementing our proposed standards:




regulation on the basis of qualitative criteria would not require any




projection of releases over time, and the validity of models for




predicting those releases would not come into question.  Instead,




selection and licensing of disposal methods could rely more upon the




informed judgments of technical experts (subject to appropriate review)




than upon uncertain numerical estimates.  Issuing only qualitative




requirements would still provide instructions that could effectively




reduce the dangers of disposal, and such criteria could be used to compare




alternative disposal systems and to select among them.  Since such




qualitative requirements—by themselves—would not be appropriate




generally applicable environmental standards, we would issue them under




our Federal radiation protection guidance authority.









     Our quantitative environmental protection requirements provide  at




least two important benefits, however:









     1.  By selecting the form of these protection requirements  as




     numerical release limits over a particular  time,  we  identify  those




     overall objectives that should be considered  in disposal  system




     design.  (One example: setting stringent  release  limits rather  than




     individual exposure limits  encourages very  good containment rather




     than suggesting dilution as part of a disposal  strategy.  Another




     example: a disposal system  designed to  limit  releases  for  1000  years




     could rely primarily on engineered barriers,  whereas a system
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      designed to retain wastes for 10,000 years also would require both




      good geological and hydrological characteristics at the disposal




      site.)









      2.   Setting quantitative standards requires a full assessment of




      system  performance to assure that these standards will not be




      exceeded—and this requirement for a comprehensive examination of the




      potential harm can provide confidence that disposal systems will




      behave  as expected and will not  create more than a small risk to




      public  health.








4.4  DELETE  OR DEEMPHASIZE  THE  QUALITATIVE  ASSURANCE  REQUIREMENTS




      In contrast  to  the previous  alternative,  this approach would increase




the reliance  on our  quantitative  containment  requirements.   It  would




provide greater flexibility for selection  and  design  of disposal systems—




since the emphasis of our action  would be  on achieving specific




environmental  protection objectives without restrictions on how those




objectives should  be met.   (For example:  system designers might choose to




use a single,  extremely effective barrier,  or  they might decide to depend




on indefinite  control of the  site to meet our  requirements.)  As part of




this alternative approach,  the  NRC  could  be expected  to provide any




appropriate procedural, site  selection, or  system  design requirements that




might be needed beyond  our  quantitative standards.
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     There are two variations of this approach that we considered:




 —  to simply delete the qualitative assurance requirements, leaving only




     the containment requirements and the procedural requirements needed




     to assess compliance, or




 —  to issue the assurance requirements as Federal radiation protection




     guidance, rather than as part of the generally applicable




     environmental standards.




     The second variation would make the qualitative criteria less binding




since—as executive directives rather than formal rules—they could be




waived if specific situations warranted.  Such waivers would be considered




in accordance with the procedures established under Executive Order 12088.









     The concept of allowing maximum flexibility to achieve  specific




environmental protection goals (saying "what to do", not "how to  do it")




can be very effective in many situations.  However, we do not believe we




should rely upon it in this case because of the unusually large




uncertainties inherent in designing disposal systems to isolate wastes  for




at least 10,000 years.  Instead, we think the cautious steps called for by




our assurance requirements should be an integral part of our standards  in




order to develop appropriate confidence that our containment requirements




will be complied with.









     Altnough the specific design requirements that will be  developed by




the NRG could meet the objectives of our assurance  requirements,  we




believe that our standards should be expected  to establish the  complete
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 framework of quantitative and qualitative environmental protection




 requirements that are needed.  Furthermore, one category of wastes covered




 by these standards—transuranic wastes from national defense




 activities—is not subject to NRG regulation.









 4.5  SELECT CONTAINMENT REQUIREMENTS ON A DIFFERENT BASIS




      We have proposed containment requirements that we believe are




 achievable by well chosen and carefully designed geologic repositories and




 that  would limit  harm to future generations to low levels.   We could have




 used  different bases for selecting these release limits.   In the following




 paragraphs we examine some alternative approaches  for picking the




 containment requirements.









 —   Develop  Projected  Release  Limits  Considering  Different  Disposal




      Methods




     An analysis  of  the  potential  performance  of several  disposal methods




 besides mined  geologic  repositories might  allow  us to set more restrictive




 limits.  These alternative disposal methods are  described in Chapter 3.




However, we cannot do the necessary analyses now or in  the near future,




because the needed information  for any other disposal method will not  be




available  for a long time.  Therefore, since this  alternative would




significantly delay  these standards, it  is essentially  the same as the




"Delay Action" alternative discussed previously.
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 —  Limit Risks to Future Generations from the Entire Nuclear Fuel Cycle,




     including Waste Disposal,  to Those That Would Exist from the Original




     Unmined Uranium Ore




     The Natural Resources Defense Council (NRDC) (CO 79) has proposed




this criterion for radioactive waste disposal. The underlying principle is




that nuclear power should cause no increase in risk to future




generations.  This alternative would involve consideration of all




long-lived wastes (not limited to high-level wastes) and effluents




associated with the nuclear fuel cycle.  It would also raise the question




of comparing risks to future generations from nuclear power with those




from any other sources of energy from which some residual risk is passed




on to future generations.  We believe this is beyond the scope of this




standards-setting effort and, therefore, have not directly used this




criterion in developing these standards.









 —  Limit Risks to Future Generations from High-Level Waste Disposal to




     Thoae That Would Exist from the Original Unmined Uranium Ore




     This alternative is related to the criterion proposed by the NRDC.




However, it is less restrictive since it considers only  one part of the




future risks from the nuclear fuel cycle.  The risks  from ore bodies vary




over a wide range, depending on local conditions such as solubility of




uranium in the local water, hydrological factors, and transport of radon




through environmental media.  There are also many uncertainties in our




ability to assess the risks from a particular ore body,  including




uncertainties in sampling, in measuring the characteristics of
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 environmental  pathways,  and in demography.   We estimated risks from




 unmined  uranium ore  ranging from somewhat  less than,  to one-thousand times




 greater  than,  the  risks  from disposal of the high-level wastes derived




 from  the ore (WI 80).









      If  we  used  this concept to  select containment  requirements,  the large




 uncertainties  would make  it difficult to decide where to set the  limits.




 We  believe  the risks from disposal  in systems complying with our  proposed




 release  limits will be no greater  than the  risks from most equivalent




 amounts  of  unmined uranium ore,  and most likely will  be less.   Therefore,




 we  did not  choose  this alternative, since  it would  not provide more




 protection  and would be a very uncertain basis for  regulation.








 4.6   SET HIGHER OR LOWER  RELEASE LIMITS




      For the release limits  we have proposed as our containment




 requirements,  the residual  risks projected  by our generalized




 environmental  pathway models would be less  than 1,000  premature deaths




 from  cancer over the 10,000  year period, an average of one premature death




 every 10 years.  To judge  the effects  on disposal costs of changing this




 level of protection, we also compared release limits  corresponding  to




 residual risk  values of:  100, 1000, 5000, and 10,000  premature deaths over




 the 10,000 year period.  We  chose this range of stringency levels because




 it appears  to  represent the  range of  performance that  may be expected of




mined geologic repositories.
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     To do this assessment, we evaluated the long-term performance  of




generic models of geologic repositories in three different geologic media:




bedded salt, granite, and basalt.   We did the assessment in two steps.




First, we used the performance projections decribed in Chapter 5 to assess




the quality of the engineering controls that would be needed in each of




the three model repositories to meet each of the four different levels  of




protection.  Second, we tried to allow for the possible effect of




alternative stringency levels on site selection.  This is particularly




relevant because our analyses indicate that the most important part of the




protection offered by a mined geologic repository comes from the




hydrological and geochemical characteristics of the site  itself.  These




calculations are further described in Chapter 9 and in our Draft




Regulatory Impact Analysis (EPA 82).









     The results of these assessments of disposal costs and alternative




stringency  levels indicate that the costs are not very  sensitive to




different  levels of protection, particularly for  the geologic media




(bedded  salt and granite)  that appear  to be better  at  reducing  long-term




risks.   The differences in costs for different  levels  are much  smaller




than  the overall uncertainties in waste management  costs.  For  example,




consider the  increased costs of complying with  the  release limits  we have




proposed,  rather than release  limits  10  times  less  stringent.   The




potential  increase  ranges  from zero  to 50 million (1981)  dollars per




year.  For comparison, the total costs of high-level  waste management  and




disposal (independent of  our action)  have been  estimated  as  between
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 700 million and almost 1.5 billion (1981) dollars per year.  Electrical




 utility revenues were about 100 billion dollars in 1980.   The possible




 impacts of this potential increase in disposal costs are  estimated to be




 less than a 0.2-percent  increase in the costs of nuclear  power and less




 than a 0.06-percent  increase in average electricity rates (EPA 82).









      These analyses,  while indicating that  disposal costs appear to  be




 relatively insensitive to differences in the level of protection,  do not




 provide a way  to determine the  acceptability of the residual risks from a




 societal  perspective, nor do they  indicate  a level of protection that is




 preferable from a balancing  o£  costs  and benefits.   One possible approach




 to  balancing costs and benefits  would be to judge the cost per life  saved




 by  different levels of protection,  perhaps  taking into account some  method




 of  discounting  costs and  benefits.  However,  our calculations of residual




 risks  are  not reliable as  absolute  values.  Thus,  we have no meaningful




 way  to  calculate  an absolute value  of the cost  per life saved by different




 levels  of  protection.









     In the absence of the ability  to make  meaningful cost and benefit




 comparisons, we have used  other  tests of  economic feasibility and




 acceptability of  risk to  judge the  appropriateness  of the level  of




 protection we have proposed.  As discussed  above,  setting the release




 limits  at  the level we chose—as opposed  to a  level  10 times less  or




 10 time's more stringent—appears to cause only  very  minor effects  on the




costs of high-level waste  disposal.  This is why  we  did not  choose higher




(less protective) release  limits.






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     To judge the acceptability of the remaining long-term risk, we




considered the risks that would otherwise be caused if the uranium ore




used to produce the wastes had not been mined.  As described in the




previous section, the magnitude of the risks from these unmined ore bodies




is very uncertain due, in part, to the wide variety of settings in which




uranium ore is found—many of which are closer to the surface than a




geologic repository would be.  Using the same generalized environmental




pathway models that were used to assess the risks from our models of




geologic repositories, the risks from a comparable amount of unmined




uranium ore are estimated to range from a few hundred to more than




1 million health effects over 10,000 years.  The lower end of this range




is roughly equal to the residual risk associated with our proposed release




limits.  Thus, the upper limit of the risk that our standards would allow




from the disposal of high-level wastes appears to pose a threat very  close




to the minimal risk posed by nature had the uranium ore never been rained




and the high-level wastes never been generated.  This is why we did not




choose lower (more protective) release limits.









4.7  SET DISPOSAL STANDARDS IN TERMS OF LIMITS ON MAXIMUM INDIVIDUAL




     EXPOSURE




     Individual exposure limits, or equivalent standards  that limit




radionuclide concentrations in air or water,  are a  traditional  form of




radiation protection standard.  Particularly  when  the limits are




comparable to or less than natural background levels, they may  be more




effective than our proposed standards at communicating how  small  the
                                    81

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 chance of harm from disposal of these wastes should be.  However, we chose




 not to use individual exposure limits in Subpart B because of two unique




 aspects of the situation regarding our disposal standards:









      First,  these disposal systems have to protect the environment from




 these highly concentrated radioactive wastes for much longer than




 institutional controls can be guaranteed to be effective.  Any individual




 exposure limit we set  could apply only at some distance from a repository,




 or it would  have  to ignore the risks from unplanned events such as




 inadvertent  intrusion—because individuals who fail to understand passive




 warnings  and  penetrate directly into or close to a disposal system




 (through  exploratory drilling for water or mineral resources,  for example)




 could receive very  large  exposures.   These exposures would probably exceed




 any  reasonable  individual  exposure standard.








      Second,  the  disposal  standards  have  to be  applied through analytical




 performance projections—implementing  such  standards  through  environmental




 monitoring and  potential remedial  actions  over  thousands  of years is  not  a




 credible  approach.  When we compared  the analyses  needed  for  compliance




 with exposure  limits to the analyses needed for compliance  with  release




 limits, we found  that  our  proposed disposal standards  would be much easier




 to implement  than exposure limits.  The NRC,  which is  responsible for




 applying our  standards for high-level waste disposal,  made  a  similar




evaluation and also found that  standards based  on  radioactivity  release




 limits could be implemented more readily than standards based  on  exposures




to individuals.






                                   82

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     The proposed standards,  although they do not directly limit doses to




individuals,  place requirements on disposal systems that make it unlikely




that many individuals will be unduly exposed.  The assurance requirements




are intended to reduce the chance that a person will intrude into the




disposal system.  Furthermore, our analyses show that the containment




requirements will limit the area of the environment that can become




contaminated and, thus, reduce the probability of significantly exposing




people who might inadvertantly disturb the lithosphere near the disposal




site.








     From these considerations, we believe that our proposed action will




better facilitate licensing of good disposal systems while providing




appropriate environmental protection from the long-term "risks presented  by




these wastes.  Therefore, we did not choose this alternative.









4.8  SET DIFFERENT LIMITS FOR RELEASES DUE TO NATURAL CAUSES AND FOR  THOSE




     CAUSED BY PEOPLE




     We could have set standards containing  two  sets of  containment




requirements: cumulative limits on releases  due  to  natural processes  and




events and limits on releases from each  individual  event caused by human




actions.  The separate containment requirements  would  then better reflect




the greater uncertainties involved in predicting human  behavior.








     For releases caused by natural processes and  events, we would use




numerical limits similar to those  in the proposed  standards.  The




probability of  these releases would  still  have  to  be evaluated  to






                                   S3

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 determine the limits on total releases from all such events over 10,000




 years.   Based upon the analyses described in Chapter 5,  we would make the




 release  limits for "reasonably foreseeable" natural processes and events




 10  times smaller than our proposed limits (which apply to both natural and




 man-caused  releases).   The limits  for "very unlikely" releases would not




 change.   The  risk associated  with  these  release limits would be about 200




 excess cancer deaths  over 10,000 years.









     For releases  caused  by actions  of humans,  we could  set numerical




 limits for  all  expected events or  for any single intrusion.  The analyses




 described in  Chapter  5  indicate that a single  intrusion  could cause  about




 50  to 100 excess cancer deaths over  10,000  years.   (This assumes that no




 one would clean up the radioactivity released  and that it would disperse




 through  the environment.)  The limits would apply to  those human actions




 the implementing agencies  judge  to be credible.   The  limits would not




 apply to recovery of the wastes  for  their resource value.  We assume that




 people knowing enough about the  wastes to want  to  recover them would also




 know about their dangers and would take  appropriate precautions.









     If  the limits on releases caused by  human  intrusion were stated for




 all such expected events over  the  10,000-year period,  the standard would




 simply be the sum of limits on releases  caused  by  natural events  and by




human intrusion, and the total  limits would  be  essentially unchanged.  The




only real difference would be  that the containment  requirements would have




two components, and each limit would  have to be met separately.   It  would
                                   84

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not resolve the uncertainty of the frequency of possible human intrusion.




This approach could introduce unnecessary complexity and would not provide




any additional public health protection.








     If the release limits for human intrusion were stated on a




single-event basis, it would be unnecessary for regulatory or implementing




agencies to estimate the number of intrusions that would be  expected over




the assessment period.  This would eliminate a major uncertainty  in




assessing the risks and in determining whether a  proposed disposal system




meets the containment requirements in the standards.   The separate release




limit standards would therefore be easier to implement.








     Standards with release  limits for each separate human  intrusion do




not, however, place a direct  limit on the total estimated harm  to future




generations.  If a disposal  system were  located so  that human  intrusion




was quite likely, the risks  from disposal of the  wastes from 100,000 MTHM




could be much larger than  the  limits proposed  in  our  standards.   (For




example, our risk assessment  assumes about 200  intrusions  into  the salt




repository over 10,000 years.  Release  limits  corresponding to  50 to 100




excess deaths per intrusion  would then be associated with  10,000 to  20,000




excess deaths over 10,000  years).  Implementation of  this  alternative




would rely on qualitative  judgments  by  the  regulatory  agencies;




quantitative estimates of  overall risk would  not  be required.
                                    85

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     There are definite advantages and disadvantages to either of these




two possible choices, and they represent two viable options.  We decided




to select the single set of containment requirements for our proposed




standards.  We believe the simplicity of the single set of values, which




adequately protect the public health, outweigh the possible advantages of




the dual requirements.
                                  86

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                                Chapter 5




                  PROJECTED HEALTH EFFECTS  FROM DISPOSAL








     This chapter is a synopsis of our assessment of the  risks from




disposal of high-level radioactive wastes  in mined geologic repositories.




We describe the model geologic repository  and show how the released




radionuclides reach the accessible environment and people.  We also




describe how we assess the risks to future generations.  The results of




the risk assessment are given in terms of the number of fatal cancers and




of the number of genetic effects to be e'xpected in the exposed population




over 10,000 years or more.  A more detailed description of the models




used, with additional references,  is provided  in Appendix B.









     Our assessment consisted of a number of  steps.  First, we defined  the




reference mined geologic repositories.  We  then analyzed various  events or




processes  that could  lead  to  releases  of radionuclides from a repository




and modeled  the movement of those  radionuclides through  the geosphere  to




the accessible environment.   From the  probability of release  and  the




associated consequence, we then calculated  the resulting risk to  the model




population (assumed rather similar in  demography to the  present United




States  population)  and summed the risks from all these releases over the




 first  10,000 years  after  repository sealing.









      The results of the assessment depend on our assumptions about the




performance  of the  repository and about  the movement of radionuclides




 through the  environment.   To  describe the behavior of each part of the






                                    87

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 disposal  system, we  chose  performance  values  that  we  believe would be




 achievable  by  a  carefully  selected  site  and a well-designed repository.




 The  numerical  values that  describe  each  physical,  chemical, or  biological




 process are  given  in Smith,  et  al.  (SMC  32).   The  assessment then shows




 how  the results  depend  on  the various  characteristics of  the disposal




 system and which characteristics are most  important.








     We evaluated  the risks  from model repositories in  five different




 geologic media:  bedded  salt, dome salt,  granite, basalt,  and shale.  Our




 detailed discussion  here concentrates  on repositories in  salt and granite,




 since these were found  to  give  lower risks.   For comparison, some results




 for  basalt and shale repositories are  also given.









 5.1  MODEL HEPOSITORIES




     Our model repositories contain radioactive waste consisting of




 100,000 metric tons of unreprocessed spent fuel, aged 10 years after




 removal from reactors.  We chose spent fuel for our assessment because




 this requires examination of the behavior of  both  fission products and




 transuranium nuclides.  The waste is assumed  to be contained in canisters




 that would last 100 years in salt and 500 years in granite.









     Figure 5-1 shows a model of the repository in bedded salt; Figure 5-2




shows a model of the repository in granite.   Both  reference repositories




cover an area of 8 square kilometers,  about one-fourth of which is
                                   88

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                                                         Surface
       Surface
       Deposits


"_T_- Shale
      Salt
.Repository
:n_" Shale
       Aquifer
410 Meters

460 Meters

510 Meters

560 Meters
590 Meters
FIGURE 5-1    REFERENCE  REPOSITORY  IN BEDDED  SALT
                        89

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                                                     Surface
                                                     200 Meters
                                                     230 Meters
                                                     460 Meters
FIGURE  5-2   REFERENCE REPOSITORY  IN GRANITE

-------
actually mined.  The reference repository in bedded salt is 460 meters




below the surface.   The salt layer is 100 meters thick,  so there are




50 meters of impervious salt above and below the repository level.   There




are an additional SO meters of relatively impermeable shale immediately




above and below the salt; then there are overlying and underlying




water-bearing permeable layers 30 meters thick.  For convenience, we refer




to these permeable layers as aquifers, although they may not be useable




sources of groundwater.  Above the upper aquifer is another 330 meters of




overburden.  The groundwater discharges into a stream 1600 meters from the




repository site.  The reference granite repository is also 460 meters




below the surface.  The granite formation continues indefinitely below




it.  Above it  there is a 230-meter thickness of granite, and above  this  is




a permeable stratum identical with the one modeled for  the bedded-salt




repository.  Above this permeable stratum are  200 meters of overburden.








     We assume that the material  removed to construct the rained cavity




will be returned as backfill  after the wastes  have been emplaced.   This




backfilled material cannot  be compressed to  its original  density;  we




assume  there will  be void  spaces  amounting  to  20  percent  of the mined




volume.   Because the repositories are below the water table and cannot be




completely  sealed  from groundwater,  we expect  that the  void spaces will




fill with water.   In granite,  this water enters from the  upper aquifer




through  fractures  in the rock and eventually fills all  the void spaces in




the  backfill.   In  salt,  it enters from  the  upper aquifer  through
                                    91

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imperfections in shaft and borehole seals.  Much of the void space is

eliminated by plastic movement of the salt, and only a small amount of

water is trapped in the salt in the form of brine pockets.



     As the canisters fail, the waste is exposed to water in the

repository and is gradually leached from the matrix and dissolved.  This

contaminated repository water is the largest source from which

radioactivity reaches the accessible environment.  We selected a release
    *                             -4
rate  of 0.01 percent per year (10   per year) for the waste.  We

assumed that the uptake of nuclides of uranium, plutonium, neptunium, and

technetium by groundwater was limited by their solubility, which was taken

to be 1 milligram per cubic meter of water.  We also assumed that the

solubility of americium was 50 grams per cubic meter of water and that of

tin was 1 gram per cubic meter.  Nuclides of these elements represent the

most important long-term concern in terms of movement through the

environment and possible hazard to people.
     *"Release rate" is defined as the fraction of material
dissolved from the high-level waste matrix by the surrounding liquid
(water) during a given period of time.  This rate generally determines
the maximum rate at which any nuclide can enter the repository water.


                                   92

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     We modeled the aquifer as a porous permeable rock.  Water flows in


the aquifer slowly, the speed being determined by the porosity  and

            **
permeability   of the rock and by the hydraulic pressure gradient in the


aquifer.  We assumed that the aquifer permeability is 10   centimeters


per second (about 30 meters per year), the aquifer pressure gradient is


0.01, and the effective porosity is 0.15.  This combination of


permeability, flow gradient, and porosity corresponds to an interstitial

              ***
water velocity    of about 2 meters per year.  Nuclides reaching the


aquifer will move in the direction of groundwater flow.  A few nuclides


will move at the same speed as the groundwater, but most will move more


slowly.  The extent to which a nuclide is slowed in comparison to the

                                                 Tlfllf "Jfllf
groundwater is expressed as a retardation factor.      We selected


conservative retardation factors for the significant nuclides.  These


factors were 100 for actinides, 10 for tin, and 1 for other nuclides.


The choice of these relatively small retardation factors means that the


calculated values for movement along the aquifer flow are'likely to be


faster than will actually be observed.
     *"Porosity" is the fraction of the rock  formation available  to
water (void fraction).


     **"Permeability" is the ability of a geologic  formation  to
transmit water.


     ***"Interstitial water velocity" is the  speed  of water flow
through the rock or other medium.


         "Retardation factor" is the ratio of the interstitial water
velocity to the velocity of the radionuclide.  For  example, a
retardation factor of 100 means that the nuclide moves at  1/100 the
speed of the water.  Retardation is caused by adsorption or other
mechanisms which slow the movement of chemical compounds through  the
environment.
                                    93

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 5.2  RELEASE  AND TRANSPORT MECHANISMS




      Under an EPA contract,  Arthur D. Little,  Inc.  (ADL)  identified  and




 analyzed possible processes  by which the waste could  be  released  from a




 repository (ADL  79).   Once released, radionuclides  could  enter  the




 environment via  air,  land  surface,  or groundwater pathways.  The  ADL study




 provides the  basic  information needed to describe the quantities  of




 radionuclides  released from  the repository  by  each  process or event,  and




 the probabilities of  release.   The  report includes  estimates of the




 fraction of the  repository affected by an event and of the groundwater




 flow  rate through the  affected area of the  repository.









      Radionuclides released directly to  air or land surface reach the




accessible  environment quickly.  Releases into groundwater can reach  the




accessible  environment only after considerable delay.  We applied a  simple




one-dimensional  transport model of  radionuclides in the groundwater.  The




model neglects dispersion, but  includes  the velocity  of groundwater  flow,




retardation of nuclides, and solubility  limits for  certain nuclides.









     Once the radionuclides enter the accessible environment, they can




reach people through a number of pathways (SMJ 82).   We considered a total




of 30 of these pathways.  For radionuclides reaching  surface water,  for




example, we considered: (a) drinking water;  (b) eating fish;  (c)  eating




three types of foods grown by irrigation—surface crops,  milk and beef;




(d) breathing air contaminated by resuspension after  irrigation; and
                                   94

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 (e)  standing on ground contaminated by irrigation.  For each nuclide, the


 fraction eaten or breathed by humans from each pathway was calculated.


 Doses  to those organs believed susceptible to radiation induced cancer


 were calculated, as well as doses to the reproductive organs.  The doses


 to  each organ were then summed over all pathways, including the external


 dose pathways.  We then estimated the total numbers of fatal cancers and


 first  generation genetic defects to the population.  Risk conversion


 factors were derived from the report of the Committee on the Biological


 Effects of Ionizing Radiation (BEIR) of the National Academy of Sciences

         &
 (NAS 72).   Table 5-1 summarizes our estimates of the fatal cancers


 caused by each curie of radioactivity released to the environment for each


 of  the three release modes: to surface water (via groundwater), to the


 land surface, or to the air (SMJ 82).





     We expect population distributions, food chains, and technology to


 change dramatically over 10,000 years.  Unlike geologic processes, they


 can  be realistically predicted only for short times.  Accordingly, our


 estimates of health effects used general models of environmental pathways,


 populations, and living habits (SMJ 82).  Rather than attempt  to predict


 future changes in populations, cancer cure rates, etc., they assume


 present values for these parameters.
      We have not recalculated these risk estimates on the basis of
the July 1980 report of the BEIR committee (NAS 80).  The risk estimates
using the 1980 report would be somewhat lower, but not by as much as a
factor of 10.
                                   95

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                   Table  5-1




Potential Health Effects  (Fatal Cancers) Caused
per Curie Released to the Environment by Different Modes
Nuclide
Am-241
Am- 24i3
C- 14
Cs-135
Cs-137
1-129
Np-237
Pu-238
Pu-239
Pu-240
Pu-242
Ra-226
Sr- 90
Tc- 99
Sn-126
Releases to
Surface Water
0.73
2.77
0.05
0.004
0.02
0.01
0.6
0.02
0.07
0.07
0.07
3.17
0.12
0.0003
0.12
Releases to
Land Surface
0.09
1.03
0.00003
0.0004
0.0006
0.00002
0.003
0.003
0.06
0.05
0.06
0.08
0.001
0.00000006
0.04
Releases to
the Air
0.16
1.14
0.0002
0.0007
0.007
0.001
0.08
0.02
0.05
0.05
0.05
0.49
0.02
0.00004
0.11
                     96

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     The risk from an event (e.g., movement of a fault or inadvertent




human intrusion by exploratory drilling) is the probability of that event




multiplied by its consequences.  For each type of possible event, we




estimated the probability of its occurence within several different time




periods after disposal.  We did this because the same event or process




will release different amounts of radionuclides—because of radioactive




decay—depending on the time when it occurs.  We then estimated the




consequences for each event occurring within each time period after




repository sealing.  We then added the risks from all individual events to




obtain the total projected risk to the population.









5.3  RESULTS OF RISK ASSESSMENTS




5.3.1  Population Risks




     For the reference bedded-salt repository,  the projected  number  of




fatal cancers in the first 10,000 years  is  about  100.  Almost all  of these




are due to releases of contaminated water  from the repository to the land




surface as a result of drilling during  resource exploration.  For  the




reference granite repository,  the projected number of  fatal  cancers  in  the




first 10,000 years is about 700.  Almost all of these  result from  transfer




of contaminated granite repository water to the surface  during  drilling in




the same way as for bedded salt.  About  10 cancers are due to routine




releases from natural processes,  in which  the  heat generated by the  waste




is the driving force for a buoyancy pump,  which raises repository  water




through the fractures  in the  granite  to the permeable  stratum,  through
                                    97

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which it is carried to surface water.  Figure 5-3 shows the projected




health effects over 10,000 years for each type of event that may affect




granite or bedded salt repositories.  Figure 5-4 provides a comparison of




projected health effects for repositories in different geological media.









     We also evaluated the health risks over longer periods, considering




only the events and processes occurring in the first 10,000 years.  In the




granite repository, many nuclides remain in transit in the groundwater




during the first 10,000 years, so that the expected number of fatal




cancers increases to about 900 in 50,000 years and to 2,800 in 100,000




years.   The corresponding number of fatal cancers from a bedded-salt




repository increases only to 150 in the first 20,000 years and remains




constant at this figure for 100,000 years.









     To evaluate the effect of changing our assumptions, we also




calculated results for different choices of the more important variables.




Resistance of the waste to leaching and the insolubility of uranium,




neptunium, plutonium,  and technetium are important in limiting the risks




from disposal.  Both the nature of the waste form and the geochemical




characteristics of the site determine leaching resistance and




insolubility.  Leaching and solubility change the availability of the




radionuclides to the groundwater in the repository.  Reduction of the




leach rate by a factor of 100 per year from the reference case reduces the




projected fatal cancers in granite to about 10; increasing the leach rate




by a factor of 100 increases them to 2,500.  Effects for the bedded-salt
                                   98

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10 5-
10 4 -
10 3-
a io2-
I lo1'
* 10°-

-------
   5000-
    7000-
    6000-
FIGURE 5-4:

PROJECTED HEALTH EFFECTS

OVER 10,000 YEARS FOR

REFERENCE REPOSITORIES

IN DIFFERENT

GEOLOGIC MEDIA
to
0)
o
o   5000.
s.
O)
o
 o  4000-
<*-
 
-------
repository are much smaller.   The effect of  varying the  leach  rates  is




shown in Figure 5-5.   The effects of assuming infinite solubility  are




shown in Figure 5-6.   Eliminating the solubility limit for selected




actinides and for technetium would increase  the projected number of




fatal cancers in the granite repository to almost 2,000, and in the




bedded salt to 9,400.








     The stability and durability of canisters are less important than




resistance of the waste form.  Modeling canister lifetime at 1,000




years produces relatively small decreases in the projected number of




fatal cancers; modeling canister lifetime at 5,000 years  reduces the




number of projected fatal cancers by a  factor of about  five.




Long-lived canisters, on the order of 50,000 to 100,000 years, wbuld




substantially reduce the number of fatal cancers because  of decay of




important nuclides, especially americium-243, while the containment




existed.  The effect of different assumptions on canister lifetime  is




shown in Figure 5-7.








     Modeling all  nuclides with  retardation factors equal to  one, so




that all nuclides  move with  the  speed of groundwater,  increases the




number of projected  fatal cancers from  the  granite repository  to  over



30,000.  Figure 5-6  shows the effects of assuming  no  retardation.




Higher retardation factors,  such as  10,000  for  the actinides  (except




neptunium), do not significantly reduce the number of projected fatal




cancers, since these radionuclides move so  slowly  through the




environment.







                                   101

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    8000-
S-
to
OJ
o
o
o
OJ
o

in
 o
 OJ
 
-------
16,000-
14,000-


12,000-
j_

-------
    8000-
    7000-
    6000-
 S-
 «3
 CD
O
    5000-
 S-
 CU

 o

 to

 «  4000


 
•a
O)
+j
o
01
•"-J
o

Q-
    3000-
    2000-
    1000-
FIGURE 5-7:


PROJECTED HEALTH EFFECTS


OVER 10,000 YEARS


VS.


DIFFERENT CANISTER LIFETIMES


(years)
                    1000
               2000        3000

          canister lifetime (years)
4000
5000
                                   104

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     Most of the fatal cancers for the reference case are associated with




intrusion into the repository by drilling;  therefore, the number of fatal




cancers is roughly proportional to the frequency of drilling.   The




frequencies suggested for such drilling events by ADL are one in 50 years




for the bedded-salt repository and one in 400 years for the granite




repository.









     For the reference cases, the calculated first generation genetic




effects are lower than the number of fatal cancers by a factor of more




than 10.  This is because most of the important nuclides involved are bone




seekers and none concentrate in gonads or soft tissue.








5.3.2  Risks to Individuals




     Without effective institutional controls, the probability  that




someone exploring for resources will drill into the  repository  at  least




once in the first 5,000 years is high, even at drilling  frequencies




significantly smaller than those estimated by ADL.   The groundwater




immediately above the repository would be contaminated by  repeated




drilling events, and some of it would remain contaminated  for  long  periods




of time.









     Individuals may be exposed to nuclides released from  the  repository




by breathing air contaminated as a result of releases  of nuclides  to land




surfaces or by drinking water contaminated as a  result of  introduction of




radionuclides into the groundwater (GO 82).  Doses  from  breathing  air were
                                   105

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 calculated as relatively  small at distances of  100 meters or more from the




 point of discharge to  the  surface.  Radionuclides move slowly in




 groundwater and exist  in  rather high concentrations over fairly small




 areas.  The most  important nuclides in groundwater are the americium




 isotopes 241 and  243.  These nuclides, particularly the long-lived




 americium-243, persist for a long time in the environment.








     We analyzed  a scenario in which an initial drilling event directly




 contacted contaminated groundwater in a granite repository.  This scenario




 results in the greatest contamination of the groundwater in the adjacent




 aquifer (GO 82).  For each such hypothetical drilling event, contamination




would spread slowly in the aquifer.  By 1,000 years after the drilling, an




 aquifer area of about 4,000 square meters would be contaminated enough to




give a dose of more than 500 millirems per year to people who might




 subsequently drill into the aquifer and drink the water.  By 10,000 years




after the initial drilling, an aquifer area of about 80,000 square meters




would be contaminated to this level.









     The probability that an individual would subsequently drill a water




well into an area contaminated by resource exploration depends on




conditions at the site, particularly the relative availability of




contaminated and uncontaminated water.  People are not likely to drill




wells into a deep aquifer if adequate water is otherwise available.  Using




the following assumptions, we made a generic estimate of the probability




that a person might encounter a contaminated area:  about 500,000 water
                                   106

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weLLs are drilled per year (GE 73), of which some 2 percent are deeper




than 150 meters (500 feet).  The U.S. Geological Survey (NW 79) estimates




that about one-third of one percent of all wells are low-yield (less than




10 gallons per minute), deep (greater than 500 feet) wells.  To allow for




increased use of ground water in  the future, we assume that 5,000 deep,



low-yield wells are drilled in the United States each year.  The number




of new wells, divided  by  the U.S. land area of 7,800,000 km ,  is




6.4x10    new wells per square meter per year.  The probability of a




person drilling into an 80,000 square meter area (the area contaminated




above 500 millirem/year at 10,000 years) would be about one in two-hundred




during any  100-year period.








5.4  CONCLUSIONS




     Our assessment of disposal  in mined geologic repositories identifies




the major  factors  in estimating  risks to future  generations.   The




following  conclusions  can be drawn  from the  assessment:








     The release of a  large  fraction of the  repository inventory should be




extremely  unlikely.   In a deep-mined geologic repository,  only a volcanic




eruption or the  impact of a  large meteorite  should produce a very large




release of  radionuclides  (more  than  a few  percent  of the inventory).  The




chance of  such releases occurring within 10,000  years  is vanishingly small




if the wastes are placed  deep enough below land or water in a  nonvolcanic




area.
                                   107

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      The movement  of  important  radionuclides  in  groundwater  from  the




 repository  to  the  accessible  environment  should  be  slow  enough  to provide




 substantial time for  decay of radioactivity.  Low permeability  and a  small




 hydraulic gradient in the groundwater  stratum will  result  in slow




 groundwater movement.  High adsorption is necessary  to further  slow the




 movement of such moderately long-lived radionuclides as  plutonium-239,




 americium-241, and americium-243.









     The entry of  radionuclides into groundwater should  be slow.  Some




 radionuclides, particularly those of uranium, neptunium, plutonium, and




 technetium,  can be  kept essentially insoluble in groundwater by proper




 selection of their  chemical form in the waste.  We can select repository




 sites where  the geochemistry  will keep these elements in their  insoluble




 forms.  For other elements, selection  of waste forms that are leached




 slowly should limit the rate  of entry  into groundwater even  though the




 chemical forms of these elements in the waste are more soluble.









     The chance that any individual will receive high doses  should be




 small.  High doses  are associated with use of contaminated groundwater or




 air close to a point where radionuclides have been released.  The




 probability of an  individual  receiving high doses can be made small by




 selecting sites where groundwater is not likely to be used,  where there




are no resources likely to be exploited, and where natural intrusive




events are  improbable.
                                   108

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     In general, we have shown that it is reasonable to expect disposal of




high-level radioactive wastes in a mined geologic repository (containing




the wastes from 100,000 MTHM) to result in fewer than an average of one




fatal cancer every 10 years for the first 10,000 years, provided the




repository is reasonably well designed and well sited.  Few of these




health effects are expected to occur  in the first 1,000 years after




emplacement.  The most important cause of radionuclide release appears to




be human  intrusion in the course of exploration for resources.  Health




effects may be reduced by using waste forms resistant to leaching  and




dissolving.  More importantly, adverse effects will be reduced by




selection of sites where: (1) the geochemistry will maintain technetium




and most actinides in an insoluble form, (2) the hydrology will retain




many nuclides in the groundwater for  long times, and (3) exploration for




resources is improbable.
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110

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                                Chapter 6




                  PROPOSED STANDARDS FOR WASTE DISPOSAL









     The proposed standards  for disposal  contain three  sections:




containment requirements,  assurance requirements, and procedural




requirements.   The containment requirements set limits  on potential




releases of radioactive materials from disposal systems to the accessible




environment for 10,000 years after disposal.  These requirements are based




on the assumption that our predictions of projected risks from disposal




systems are good enough to use in selecting, designing, and implementing




disposal methods.  These predictions can then be used to assess the




long-term releases from a disposal system and to decide whether it will




provide adequate protection to present and  future generations.









     Closely coupled with the numerical containment  requirements are seven




qualitative assurance requirements that are needed to have confidence  that




the  long-term release limits will be met.   The assurance requirements




address and compensate for the uncertainties that necessarily accompany




plans to isolate high-level and transuranic wastes from  the environment




for  a very long time.  No matter how promising analytical projections  of




disposal system performance appear to  be,  these  wastes should be disposed




of in a cautious manner that  reduces the  likelihood  of unanticipated




releases.  Our assurance requirements  provide  the  context necessary for




application of our containment  requirements, and they  should  ensure very




good long-term protection of  the  environment.
                                   Ill

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     Finally, our procedural requirements provide  instructions that must




be followed to ensure that the containment requirements are properly




implemented for particular disposal  systems.  These requirements describe




the assumptions that are appropriate when assessing the performance of




disposal systems to determine compliance with our  long-term release limits.








     In the following sections, we discuss the various features of all




three sets of requirements, explain why we selected them, and describe how




they should be used.









6.1.  CONTAINMENT REQUIREMENTS




     The primary requirements of our proposed environmental standards are




numerical limits on the amount of radioactive waste that may be released




from a disposal system to the accessible environment for 10,000 years




after disposal.  These containment requirements should provide excellent




protection of public health and the environment.  In fact, they should




limit the risks to present and future generations to a level no greater




than the risks from equivalent amounts of unmined uranium ore.









     The following sections indicate how we selected the form of these




containment requirements, why we chose the specific level of risk they




are based on, and how the release limits are to be used.  Because they




address such a long time period, and because they include unplanned




releases,  these containment requirements can be implemented only through




analytical projections of disposal system performance.  There will be many
                                   112

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uncertainties in making such long-term performance projections.

Accordingly, our proposed standards require only a "reasonable expectation"

that these containment requirements will be met; this determination must

be made by the implementing agencies.



6.1.1  The Accessible Environment

     The proposed containment requirements apply to releases to the

"accessible environment," which we identify by a set of definitions:


          "'Accessible environment1 includes (1) the atmosphere,
     (2) land surfaces, (3) surface waters, (4) oceans, and
     (5) parts of the Lithosphere that are more than ten kilometers
     in any direction from the original location of any of the
     radioactive wastes in a disposal system."

          '"Lithosphere1 means the solid part of the Earth,
     including any groundwater contained within it."

          "'Groundwater1 means water below the  land surface in a
     zone of saturation."



     Through these definitions we intend to protect all portions  of  the

environment that may be in direct contact with  man.  (For  example,  the

atmosphere, land surface, and any surface water directly over  a geologic

repository  is included within the "accessible environment".)   In  addition,

our definition of the "accessible environment"  includes all groundwater

formations  that are more than 10 kilometers away  from  a disposal  system.

Groundwater is an important resource; currently,  more  than 20  percent of

the fresh water used  in the United States  comes from underground  sources.

However, geologic formations that contain  relatively small amounts of
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groundwater can serve as an  important part of the containment system for a




geologic repository.  Thus, we need to allow for contamination of some




groundwater while we are ensuring the long-term isolation of these wastes.








     Accordingly, our approach does not provide any direct protection




for the relatively small amount of groundwater that could be within




10 kilometers of a geologic repository.  However, since the amount of




groundwater left unprotected should be kept as small as possible,




consistent with other requirements, we expect that the Federal




environmental impact statement for each disposal system will identify all




sources of groundwater within 10 kilometers of the disposal system, will




describe the potential long-term environmental effects of possible




contamination of these sources of groundwater, and will consider these




effects as one of the factors in evaluating alternative sites.









     On the other hand, our definition of "accessible environment"




includes all groundwater sources—regardless of their quality—that are




more than 10 kilometers away from a disposal system because our analyses




indicate that this should be an ample distance to provide the protection




expected from the geologic barriers of a repository.  (In fact, all of the




analyses described in Chapter 5 assume that the accessible environment,




with respect to groundwater, is only 1 mile away from the repository—the




longer distance of 10 kilometers would tend to reduce the residual risks




identified in Chapter 5.)
                                   114

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6.1.2  Radionuclide Release Limits




     The proposed containment requirements set limits on the projected




releases of radionuclides to the accessible environment for 10,000 years




after disposal.   Because of the uncertainties of such long-term




projections, we require a disposal system design to provide a "reasonable



expectation" that the release limits will be met.  A comprehensive




performance assessment would be the basis for making this judgment.








     These requirements would limit "reasonably foreseeable" releases to




the quantities derived from Table 2 of the proposed standards: "Release




Limits for Containment Requirements" (see Appendix A of this document).




"Very unlikely" releases would be limited to 10 times the quantities for




"reasonably foreseeable" releases.  "Reasonably foreseeable" releases are



those estimated as having a greater than  1 in 100 chance of occurring




within 10,000 years.  "Very unlikely" releases are those with a chance  of




occurring in 10,000 years estimated to be less than  1  in 100 but  more  than




1 in 10,000.  We selected the qualitative terms "reasonably foreseeable"




and "very unlikely" to emphasize  that these are not  precise estimates.




The implementing agencies may use some discretion  in assessing  the




probabilities of releases.








     The values  in Table 2 of the proposed standards are the  quantities of




each radionuclide that would result in 10 projected  fatal  cancers over




10,000 years (SMJ 82) if released into surface water—the most  dangerous




of the three release modes shown  in Table 5-1.  The  values  in Table  2  are
                                   115

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 expressed in terms of releases from wastes generated from 1000 metric tons




 of heavy  metal (MTHM) originally placed in the reactor;  they correspond to




 1,000 projected fatal cancers in 10,000 years from the 100,000 MTHM




 geologic  repository used  in our risk assessment.   We estimated that the




 long-term risks from disposal of wastes from 100,000 MTHM in our model




 repositories for salt and granite would be about  100 to  800  fatal cancers




 in 10,000 years.   We choose release limits for Table 2 of the proposed




 standards that correspond to somewhat higher long-term risks.









      We selected the value of 1 million curies of transuranic elements as




 a  reference  value for transuranic wastes to require about the same degree




 of control for the long-lived alpha-emitting radionuclides in transuranic




 wastes that  we are requiring for those radionuclides in  high-level




 wastes.   Since a  100,000-MTHM repository contains about  300  million curies




 of transuranic  nuclides,  the 1,000,000-curie reference quantity for




 transuranic  wastes is in  about  the same ratio to  this total  as the




 1,000-MTHM reference  quantity of spent fuel is with respect  to the




 100,000-MTHM model  repository.   Or,  in other words,  the  reference values




 were  selected  so  that about  the  same  fraction of  transuranic radionuclides




 would be  retained  for either high-level or transuranic wastes.









      Since more  than  one  nuclide would be  released  in any actual




 situation, the  limitation  for each nuclide must be  lowered (made more




 stringent) to  allow  for the  risks  attributable to other  nuclides.   This is




done through summing  the  terms:




              a/A + b/B +  c/C +  ...






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where a, b, c... are the cumulative releases of the various nuclides and




A, B, C... are the appropriate Table 2 values.  The total releases are




summed over all radionuclides as well as over all release pathways.








     The containment requirements provide for less stringent limitations




on "very unlikely" releases of radionuclides to the accessible




environment.  They would set release limits that are higher by a factor of




10, with the requirement that the probability of such a release must be




lower by a factor of 100.  The requirement for a greater decrease  in




probability than the increase in released quantity reflects our aversion




to large consequences.  This provision amounts to about a 10-percent



increase in effects on health.








     These containment requirements do not permit large projected




cumulative releases, which would have severe consequences, unless  their



probabilities are less than 1 in 10,000 over 10,000 years.  Such large




releases could result only from extremely disruptive events, such  as a




volcanic eruption or the impact of a large meteorite, occurring when the




repository still contained a significant inventory of undecayed waste




radionuclides.  We do not believe that releases that are so highly




improbable should be regulated.
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6.1.3  The Level of Protection




     We used the information in Chapters 3 and 5 to select a limit on harm




to future generations that we believe is very small and that we believe




good geologic repositories can meet.  The limit on which we based our




proposed containment requirements is 1,000 excess cancer deaths over




10,000 years for a 100,000-MTHM repository.









     As shown in Figure 6-1 (which reproduces Figures 5-4 through 5-7 from




Chapter 5, with the proposed level of protection indicated), there are




several combinations of geologic media,  engineered barriers, and site




geochemistry that should provide at least this much protection.  We




believe, with careful site selection and with the technology available to




develop engineered barriers, mined geologic repositories can keep risks




below this level.
                                   118

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8000-

7000-

6000-
5000 •
4000-
3000-
2000-

1000 •
n _
PROJECTED HEALTH EFFECTS
OVER 10,000 YEARS FOR 	
REFERENCE REPOSITORIES
IN DIFFERENT
GEOLOGIC MEDIA
















PROPOSED STANDARDS
i 	 1 I 	 1






















°ciET B
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6.2  ASSURANCE REQUIREMENTS

     Our seven qualitative assurance requirements address and compensate

for the uncertainties  that necessarily accompany plans  to isolate

high-level and transuranic wastes  from the environment  for a very  long

time.  They call for a cautious and "common-sense" approach to disposal

ttiat provides the necessary context for application of  our containment

requirements.  We expect that the  specific steps taken  by the implementing

agencies to comply with each of the following criteria  will be described

in the Federal environmental impact statement—and other decision

documents—for each disposal system:



6.2.1  Criterion 1:
         "Wastes shall be disposed of promptly once disposal systems
     are available and the wastes have been suitably conditioned  for
     disposal."
     Waste storage systems require active human controls  for periodic

inspection and maintenance.  To reduce dependence on active human

controls, this assurance requirement specifies that the wastes  should be

removed from storage and disposed of promptly once adequate disposal

systems are available and the wastes have been appropriately processed  for

disposal (i.e., placed into a suitably stable chemical and physical form).



     We have not established a specific time period in this requirement,

because the appropriate length of storage may depend on details of the

disposal system design.  For example, it may be desirable to store
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high-level wastes for several decades to allow for decay of most of the

short-lived radionuclides.  The primary intent of this criterion is to

prevent wastes from being stored indefinitely to avoid ultimate disposal.



6.2.2  Criterion 2;


         "Disposal systems shall be selected and designed to keep
     releases to the accessible environment as small as reasonably
     achievable, taking into account technical, social, and economic
     considerat ions."


     As discussed in Chapter 2, we assume that any exposure to  radiation

can produce some harm.  Accordingly, this assurance requirement stipulates

that releases of radioactive wastes be kept as small as reasonably

achievable.  It instructs the regulatory agencies to reduce releases  below

the limits of the containment requirements on the basis of site- or

method-specific information.



     This principle has been used in radiation protection  for  a long

time.  It is not a totally quantitative concept,  and we have not placed

numerical specifications  on  it here.  The technical judgment of the

implementing agencies would determine when  this  criterion  has  been

satisfied.
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6.2.3  Criterion 3:
         "Disposal systems shall use several different types of
     barriers to  isolate the wastes from the accessible
     environment.  Both engineered and natural  barriers shall be
     included.  Each such barrier shall separately be designed to
     provide substantial isolation."
     This assurance requirement emphasizes the need for caution in

designing a disposal system by requiring multiple barriers.  We will

always be uncertain about how well a barrier may perform over the long

lifetimes required of disposal systems.  Furthermore, if a barrier  in a

disposal system fails, the failure would probably not be detected for a

long time, if at all.



     Therefore, the barriers should retain the radioactivity by different

chemical and physical mechanisms.  We believe that both engineered  and

natural barriers should be included.  Each of the barriers should be

designed separately to provide substantial isolation, even if the other

barriers perform poorly.  Extra barriers should be added when they  can be

provided at reasonable cost,  even if they do not appear to be needed to

meet the performance requirements for the overall disposal system.
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6.2.4  Criterion 4;


         "Disposal systems shall not rely upon active institutional
     controls to isolate the wastes beyond a reasonable period of
     time (e.g., a few hundred years) after disposal of the wastes."


     This assurance requirement would limit reliance on "active"

institutional controls to a reasonable period after disposal,  which should

be no more than a few hundred years.  "Active" institutional controls

include guarding a disposal site, performing maintenance operations or

remedial actions at a disposal site, or controlling or cleaning up

releases from a disposal site.



     This requirement does not mean we think society will lose all

knowledge of radioactivity, nuclear energy, radioactive wastes, or even

specific disposal sites after a few hundred years.  On the contrary, we

believe that such information is likely to survive, even without  the

extensive markers and records called for by the next assurance

requirement.  However, merely having this knowledge does not  guarantee

that it will be widely disseminated or effectively acted upon.  We believe

it is prudent to assume that  society may not  retain  active  controls over

disposal systems for very  long, and  that unrelated  activities may resume

at a disposal site even though  the  presence of radioactive  wastes is

documented.  The assumptions  that we believe  are  appropriate  when

considering  the effectiveness of "passive"  institutional controls are

described under our procedural  requirements.
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6.2.5  Criterion 5:
         "Disposal  systems  shall  be  identified  by  the most  permanent
     markers and records practicable  to  indicate the dangers  of  the
     wastes and their  location."
     During our public workshops on radioactive waste  criteria, we

explored whether the location of radioactive wastes  should  be  concealed

from future generations or should be clearly marked  (EPA  77).  We became

convinced that concealment was impractical because the inevitable

anomalies of a disposal site would eventually attract  attention.

Therefore, trying to limit future knowledge would probably  only make  the

situation more dangerous.



     Accordingly, this assurance requirement calls for comprehensive

actions to be taken to pass on information about the wastes and the way we

disposed of them.  However, we do not assume that this  information can be

guaranteed to prevent disruption of the site if active  controls are lost.

Unfortunately, warning signs of many kinds are often ignored today; we

cannot assume they would be much more effective in the  future.
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6.2.6  Criterion 6;


         "Disposal systems shall not be located where there has been
     mining for resources or where there is a reasonable expectation
     of exploration for scarce or easily accessible resources in the
     future.  Furthermore, disposal systems shall not be located
     where there is a significant concentration of any material
     which is not widely available from other sources."


     This assurance requirement is a logical complement to our limitation

on active institutional controls.  Resources will attract exploration, and

the people doing the exploring may not know of the locations or dangers of

radioactive wastes.  Since active controls may not be available to  limit

access, disposal systems should not be located where people are likely to

explore for scarce or easily accessible resources.



     Obviously, exploration may occur anywhere, and careful judgment must

be used in applying this criterion.  Past and current mining is an

important indicator, but future trends must also be considered.  Some

resources that are not economical to recover today may  become a

substantial part of our resource base in the future; oil shales are an

example of such a resource.  Also, minerals not important  now might become

so in the future.



     Because of these considerations, we believe that  disposal  systems

should not be  located where  there  is a substantial concentration of any

material not widely available elsewhere.  Nor  should  they  be  located  where

exploration or mining is likely in the future, particularly  if  related to
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resources that can be neither renewed nor significantly recycled (such as




energy resources).  This assurance requirement could also be important




even if active institutional controls survive for a long time.  A poorly




located disposal system could deny or seriously hinder use of critical




resources to future generations.








     This assurance requirement should discourage the use of geologic




formations frequently associated with resources.  For example, salt domes




are often mined either for their relatively pure salt or for use as




effective, inexpensive storage caverns.  Of the 130 domes in the Gulf




Coast area judged suitable for such exploitation, 48 were in use as of




1965 (GR 81).  As discussed under Criterion 4, we do not believe that




active institutional controls should be relied upon to prevent exploration




for resources for more than a few hundred years after disposal.




Therefore, the frequent use of salt domes as resources would argue against




locating a repository in this type of structure.








     These particular concerns generally would not apply to bedded-salt




deposits, which are far more extensive.  A specific site in a bedded-salt




deposit appears much less vulnerable to inadvertent intrusion than a




particular site in a salt dome.
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6.2.7  Criterion 7:
         "Disposal systems shall be selected so that removal of most
     of the wastes is not precluded for a reasonable period of time
     after disposal."
     This assurance requirement would permit corrective action if

development of technical and scientific knowledge should indicate that a

disposal system poses much greater risks than had originally been

expected.  High-level radioactive waste disposal is a new technology.

Future generations may find it necessary or desirable to recover the

wastes and change the disposal site or methodology in light of new

knowledge.



     This requirement is different from the more familiar concept of

retrievable storage.  Recovering the wastes after the conventional

retrieval period of a few decades is likely to be expensive and, perhaps,

even dangerous.  It could require, for example, mining  the entire

repository volume to recover the wastes.  This criterion merely  requires

that this option be left open to our successors.



     Current plans for mined geologic disposal would comply with this

assurance requirement.  However, methods involving  irreversible

incorporation of radionuclides into disposal  sites  would not.  Deep-hole

placement, rock-melting, or hydrofracturing are  examples of disposal

methods  that do not appear to comply with  this assurance requirement.
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 6.3  PROCEDURAL REQUIREMENTS

     Compliance with our containment requirements will be determined

 through long-term projections of disposal  system performance.  The

 "performance assessments" used to make these judgments should be done  in

 accordance with the proposed definition included in our disposal standards:


         '"Performance assessment* means an analysis which
     identifies those events and processes which might affect the
     disposal system, examines their effects upon its barriers, and
     estimates the probabilities and consequences of the events.
     The analysis need not evaluate risks  from all identified
     events.  However, it should provide a reasonable expectation
     that the risks from events not evaluated are small in
     comparison to the risks which are estimated in the analysis."


     When our containment requirements are applied to a particular

disposal system, some of the procedures we used in our performance

assessments (described in Chapter 5) must be retained so that the intent

of our standards is met.  On the other hand, some of the assumptions we

made should be replaced with specific information appropriate for each

disposal system.  Our procedural requirements set forth three instructions

needed to ensure that performance assessments for particular disposal

systems are properly evaluated against the containment requirements.  Each

of these three procedural requirements is discussed here:
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         "The assessments shall consider realistic projections of
     the protection provided by all of the engineered and natural
     barriers of a disposal system."


     In developing the containment requirements, we considered the overall

protection that should be achievable by the combination of barriers in a

geologic repository.  Accordingly, the analyses used by NRG and DOE to

evaluate compliance with our requirements should consider realistic

assessments of the protection provided by all of the engineered and

natural barriers of a disposal system.  For example, performance

assessments of a geologic repository system should include the protection

afforded by geochemical retardation of radionuclides and by the  limited

solubility of radionuclides in groundwater, provided that reasonable

evidence is developed to support such mechanisms for that particular  site.


         "The assessments shall not assume that active  institutional
     controls can prevent or reduce releases  to the accessible
     environment beyond a reasonable period (e.g., a few hundred
     years) after disposal.  However, it should be assumed that  the
     Federal Government is committed to retaining passive
     institutional control of disposal sites  in perpetuity.   Such
     passive controls should be effective in  deterring  systematic or
     persistent exploitation of a disposal site, and it should  be
     assumed that they can keep the chance of inadvertent human
     intrusion very small as long as the Federal Government  retains
     such passive control of disposal sites."


     The assumptions we made in our performance assessments  about the

frequency of human intrusion were conservative because  they  ignored the

substantial protection that passive institutional controls should offer.

The performance assessments made for specific sites by  the implementing

agencies do not need to be as pessimistic with regard to human intrusion.

Because of the uncertainties of controls requiring the active participation
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 of  people  over  a  long  time,  performance  assessments  should not  assume  that

 active  institutional controls  can  prevent  or  reduce  releases  beyond a

 reasonable period  of time  (e.g., a few hundred years) after disposal.

 However, because  the Federal Government  is committed to  retaining control

 over  these disposal sites  in perpetuity, passive  institutional  controls

 should  substantially reduce  the chance of  inadvertent human intrusion  well

 beyond  this period.  These passive  controls should not be assumed to

 prevent all possibilities of inadvertent intrusion, because there is

 always  a chance that the controls will be  overlooked or  misunderstood.

 However, such measures should  be effective in deterring  systematic or

 persistent exploitation of a disposal site.  Furthermore, the chance of

 human intrusion should be very small as  long as the Federal Government

 retains passive control of disposal sites.


         "The assessments shall use information regarding the
      likelihood of human intrusion, and all other unplanned events
      that may cause releases to the accessible environment, as
     determined by the implementing agency  for each particular
     disposal site."


     We based our performance assessments  on relatively  simple  generic

models of geologic repositories and the data that was available for such

models.  Where information was uncertain, we made conservative  assumptions

 that should tend to overestimate the long-term risks of  disposal.

However, we do not intend that the  implementing agencies should use all of

the same models, data,  and assumptions that we did in making performance

assessments.  Instead,  the implementing agencies generally should use  the

best  information available for each particular disposal  site.
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                                Chapter 7




            PROPOSED STANDARDS FOR WASTE MANAGEMENT AND STORAGE








     Subpart A of the proposed standards establishes limits on the




annual dose equivalent to any member of the public as the result of the



management and storage of spent fuel and high-level and transuranic




wastes.  Management and storage operations for these materials include all




treatment and handling preparatory to disposal, including storage,




processing, packaging, and emplacement in the final disposal configuration.




Transportation is not included (see Section 2.6.1).








     Selection of a disposal technology defines the nature of the




operations necessary before disposal.  Several disposal technologies have




been proposed, such as geologic, extraterrestrial and transmutation.




These technologies have been reviewed for us by the MITRE Corporation




(AL 79a) and for the President by the Interagency Review Group (IRG 79).








7.1  WASTE MANAGEMENT OPERATIONS FOR VARIOUS DISPOSAL OPTIONS








7.1.1  Operations for Disposal of Unprocessed Spent Fuel




     Disposal of unprocessed spent fuel is applicable only to commercial



power generation; defense fuel is reprocessed for recovery of weapons




grade plutoniura and for its highly enriched uranium content.  Initial




storage of commercial spent fuel at the reactor site is required  to reduce




radioactivity and heat generation by decay before further handling is
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attempted.  Additional storage after packaging may or may not take place,




depending on the availability of a disposal site.  The management of




unprocessed spent fuel prior to terrestrial disposal includes:  storage




before packaging (if required), packaging, additional storage after




packaging, and emplacement.








     Storage Before Packaging




     Almost all spent fuel is now being stored, and considerable quantities




have been stored for a number of years.  This storage experience has




provided considerable information on environmental impacts.









     The Uranium Fuel Cycle standards (40 CFR 190) regulate storage of




spent fuel at nuclear power stations.  These proposed standards apply to




storage at independent spent fuel storage facilities (ISFSF), which are




not located at the reactor sites.  This is sometimes referred to as




away-from-reactor (AFR) storage.  The Department of Energy (DOE) estimates




the major releases from an ISFSF, in which 3000 metric tons of heavy metal




(MTHM) have been stored in water basins for 6.5 years, to be about




890 curies of krypton-85 and 2.4 curies of tritium, per year.  Almost all




of the krypton releases and about half of the tritium releases occur at




the time the spent fuel is received (HO 80a; DOE 79).  The projected




krypton-85 releases are small compared to the 50,000 curies per gigawatt




(electrical)-year (GWe-y) permitted under the standards of 40 CFR  190,




since 3000 MTHM corresponds to about 90 GWe-y, based on 33,000 megawatt




(thermal)-days per metric ton and an electrical efficiency of 32 percent.
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     We estimated that the annual dose to a maximally exposed individual




from the ISFSF (including dose associated with packaging) is




0.0006 millirem (mrem) to the whole body; the highest dose to any organ



of an individual is 0.015 mrem per year to the thyroid.  DOE estimates the




aggregate annual doses to the population within 80 kilometers of a




generic ISFSF, including the dose from a packaging operation, to be




0.47 organ-rems to the thyroid and 0.025 person-rems, whole body (DOE 79).








     Packaging




     Packaging spent fuel for final disposal is assumed  to be carried out




at the storage site rather than at the final disposal site.  We estimate




the major releases from packaging are 810 curies of krypton-85 and




1.3 curies of tritium per year (HO 80a; DOE 79).  The individual and




population doses are included in the ISFSF doses.








     Extended Storage




     We estimate that releases during recovery and handling of packaged



spent fuel stored for 30 years are about 0.01 curies of  krypton-85 and




about 0.1 microcuries of tritium per year (HO 80a; DOE 79).  Doses to any




organ of a maximally exposed individual are less than 1  microrem per year;



population doses are also negligible.
                                  133

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     Emplacement




     Releases of waste materials during emplacement occur only during the




handling of waste packages.  DOE estimates these releases to be about



  —20
10    curies per year (DOE 79) .  Resulting doses, both individual and




population, are below the 40 CFR 190 standards.
     Naturally occuring radionuclides are released from repository




materials during construction and operation.  However, the resulting doses




to individuals and populations are very small (DOE 79).  These releases




are not considered further, since they are small and are not significantly




affected by these standards.








7.1.2  Operations Prior to Disposal of Processed High-Level Wastes




     The management of processed high-level wastes prior to terrestrial




disposal includes five steps  (HO 80a):




         1.  High-level liquid waste (HLW) storage




         2.  Waste solidification




         3.  Interim storage  of solidified waste




         4.  Emplacement




         5.  Treatment and emplacement of retained krypton-85 and




             iodine-12 9









     High-Level Liquid Waste  Storage




     High-level liquid wastes result from defense activities and would




also result from the reprocessing of commercial spent fuel if reprocessing




were undertaken.  The technology for storing high-level liquid wastes is






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based on 30 years'  experience with defense wastes.   Wastes are stored in




subsurface tanks of stainless or mild steel. It is  expected that stainless




steel tanks will be required for the storage of high-level wastes from




reprocessing commercial spent fuel (DOE 79).  Double containment reduces




the possibility of leakage to soil.








     DOE has estimated the releases from a commercial reprocessing plant




with an annual capacity for treating 1140 cubic meters (300,000 gallons)




of high-level waste from 2000 metric tons of heavy metal.  This is the




spent fuel that corresponds to about 60 GWe-y.  About 27,000 curies of




tritium per year are estimated to be released to the atmosphere, while




releases of any other nuclide would be about 1 millicurie per year or




less.  DOE calculated the maximum annual dose equivalent  to any individual




to be about 0.1 millirem and the population dose within an 80-kilometer




radius to be about 11 person-rems.









     Waste Solidification




     In the event that commercial reprocessing of  spent  fuel  becomes  a




part of the national program, Nuclear Regulatory Commission  (NRG)




regulations (10 CFR 50, Appendix F) require that the high-level  liquid




wastes generated from this reprocessing be converted to  a solid  form




within 5 years and be transferred to the Federal Government within 10




years of its generation.  Some alternatives proposed for  long-term




management of defense wastes also include waste solidification (ERDA 77a;




ERDA 77b;  ERDA 77c).  Calcination is the most likely method of




solidification prior to incorporation into a solid matrix, while






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glassification provides a solid matrix for disposal.  The Idaho National




Engineering Laboratory has calcined high-level wastes since about 1963,




and Hanford has been developing various glassification techniques since




the early 1960's.








     We have assessed the environmental effects of calcination followed by




glassification (HO 80b).  The study considered fluidized-bed or spray




calcination followed by in-can or continuous-tnelter glassification.  Other




treatments are possible, including formation of crystalline (e.g.,




synthetic rock or metal) matrices.  Effluent releases from a generic plant




using available or near-term available effluent cleaning technology were




calculated.  The assessment covered the seven radionuclides (tritium,




iodine-129, ruthenium-106, cesium-134, cesium-137, strontium-90, and




plutonium-239) most likely to be released during normal operations of the




four potentially useful technologies available for calcination and




glassification.   We reviewed the control technology for particulates and




for off-gas cleanup, estimated decontamination factors for typical off-gas



control equipment for the radionuclides under consideration, and we




estimated the overall system decontamination factors for iodine,




ruthenium, and particulates to be 1x20 , 1x10 , and 1x10  ,




respectively.








     Radioactive effluent releases from the generic solidification plants




were calculated for two locations, a rural site typified by Barnwell,




South Carolina,  and an urban site typified by St. Louis.  Dose equivalents
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for whole-body and several organs were calculated for the maximally




exposed adult at each location.  Results are summarized in Tables 7-1




and 7-2.  The largest organ dose was determined to be to the lower large




intestine (LLI).  Ruthenium-106 is the major dose contributor for




materials up to one year old.  As ruthenium-106 decays, tritium become the




most significant nuclide in the older material.









     Population doses were calculated for the areas within an 80-kilometer




radius of the generic plant at each site.  Results are provided in




Table 7-3.  Seven pathways were considered: air submersion, water




submersion, surface contamination, inhalation, and ingestion of




vegetables, meat and milk.









     A number of factors contribute to the high population dose at the




urban site.  There are about 2.5 million persons within 80 kilometers of




the urban site and about 0.5 million within the same  distance of  the rural




site.  There also are larger numbers of meat animals  (about  700,000  versus




200,000) and dairy animals (about 38,000 versus 14,000)  and  more  vegetable




food crops produced (about 120,000 metric  tons versus 30,000 metric  tons)




within 80 kilometers of the urban site.  Meteorological  factors  also




contribute to the higher dose  at the urban site.








     Interim Storage of Solidified Waste




     Interim storage of solidified wastes  would cause releases similar  to




those for extended storage of  spent fuel, which was found  to have




negligible*impact.   DOE and EPA have estimated that the storage of






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                                 Table  7-1




    Estimated Dose Equivalents  to the Maximally Exposed  Individual from
3H
106Ru
ALL

Total
(millirem
1 yr 5 yr
0.5 **
1 **
2 **
Dose Equivalents

Solidification
(Rural Site)
Body
per year)
10 yr
0.3
0.0
0.4
Lower Large Intestine
(millirem per year)
1 yr 5 yr 10 yr
0.5 ** 0.3
16 ** 0.1
20 ** 0.4
Table 7-2
to Maximum Individual from Solidification
(Urban Site)
Total Body
(millirem per year)
1 yr 5 yr 10 yr
3H
106Ru
ALL
1.6 **
8.7 **
11 **
1.6
0.1
2.2
Table 7-3
Population Doses from Waste
1 yr
5 yr
10 yr
Rural
Total Body
(Person-rem)
23
8
6
Site
LLI
( Organ- rem)
100
11
6
Lower Larger Intestine
(millirem per year)
1 yr 5 yr 10 yr
1.6 ** 1.9
130 ** 0.3
140 ** 2.4
Solidification
Urban Site
Total Body LLI
(Person-rem) (Organ-retn)
200 800
90 100
40 50
**  not calculated
                                   138

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solidified waste would have even smaller impact,  since krypton-85 and




tritium, the only nuclides released in any quantity from spent-fuel




storage, are not contained in the solidified waste.








     Emplacement




     Emplacement of solidified wastes is essentially the same process as



emplacement of spent fuel.  The doses estimated by DOE are extremely small




(DOE 79).








7.1.3  Collection and Disposal of Krypton-85 and Iodine-129 Wastes




     Under our environmental standards for the Uranium Fuel Cycle




(40 CFR  190), krypton-85 and iodine-129 are considered to be wastes  that




must be  collected for disposal.  The doses from krypton-85 and  iodine-129




releases occurring at fuel reprocessing plants have been examined  in




detail  in connection with  the development of the standards in 40 CFR 190.




Therefore, we undertook  further  analysis of only the  doses associated with




these nuclides  after  removal  from  the  reprocessing plants.








     Krypton-85  Collection and  Disposal Options




     Our review indicates  a  number of  collection methods suitable for




krypton-85  (AL  79b) .  A  model reprocessing  plant would separate



 12.7 million curies of krypton-85  from spent  fuel  as  krypton gas.   This




gas would be contained  in cylinders,  either as pressurized  gas  or adsorbed




on zeolite  molecular  sieves.  Immobilization by ion implantation is being
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 investigated but has not been as fully developed as the other techniques.




 Either engineered  storage or geologic disposal  is suitable  for final




 disposal of the collected krypton.









     DOE has estimated krypton  leakage of 0.1 percent per year from




 storage in cylinders as a pressurized gas (DOE  79).  The gas may be stored




 onsite for eventual release to  the atmosphere.  For this disposal option,




 these releases would be in addition to discharges from normal operation.




 Under these conditions it would require about 52 years to meet the Uranium




 Fuel Cycle Standard limit of 50,000 curies per GWe-y.  Disposal in a mined




 geologic repository would provide smaller total releases, provided the




 containers were sufficiently stable in the mine environment to prevent




 increased leakage.









     Iodine-129 Collection and Disposal Options




     We also reviewed collection and disposal methods for iodine-129




 (AL 79b).  A generic reprocessing plant separates about 66  curies of




 iodine-129 per year in a total of about 600 to 650 kilograms of iodine.




The iodine can be immobilized either as insoluble barium iodate in




 concrete or by sorption on zeolites.  The immobilized iodine would then be




 suitable for terrestrial disposal,  although extraterrestrial or seabed




disposal options are being considered.









     No detailed characterization of the effects associated with




preparation of iodine-129 for final disposal is available.  However, these
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effects can be compared with the effects of similar treatments of




Low-level and intermediate wastes, which are insignificant (DOE 79).








7.1.4  Extraterrestrial Disposal




     Chemical processing is required if spent fuel is to be reprocessed




and is also required for transmutation and extraterrestrial disposal.




Both transmutation and extraterrestrial disposal of entire spent-fuel




elements appear to be impractical because of the amount of material




involved.  Chemical processing is required to separate the long-lived




radionuclides from the shorter-lived elements.  This reduces the material




to be transmuted or disposed extraterrestrially.  DOE has prepared




environmental assessments of conventional disposal technologies for spent




fuel and for chemically reprocessed high-level liquid wastes (DOE 79,




DOE 80).  Limited descriptions of the environmental effects of




technologies for extraterrestrial disposal and transmutation are available



(AL 79a, DOE 80).









     DOE and EPA analyses of the operations required to prepare high-level




waste for extraterrestial disposal indicate that separation of the




actinide fraction of the wastes (partitioning) would be required to avoid




an excessive number of launches of space vehicles  (AL 79a).  After




separation, nonactinide wastes would be disposed by terrestial methods.




The separated actinide wastes would be packaged to make a waste form




resistant to dispersion in the event of a launch accident or failure to




achieve proper orbit.  In some schemes, iodine-129 and other long-lived



fission products would also be sent into space.






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      There  is  insufficient  information  for  a  detailed  environmental




 analysis  of the  release  of  radionuclides  during  preparation  for




 extraterrestrial disposal and  in  the  event  of accidents  during launching




 (AL  79a;  DOE 80).  Since partitioning of  actinides  is  a  chemical  operation




 similar to  fuel  reprocessing,  it  is reasonable to expect somewhat similar




 releases.








 7.1.5  Transmutation




     Transmutation is the conversion  of radionuclides  into other  stable or




 radioactive nuclides, usually  by  irradiation  with neutrons.  Our  review of




 this disposal  option indicates that it  is in  the early conceptual design




 stage and that there is no  firm information on potential releases or doses




 (DOE 79).   The process requires that  the nuclides to be  transmuted be




 separated from the other materials.   The actinides must  be partitioned,




 and  it is usually necessary to further separate  individual actinides




 (fractionation).  The actinides are then  incorporated  into fuel rods for




 the  irradiation.  The requirements for transmutation are similar  to known




 processes in the nuclear fuel cycle—partitioning and  fractionation are




 similar to  fuel reprocessing, and incorporation of actinides into fuel




 rods is similar to fuel  fabrication.








 7.1.6  Other Separations




     There  have been many proposals for separation of various nuclides




preparatory to waste disposal (DOE 80):  separation for  special disposal




of long-lived nuclides (e.g., iodine-129, carbon-14, technetium-99) and
                                   142

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separation of short-lived nuclides,  particularly strontium-90 and




cesium-137, to minimize heat loading for geologic disposal.   Other




proposals include more complete separations of lanthanides and actinides




for space disposal or chemical synthesis of a waste form.  None of these




proposals is beyond the conceptual stage, and no information on




environmental releases or effects is available.  All of these separation




techniques are chemical processes similar to fuel reprocessing.








7.2  DERIVATION OF THE STANDARDS




     Our review of the waste management operations to be undertaken prior




to terrestrial disposal reveal that it is feasible, with available




technology and proper siting of facilities, to limit maximum individual




doses to those established by the Uranium Fuel Cycle Standard




(40 CFR 190).  Many of the processes to be undertaken are similar to




processes regulated under 40 CFR 190.  Therefore, we are confident these




processes can be regulated within these limits.  The individual doses from




some chemical processes have not been fully evaluated, but we  believe that




they can be expected to fall in the range of those associated  with fuel




reprocessing plants.  Storage operations for collected krypton-85 and




iodine-129 are also found to be within 40 CFR  190, although  special




provisions must be made for adequate time of decay before releasing




krypton to the atmosphere.
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     Although we have little information on the waste management




operations required for extraterrestrial disposal and transmutation, we




believe the limit established under 40 CFR 190 will not preclude these




options.









7.2.1  Selection of Standards Format




     The proposed standards for waste management operations are expressed




in terms of limits on annual radiation dose to individuals and to body




organs. This is the format used in our environmental standards for the




Uranium Fuel Cycle, 40 CFR 190.10(a).  For a number of reasons, it is




desirable to express the waste management operations standard in the same




format, unless further investigation shows this format is unsuitable:.




     1.  The Uranium Fuel Cycle standards are conceptually simple and




         relatively easy to implement.




     2.  Many of the procedures covered by the waste management operations




         standard are similar to operations covered by the Uranium Fuel




         Cycle standards.




     3.  It is possible that facilities regulated by the waste management




         operations standard will be sited adjacent to, or in proximity




         to,  facilities regulated under the Uranium Fuel Cycle standards.




         It would simplify procedures to use the same format in regulating




         these facilities.
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7.2.2  Selection of Numerical Limits




     Our evaluations of waste management operations indicate that



population doses are relatively small and that limiting doses to



individuals is the optimal method to control radionuclide releases.




Estimated releases are small enough that they conform to the 25 mrem/year



limit established in 40 CFR 190.








     Our estimates show that the highest individual doses for the waste




management operations preparatory to terrestrial disposal of either spent




fuel or reprocessed wastes are about 15 to 20 millirem per year  at a rural




site and about 135 millirem per year at an urban site.  These maximum




doses are caused by the solidification of 1-year-old  liquid wastes.  The




highest individual doses have not been estimated for  operations  serving




extraterrestrial waste disposal and  transmutation.  These operations are




similar to some of  the operations in the Uranium Fuel Cycle and  are,




therefore, expected to have  similar  doses.








     Dose Limits Higher than  those  in 40 CFR 190




     Setting  permissible dose  limits higher  than  those in 40 CFR 190




(e.g., a  level of  250 millirem per year) would  accomodate the doses




estimated for solidification  at an urban site and  would allow latitude for




other operations.   We do not  believe, however,  that there  is justification




for  setting  limits  above  those established  by 40  CFR 190.   Our




investigations  indicate that  there  is no need for latitude  in dose




limitations  for any operations other than solidification.   We estimate
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 that only a few solidification facilities will be required to carry out




 the waste management program and that most of these are expected to be




 operated in conjunction with fuel reprocessing.   Since fuel reprocessing




 requires large exclusion areas,  we see little or no problem with




 restricting waste  solidification to rural sites.




      Therefore we  reject dose limits higher than those in 40 CFR 190.








      Dose Limits Lower  than  those in 40  CFR 190




      Our assessment  is  that  the  waste management  operations required for




 terrestrial  disposal  could be carried out with annual  dose limits  to the




 maximally exposed  individual  below those required by 40 CFR 190.   The




 principal requirement is that extended storage (for periods greater than




 1 year)  be provided before solidification of  reprocessed wastes.   There is




 insufficient information to establish that  the operations in support of




 extraterrestrial disposal or  transmutation  could  meet  40 CFR 190.









      In  the  environmental impact  statement  for the  40  CFR 190  standards




 (EPA  76), we considered  the alternatives  of reducing the annual  whole-body




 dose  limits  from 25 millirem  to  15 millirem,  then to 5 millirem, with




 comparable reductions of permissible  dose  limits  to thyroid and  other




 organs.   We  judged the  improvements  to be "not significant" and  "small"




 respectively.  The improvements  in public health  from  dose limits  below




 25 millirem  for waste management  operations would be even smaller  and




would affect a smaller  industry.
                                   146

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     Implementation of standards with lower limits than those in




40 CFR 190 would be difficult.  Doses could be estimated by establishing




environmental transport models for measured effluent releases, but




confirmation of doses by direct measurement of environmental




concentrations of nuclides would be at or beyond the state of the art for



environmental monitoring.








     Establishment of dose limits substantially more restrictive than the




40 CFR 190 limits might preclude operations supporting extraterrestrial




disposal and transmutation.  Environmental protection must include all




parts of the management and disposal of spent  fuel and high-level and




transuranic wastes.  The possible benefits that might result  from further



investigation of extraterrestrial disposal and transmutation  outweigh the




small benefits of reducing the  limits for the  operational  standards.  The




Interagency Review Group has  recommended further  research  on




extraterrestrial disposal options (IRG 79).








     We reject dose  limitation  standards below those of  40 CFR  190,




because they would not provide  significant reductions  in risk.  They would




be more difficult to implement, and  they would preclude  consideration of



ottier potentially useful technologies.








     Limits Equal to those of 40 CFR 190




     As discussed in section  7.2.1,  administration  of  the individual dose




limits for the waste management operations standard  is  simplified if  its




format is the  same as that for  the Uranium Fuel Cycle  standards.   There






                                   147

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is, in our judgment, no justification  for setting the  individual dose




limits above those established by 40 CFR 190.  There is  little to be




gained in setting the limit lower than 40 CFR 190.  Therefore, we have




chosen individual dose limits equal to those of 40 CFR 190 as the proposed




standard.









     The proposed standard establishes limits of 25 millirem per year to




the whole body, 75 millirem per year to the thyroid, and 25 millirem to




any other organ of any member of the general population.  These limits are




known to be readily itnplementable, are protective of the public, and do




not impose prohibitive costs.   These dose limits would permit the use of




available monitoring capabilities and would require the use of the best




available technology to prevent excessive radiation exposure of the




public.  We estimate that these limits would'not preclude extraterrestrial




disposal or transmutation.









7.2.3  Implications of the Operational Standards




     The proposed waste management operations standards extend the




protection of 40 CFR 190 to similar activities not included under that




regulation.   With the exception of transportation, generally applicable




environmental standards will apply to all operations in the generation of




commercial electrical power, from milling of uranium ore through




emplacement of wastes in their final disposal site.
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     The proposed standards provide that doses to individuals from




combined operations covered by 40 CFR 190 and 40 CFR 191 shall not exceed




the limits of either part.  This provision is analagous to the provision




that the dose limits for any individual from any combination of the




activities covered by 40 CFR 190 shall be the same as those from any




single activity. Together, the two standards prevent an individual from




receiving excessive doses from any combination of nuclear activities.
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150

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                                Chapter 8




                          ENVIRONMENTAL IMPACTS








     Our proposed standards address the operations required to store,




prepare, and dispose of high-level and transuranic wastes,  as well as  the




long-term performance of their disposal systems over 10,000 years.  The




intent of these standards is to minimize releases of radionuclides to  the




environment and to assure that the risks to the public health and safety




are small.








8.1  HEALTH IMPACT








8.1.1  Standards for Waste Management Operations




     The proposed standards for waste management operations limit the




maximum dose to an individual in the general population to 25 millirem per




year.  Evaluations of emissions from existing facilities indicate that




actual doses can be expected to be much lower than 25 millirem per year,




and will generally range between 5 and 10 millirem per year.  However, the




possible proximity of waste management facilities to other operations  of




the uranium fuel cycle already governed by a limit of 25 millirem per  year




under 40 CFR 190—and the difficulty of measuring such small  emissions—




make it impractical to impose a smaller limit.   In addition,  we  do not




want to preclude development of alternative disposal methods  that, while




they may  lead  to slightly  larger doses to  some people during  the




operational phase, could lead to a significantly lower long-term risk  to
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 future generations.  The  overall  risk  to  populations,  as  opposed  to  the




 risk  to  individuals, from releases  from management  operations  for




 high-level and  transuranic waste  disposal  is  expected  to  be  small, because




 we anticipate that only a few  such  facilities will  be  needed to process




 both  existing and future  inventories.  Thus,  we  conclude  that  the risk to




 both  individuals and to populations  from management operations for




 high-level waste disposal carried out  according  to  these  standards will be




 small and will  present no significant  hazard  to  the public health and




 welfare.









 8.1.2  Disposal




     The proposed standards for disposal  set  limits for releases  of




 specific radionuclides over the first  10,000  years  after  disposal.   The




 maximum  impact  expected over this time period from  disposal  in compliance




 with these containment requirements  is estimated  to be about 10 excess




 premature deaths from cancer for  each  1000 metric tons of heavy metal




 (MTHM) charged  to the reactor.  In making  this estimate,  we  have  used an




 approach we believe makes  the estimates conservatively large.









     We expect  that the first method to be utilized for disposal  of




high-level radioactive and transuranic wastes will  be  disposal in a  mined




 repository deep in a continental  rock  formation.  To obtain  a basis  for




judging whether such disposal will meet our standards, we have made  risk




assessments for a model repository in different geologic  media.   The model




repository is similar to current  concepts of  repositories discussed  in
                                   152

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recent literature.  The model assumes release of radionuclides might occur




by any of several modes after disposal in the repository.  The




radionuclides considered are those that our evaluation indicates have a




potential to make a significant contribution to the total exposure




expected from the release of a fraction of the radioactive waste.  Our




estimates show about 700 expected health effects from a repository in




granite and about 100 from a repository in bedded salt over 10,000 years.




We judge this impact to be small.









8.2  CONTAMINATION OF THE ENVIRONMENT




     The limits that the standards for disposal set for releases to  the




accessible environment generally prevent contamination of any part of that




environment to a level that would interfere with its use.  We made the




health impact estimates discussed in -section 8.1.2 on the assumption that




the accessible environment was used without restriction once  100 years had



passed after disposal.









     The standards for disposal do not, however, limit releases  to  sources




of groundwater within 10 kilometers  from the emplacement of  these




radioactive wastes.  Our assessment  of individual doses  has  shown  that




inadvertent human intrusions for resource  exploration could  significantly




contaminate groundwater areas of a few tens of  thousands of  square




meters.  More serious contamination  could  occur as a result  of  severe




improbable natural events such as faulting.
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     However, the standards for disposal do tend to restrict the nature of




the sources of groundwater that could be contaminated.  Because the




standards impose limits on the amounts of radioactive nuclides that can




reach groundwater more than 10 kilometers from the point of emplacement of




wastes, any unprotected groundwater must move very slowly.  Slowly moving




groundwater is usually contained in relatively impermeable formations that




are poor producers of water and, therefore, are unlikely to be exploited.




Although a certain amount of groundwater may become contaminated—and lost




for potential future use—we believe that this minor resource loss would




be justified by the very good long-term protection that would be provided




by disposal systems complying with our standards.








8.3  CONCLUSION




     The proposed standards for waste management operations limit doses to




persons in the general population to 25 mrem/year.  In actual practice,




few people are expected to be exposed, and doses are expected to be




substantially below this limit.  Therefore, we conclude that these waste




management operations do not represent a significant risk to the health




and safety of the population.









     The maximum impact expected from a disposal system in compliance with




this standard is estimated to be less than 1,000 premature cancer deaths




over the first 10,000 years for disposal of the high-level wastes from




100,000 MTHM of reactor fuel (about as much as all currently operating




reactors should produce over their lifetimes).  The environmental impacts
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of the alternatives we considered range up to about 10 times as large as




this; these impacts are discussed in Chapter 4.  Our analyses indicate




that construction of repositories in bedded salt or granite can achieve



our proposed containment requirements, and repositories  in other geologic




media may as well.  Comparison with other environmental  radiation hazards




indicates that disposal in compliance with our proposed  standards would




represent a very small added risk to the population over such an extended




period of time.
                                   155

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                                Chapter 9




                            REGUIATORY IMPACT








     This chapter reviews the projected costs associated with management




and disposal of high-level radioactive waste, and it summarizes the




potential regulatory impacts of our proposed standards.  The details of



the analyses described here are presented in our Regulatory Impact




Analysis (EPA 82), prepared in accordance with Executive Order 12291.








     The situation regarding the disposal of high-level waste is unusual




from a regulatory standpoint.  In most cases, a regulation concerns an




ongoing activity.  Any modifications that the regulation causes in  the




activity may be considered  to be costs that  should be  outweighed by the




regulatory benefits.  For high-level waste disposal, however, the



appropriate regulations must be developed well before  the activity  to  be




regulated can  even begin.   Thus, the typical perspectives about balancing




regulatory costs and  benefits do not apply.








     To  investigate  the  potential  impacts  of this proposed  action,  we




evaluated how  the costs  of  high-level  waste  management and  disposal might




change due  to  alternative  stringency  levels  of our containment




requirements—or due to  changes in our assurance requirements.   Because




there  is no "baseline" program to  consider,  we could not quantify the




costs  and benefits  of our proposed action compared to the consequences of




no regulation.
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     The most important benefit of our action should be the assurance that




these wastes will be disposed of with adequate protection of public health




and the environment.  This assurance, in turn, should allow the Federal




program to proceed expeditiously to develop acceptable disposal methods at




appropriate sites.  It may be argued that a further benefit would be the




resolution of a key issue that might lead to expanded commercial use of




nuclear power.  This would be a benefit if nuclear power has clear




advantages, economic and otherwise, compared to alternative methods of




generating electricity; however, we have not analyzed this issue.









     Our containment requirements consist of limits on potential releases




of radioactivity from a disposal system; these limits are to be used as




overall design objectives.  These requirements are stated in terms of




projected releases for 10,000 years after disposal of the wastes.  To




judge the risks associated with these release limits, we have used




generalized environmental pathway models to assess the potential health




impacts of the releases that would be allowed by our standards.  However,




calculations of these "residual risks" are clearly not reliable as




absolute values, since projections of population distributions, ways of




life, and human behavior over 10,000 years cannot be meaningful.  Rather,




these calculations are valuable only for understanding the relative




"residual risks" from different sources of radiation exposure (such as




risks from different disposal system designs, or risks from natural ore




bodies).
                                   158

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     For the release limits we have proposed as part of our environmental




standards,  the residual risks projected by these models would be less than




1,000 premature deaths from cancer over the 10,000 year period, an average




of one premature death every 10 years.   To judge the effects on disposal




costs of changing the level of protection, we also compared release limits




corresponding to residual risk values of: 100, 1,000, 5,000, and 10,000




premature deaths over the 10,000 year period.  We chose this range of




residual risks because it appears to represent the range of performance




that may be expected of mined geologic repositories.









     To do this analysis, we evaluated the long-terra performance of




generic models of geologic repositories in three different geologic media:




bedded salt, granite, and basalt.  We did the analysis  in two  steps:









     First, we used our performance projections to assess the  quality of



the engineering controls that would be needed in each of the three model




repositories to meet each of the four different levels  of protection.




In doing so, we encountered the problem that development of  specific




engineered barriers (e.g., waste forms and canisters) has not  yet




progressed enough to clearly associate the costs of manufacturing  these




engineered barriers with their performance levels.  Thus, we had  to  make




some rather speculative judgments to associate disposal costs  with




alternative stringency levels.  The results  of this analysis are  displayed



in Figure 9-1.
                                   159

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c*
o
     600





5    soo




S    400


VI


"    300
i

S    100
S
                      SALT  REPOSITORY
                                          (Engineering  Barrier Costs Only)
                                               600
                                               400
                                               200
                                           GRANITE  REPOSITORY
                                                                    600'
                                                                    400
                                                                    200
                                                                          BASALT REPOSITORY
                        100    1000   5000   10000         100    1000   5000   10000          100   1000   5000   10000



                              — . 	 _ level of protection (health effects over 10,000 years) 	
                        FIGURE 9-1:   VARIATIONS IN  WASTE MANAGEMENT  COST vs.  LEVEL OF PROTECTION

-------
      Second, we  tried to allow for  the possible  effect  of  alternative




 stringency  levels on site selection.  This  is particularly relevant



 because  our analyses indicate that  the most important part of  the




 protection offered by a mined geologic repository comes  from the




 hydrological and geochemical characteristics of  the  site itself.




 The costs of usin8 a "good" site rather than a "bad" site  (within  the  same




 type  of  geologic media) do not involve differences in construction cost.




 Instead, they  involve the difficulty of finding  a site  that is "good




 enough." Since  there are so few data on  site characterization, we have no




 good  basis for judging how many sites might have to  be  studied to  meet




 different levels.  However, we did  made some assumptions about how site




 selection costs might increase in order to  meet  more stringent standards.




 We then  combined these assumptions  with our evaluations  of the variations




 in engineered barrier costs to arrive at  our second  set  of disposal cost




 estimates.   The results from this analysis  are shown in  Figure 9-2.








     The results of these assessments of  disposal costs  and alternative




 stringency levels indicate that the costs are not very sensitive to




different levels of protection,  particularly for the geologic media




 (bedded  salt and granite) that are  better at reducing long-term risks.




Even with our hypothesis about increased  site selection  costs with more




 stringent levels, the difference in costs for different  levels are much




smaller  than the overall uncertainties in waste management costs.  For




example, consider the increased costs of  complying with  the release limits




we have proposed, rather than release limits 10  times less stringent.
                                   161

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                    (Engineering Barrier Costs and  Site Selection Costs)
     600





l|    500-


o>
.*
-v.

S   400

+j
VI
o

^    300
i
in
s
     200




     100
           SALT REPOSITORY
                                     600    GRANITE REPOSITORY      600l    BASALT REPOSITORY
400
200
                                                                     400-
                                                                     200
            100    1000    5000  10000          100   1000    5000   10000          100   1000    5000   10000



                  . 	 ...  level Of protection (health effects over 10,000 years) 	
           FIGURE  9-2:  VARIATIONS  IN WASTE  MANAGEMENT COST  vs.  LEVEL  OF PROTECTION

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The potential increase ranges from zero to 50 million (1981) dollars per




year.  For comparison, the total costs of high-level waste management and




disposal (independent of our action) have been estimated as between




700 million and almost 1.5 billion (1981) dollars per year.  Electrical




utility revenues were about 100 billion dollars in 1980.








     These analyses, while indicating that disposal costs appear to be




relatively insensitive to differences in the  level of protection, do not




provide a way to determine the acceptability  of the residual risks from a




societal perspective, nor do they indicate a  level of protection that is




preferable from a balancing of costs and benefits.  One possible approach




to balancing costs and benefits would be to judge the cost  per  life saved




by different levels of protection, perhaps taking into account  some method




of discounting costs and benefits.  However,  our calculations of residual




risks  are not reliable as absolute values.  Thus, we have no meaningful




way  to calculate an absolute value of the cost per  life saved by different




levels of protection.








     In the  absence of the ability  to make meaningful cost  and  benefit




comparisons, we have used other tests of economic feasibility  and




acceptability of risk to judge  the  appropriateness  of the level of




protection we have proposed.  As  discussed above, setting the  release




 limits at the  level we chose—as  opposed  to  a level 10  times less  or




10  times more  stringent—appears  to  cause only very minor effects  on the




costs  of high-level waste disposal.  To judge the acceptability of the




remaining long-term risk, we  considered the  risks that  would otherwise be






                                   163

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 caused  if  the uranium ore  used  Co  produce  Che  wastes  had  not been mined.

 The magnitude of  the  risks from these  unmined  ore  bodies  is  very  uncertain

 due,  in part, to  the  wide  variety  of settings  in which  uranium ore is

 found—many of which  are closer to the surface than a geologic repository

 would be.  Using  the  same  generalized  environmental pathway  models that

 were  used  to assess the risks from our models  of geologic  repositories,

 the risks  from a  comparable amount  of  untained  uranium ore  are  estimated  to

 range from a few  hundred to more than  1 million health  effects over

 10,000  years (WI  80).  The lower end of this range is roughly  equal to the

 residual risk associated with our  proposed release limits.   Thus,  the

 upper limit of the risk that our standards would allow  from  the disposal

 of high-level wastes  appears to pose a threat  very close to  the minimal

 risk  posed by nature,   had  the uranium  ore never been  mined and the

 high-level wastes never been generated.



     Our assurance requirements provide seven  qualitative criteria that

 should reduce some of  the uncertainties inherent in disposing  of wastes

 that must be isolated   for a very long  time.  The specific provisions of

 these assurance requirements are described in Chapter 6 of this document.

Only three of the criteria have a significant potential to increase the

costs of high-level waste disposal.  These are:

     Criterion 2,  which calls for disposal systems to keep  radioactive
     releases as  small as reasonably achievable;

     Criterion 3,  which calls for disposal systems to use multiple
     barriers,  botti engineered and natural; and

     Criterion A,  which restricts reliance on  active institutional
     controls  to  a resonable period after disposal (e.g., a  few hundred
     years).
                                   164

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     Each of these three assurance requirements might have the effect of




requiring better engineered barriers than would otherwise be needed to




meet our environmental standards.  This would be particularly true for a.




repository sited in a relatively good geologic media (such as our generic




models for bedded salt or granite).  However, even if no engineered




barriers at all appeared to be needed for long-term protection after




disposal, fairly protective canisters and waste forms would be needed for




other phases of waste management, such as transportation to and




emplacement in a repository.  Therefore, we believe that these assurance




requirements would necessitate—at most—only moderate improvements  in




waste form performance, and we judged that the impact that these




improvements might have on disposal costs should be less than 10 million




(1981) dollars per year.  Since this impact concerns improvements to




engineered barriers, the potential cost increase would be duplicative of




any engineered barrier impacts caused by our environmental standards.




Thus, the potential cost effects of our containment requirements and our




assurance requirements should generally not be added together.   (For some




unusual possibilities, adding the effects of the containment  requirements




and the assurance requirements might be appropriate, but  these




possibilities would tend to involve relatively small impacts.)
                                  165

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166

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      Appendix A
THE PROPOSED STANDARDS
         167

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168

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     A new Part  191  is proposed to be added to Title 40, Code of Federal




 Regulations,  as  follows:









               SUBCHAPTER F - RADIATION PROTECTION PROGRAMS









 PART  191  - ENVIRONMENTAL RADIATION PROTECTION STANDARDS FOR




     MANAGEMENT  AND  DISPOSAL OF SPENT NUCLEAR FUEL, HIGH-LEVEL AND




     TRANSURANIC RADIOACTIVE WASTES









      Subpart A  - Environmental Standards  for Management and Storage




 191.01    Applicability




 191.02    Definitions




 191.03    Standards  for Normal Operations




 191.04    Variances  for Unusual Operations




 191.05    Effective Date









              Subpart B - Environmental Standards  for Disposal




 191.11   Applicability




 191.12   Definitions




 191.13   Containment Requirements




 191.14   Assurance Requirements




 191.15   Procedural  Requirements




 191.16   Effective Date









AUTHORITY:  The Atomic Energy Act of 1954, as amended; Reorganization Plan




No. 3 of 1970.






                                   169

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      SUBPART A - ENVIRONMENTAL STANDARDS FOR MANAGEMENT AND STORAGE









191.01  Applicability




     This Subpart applies to radiation doses received by members of the




public as a result of the management (except for transportation) and




storage of spent nuclear fuel, high-level, or transuranic radioactive




wastes, to the extent that these operations are not subject to the




provisions of Part 190 of Title 40.








191.02  Definitions




     Unless otherwise indicated in this Subpart, all terms shall have  the




same meaning as in Subpart A of Part 190.




     (a)  "Spent nuclear fuel" means any nuclear fuel removed from a




nuclear reactor after it has been irradiated.




     (b)  "High-level radioactive wastes" means any of the following that




contain radionuclides in concentrations greater than those identified  in




Table 1: (1) liquid wastes resulting from the operation of the first cycle




solvent extraction system, or equivalent, in a facility for reprocessing




spent nuclear fuels;  (2) tJ-<=>. concentrated wastes from subsequent




extraction cycles, or equivalent; (3) solids into which such liquid wastes




have been converted;  or (4) spent nuclear fuel if disposed of without




reprocessing.




     (c)  "Transuranic wastes," as used in this Part, means wastes




containing more than 100 nanocuries of alpha emitting transuranic




isotopes,  with half-lives greater than one year, per gram of waste.
                                   170

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     (d)  "Storage" means placement of radioactive wastes with planned




capability to readily retrieve such materials.




     (e)  "Management and storage" means any activity, operation, or




process, except for transportation, conducted to prepare spent nuclear




fuel, high-level or transuranic radioactive wastes for storage or




disposal, the storage of any of these materials, or activities associated




with the disposal of these materials.




     (f)  "General environment" means the total terrestial, atmospheric,




and aquatic environments outside sites within which any  operation




associated with the management and'storage of spent nuclear fuel,




high-level or transuranic radioactive wastes  is conducted.




     (g)  "Member  of the public" means any individual who  is  not engaged




in operations involving  the management,  storage,  and  disposal of materials




covered by these standards.  A worker so engaged  is a member  of  the public




except  when on duty  at a  site.








191.03   Standards  for Normal  Operations




     Operations  covered  by  this  Subpart  should  be conducted so as to




reduce  exposures to members of  the public to the extent reasonably




achievable,  taking into  account  technical,  social, and economic




considerations.  As an upper  limit, except for variances in accordance




with 191.04,  these operations shall be conducted in such a manner as to




provide reasonable assurance  that the combined annual dose equivalent to




 any member of the public due to: (a) operations covered by Part 190,
                                    171

-------
(b) planned discharges of radioactive material  to  the general environment




from operations covered by this Subpart, and (c) direct radiation from




these operations; shall not exceed 25 millirems to the whole body,




75 millirems to the thyroid, or 25 millirems to any other organ.









191.04  Variances for Unusual Operations




     (a)  The implementing agency may grant a variance temporarily




authorizing operations which exceed the standards  specified in  191.03 when




abnormal operating conditions exist if: (Da written request justifiying




continued operation has been submitted, (2) the costs and benefits of




continued operation have been considered to the extent possible, (3) the




alternatives to continued operation have been considered, and (4)




continued operation is deemed to be in the public  interest.




     (b)  Before the variance is granted, the implementing agency shall




announce, by publication in the Federal Register and by letter  to the




governors of affected States: (1) the nature of the abnormal operating




conditions,  (2) the degree to which continued operation is expected to




result in doses exceeding the standards, (3) the proposed schedule for




achieving conformance with the standards, and (4)  the action planned by




the implementing agency.









191.05  Effective Date




     The standards in this Subpart shall be effective 12 months  from the




promulgation date of this rule.
                                   172

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             SUBPART B - ENVIRONMENTAL STANDARDS FOR DISPOSAL








191.11  Applicability




     This Subpart applies to radioactive materials released into the




accessible environment as a result of the disposal of high-level or




transuranic radioactive wastes,  including the disposal of spent nuclear




fuel.  This Subpart does not apply to disposal directly into the oceans or




ocean sediments.








191.12  Definitions




     Unless otherwise indicated in this Subpart, all terms shall have the




same meaning as in Subpart A of this Part.




     (a)  "Disposal" means isolation of radioactive wastes with no  intent




to recover them.




     (D)  "Barriers" means any materials or structures that  prevent or




substantially delay movement of the radioactive wastes toward  the




accessible environment.




     (c)  "Disposal  system" means any  combination  of engineered and




natural  barriers  that contains radioactive wastes  after  disposal.




      (d)  "Groundwater" means water below the land surface in  a zone  of




saturation.




      (e)  "Lithosphere" means the solid  part  of the Earth, including  any




groundwater  contained within  it.
                                    173

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     (f)  "Accessible environment" includes (1) the atmosphere,  (2)  land




surfaces, (3) surface waters, (4) oceans, and (5) parts of the lithosphere




that are more than ten kilometers in any direction from the original




location of any of the radioactive wastes in a disposal system.




     (g)  "Reasonably foreseeable releases" means releases of radioactive




wastes to the accessible environment that are estimated to have more than




one chance in 100 of occurring within 10,000 years.




     (h)  "Very unlikely releases" means releases of radioactive wastes to




the accessible environment that are estimated to have between one chance




in 100 and one chance in 10,000 of occurring within 10,000 years.




     (i)  "Performance assessment" means an analysis which identifies




those events and processes which might affect the disposal system,




examines their effects upon its barriers, and estimates the probabilities




and consequences of the events.  The analysis need not evaluate  risks  from




all identified events.  However, it should provide a reasonable




expectation  that  the  risks from  events not evaluated are  small  in




comparison to  the  risks which  are estimated  in  the  analysis.




      (j)  "Active  institutional  controls" means  (1) guarding  a  disposal




site, (2) performing  maintenance operations  or  remedial  actions at  a




disposal  site, or  (3) controlling or  cleaning up  releases from  a disposal




site.
                                   174

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     (k)  "Passive institutional controls" means (1) permanent markers




placed at a disposal site, (2) public records or archives, (3) Federal




Government ownership or control of land use, or (4) other methods of




preserving knowledge about the location, design, or contents of a disposal




system.




     (1)  "Heavy metal" means all uranium, plutoniutn, or thorium placed



into a nuclear reactor.









191.13  Containment Requirements




     Disposal systems for high-level or transuranic wastes shall be




designed to provide a reasonable expectation that  for 10,000  years after




disposal:




     (a)  Reasonably foreseeable releases of waste  to the accessible




environment are projected to be less than the quantities  calculated




according to Table 2.




     (b)  Very unlikely releases of waste to the accessible environment




are projected to  be less  than  ten times the quantities  calculated



according to Table 2.









191.14  Assurance Requirements




     To provide the confidence needed  for compliance with the containment




requirements of 191.13, disposal of  high-level  or  transuranic wastes shall




be conducted in accordance with  the  following  requirements:
                                   175

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     (a)  Wastes shall be disposed of promptly once disposal systems are




available and the wastes have been suitably conditioned for disposal.




     (b)  Disposal systems shall be selected and designed to keep releases




to the accessible environment as small as reasonably achievable, taking




into account technical, social, and economic considerations.




     (c)  Disposal systems shall use several different types of barriers




to isolate the wastes from the accessible environment.  Both engineered




and natural barriers shall be included.  Each such barrier shall




separately be designed to provide substantial isolation.




     (d)  Disposal systems shall not rely upon active institutional




controls to isolate the wastes beyond a reasonable period of time (e.g., a




few hundred years) after disposal of the wastes.




     (e)  Disposal systems shall be identified by the most permanent




markers and records practicable to indicate the dangers of the wastes and




their location.




     (f)  Disposal systems shall not be located where there has been




mining for resources or where there is a reasonable expectation of




exploration for scarce or easily accessible resources in the future.




Furthermore, disposal systems shall not be located where there is a




significant concentration of any material which is not widely available




from other sources.




     (g)  Disposal systems shall be selected so that removal of most of




the wastes is not precluded for a reasonable period of time after disposal.
                                   176

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191.15  Procedural Requirements




     Performance assessments to  determine compliance with the  containment




requirements of 191.13 shall be  conducted in accordance with the following:




     (a)  The assessments shall  consider realistic projections of the




protection provided by all of the engineered and natural barriers of a




disposal system.




     (b)  The assessments shall  not assume that active institutional




controls can prevent or reduce releases to the accessible environment




beyond a reasonable period (e.g., a few hundred years) after disposal.




However, it should be assumed that the Federal Government is committed to




retaining passive institutional control of disposal sites in perpetuity.




Such passive controls should be effective in deterring systematic or




persistent exploitation of a disposal site, and it  should be assumed that




they can keep the chance of inadvertent human  intrusion very small as long




as the Federal Government retains such passive control of disposal sites.




     (c)  The assessments shall use information regarding the  likelihood




of human intrusion, and all other unplanned events  that may cause releases




to the accessible environment, as determined by the implementing agency




for each particular disposal site.









191.16  Effective Date




     The standards in this  Subpart shall  be effective immediately upon




promulgation of  this  rule;  however, this  Subpart  does not apply to  wastes




disposed of before promulgation  of this  rule.
                                   177

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    TABLE  1 - CONCENTRATIONS  IDENTIFYING HIGH-LEVEL  RADIOACTIVE WASTES
     Radionuclide                                       Concentration
                                                  (curies per gram of waste)
     Carbon-14   	8x 10~6


     Cesium-135	8x 10~4


     Cesium-137	5x 10~3


     Plutonium-241	3x 10~6


     Strontium-90	  7x 10~3


     Technetium-99	3x 10~6


     Tin-126	7x 10"7


     Any alpha-emitting transuranic


       radionuclide with a half-life	Ix 10~7


       greater than 20 years


     Any other radionuclide with a half-life

                                                                  _o
       greater than 20 years	1x10
     NOTE:  In cases where a waste contains a mixture of radionuclides, it


shall be considered a high-level radioactive waste if the sum of the


ratios of the radionuclide concentration in the waste to the concentration


in Table 1 exceeds one.
                                   178

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     For example,  if a waste containing radionuclides A, B, and C  in




concentrations Ca, C^, and Cc,  and if the concentration  limits from




Table 1 are GLa, CL^, and CLC,  then the waste shall  be considered




high-level radioactive waste if the following relationship exists:
                 .ca       cb       cc
                  CLa      CLb      CLC
                                    179

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      TABLE 2 - RELEASE LIMITS FOR CONTAINMENT REQUIREMENTS

        (Cumulative  Releases  to  the Accessible Environment

                 for 10,000 Years After Disposal)
Radionuclide
     Release Limit
(curies per 1000 MTHM)
Americiura-241 ---------------------     10

Americium-243 ---------------------     4

Carbon-14	   200

Cesium-135	2000

Cesium-137	   500

Neptunium-237	     20

Plutoniura-238 	   400

Plutoniura-239 	   100

Plutonium-240	•	   100

Plutoniura-242 	   100

Radiura-226	     3

Strontium-90	     80

Technetium-99	10000

Tin-126	     80

Any other alpha-emitting

  radionuclide  --------------------     10


Any other radionuclide which does

  not emit alpha particles  --------------   500
                              180

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     NOTE 1:  The Release Limits in Table 2 apply either to the amount of

high-level wastes generated from 1,000 metric tons of heavy metal (MTHM),

or to an amount of transuranic (TRU) wastes containing one million curies

of alpha-emitting transuranic radionuclides.  To develop Release Limits

for a particular disposal system, the quantities in Table 2 shall be

adjusted for the amount of wastes included in the disposal system.

For example:




     (a)  If a particular disposal system contained the high-level wastes

from 50,000 MTHM, the Release Limits for that system would be the

quantities  in Table 2 multiplied by 50 (50,000 MTHM divided by 1,000 MTHM).

     (b)  If a particular disposal system contained five million curies  of

alpha-emitting transuranic wastes, the Release Limits for that system

would be the quantities in Table 2 multiplied by five (five million curies

divided by  one million curies).

     (c)  If a particular disposal system contained both  the high-level

wastes  from 50,000 MTHM and 5 million curies of  alpha-emitting  transuranic

wastes, the Release Limits for  that  system would be the quantities  in

Table 2 multiplied by 55:


              50,000 KTHM     5,000,000  curies TRU
              	  +  	;	 „  55
                1,000 KTHM      1,000,000  curies TRU
                                   181

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     NOTE 2:  In cases where a mixture of radionuclides  is projected




to be released, the limiting values shall be determined  as follows:




For each radionuclide in the mixture, determine the ratio between  the




cumulative release quantity projected over  10,000 years  and the  limit




for that radionuclide as determined from Table 2 and Note 1.  The  sum




of such ratios for all the radionuclides in the mixture  may not  exceed



one.









     For example, if radionuclides A, B, and C are projected to  be




released in amounts Qa, Qb, and Qc, and if  the applicable Release




Limits are RLa, RL^, and RLC, then the cumulative releases over




10,000 years shall be limited so that the following relationship exists:
                  Qa       Qb       Qc




                  KL       RL       RL
                                   182

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               Appendix B
RISK ASSESSMENTS OF GEOLOGIC REPOSITORIES
                  183

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184

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     Our assessments of the health risks associated with disposal  in mined




geologic repositories are intended to provide conservative estimates of




the potential risks from our model repository systems.   (The repository




characteristics were chosen to be stringent enough for adequate




protection, but lenient enough to be reasonably achievable.)  More




sophisticated risk assessments of these model repositories—using more




realistic repository performance characteristics—should predict less harm




than our estimates.  Some potential repository sites should be more




environmentally acceptable than our models; others could be less




environmentally acceptable.  The Department of Energy and the Nuclear




Regulatory Commission will perform assessments of specific candidate




repository sites.









     Our risk assessments  include estimates of the potential harm to a




representative population and estimates of the annual dose  equivalents to




highly  exposed individuals.  We chose to express  the harm to the




population as numbers of fatal cancers  and numbers of genetic  effects—and




the  individual dose  assessments in annual dose equivalent—rather  than as




the  quantities or  concentrations  of  radioactive materials released  to  the




environment.  This approach  allows us to combine  the  effects  of the




various radionuclides  and  to represent  more  clearly  the potential risks  of




high-level waste disposal.









     By varying  the hypothetical  conditions  of  our generic  model




repositories, we can explore a  wide  variety  of  circumstances.   These




projections  then  become a  tool  for comparing various  conditions as they






                                   185

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affect repository performance.  Some postulated conditions may produce

risks too nigh to be considered acceptable; others may greatly reduce the

risks but may be technologically or economically unattainable.  The full

report of our population risk assessment and our individual dose

assessment are presented in Smith, et al. (SMC 82) and Goldin, et al.

(GO 82), respectively.



B.I  GENERAL FEATURES OF THE MODEL

     The model system used for our assessment is made up of five parts:

     1.  the waste container;

     2.  the physical and chemical characteristics of the waste itself
         (the waste form);

     3.  the geologic medium into which the waste is placed (the host rock);

     4.  the geologic media between the host rock and man's
         environment (the geosphere);

     5.  the environmental (i.e., above ground) pathways through which
         people can come into contact with waste radionuclides
         (the biosphere).


     The first four parts are characteristics of the disposal system and

are common to both the population and the individual risk assessments; the

fifth part, while considered in both assessments, is treated differently

because of the different approaches needed to assess individual and

population exposures.  These environmental pathways are discussed more

completely by Smith, Fowler, and Goldin (SMJ 82).  The events and

processes leading to radionuclide releases are discussed in section B.2,

together with the population risk assessments.  The individual dose

assessments are presented in section B.3.
                                   186

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B.I.I  Waste Characteristics




     Table B-l lists the major radionuclides contained in the radioactive



inventory of our model repository.  We selected these 15 radionuclides on




the basis of their quantity in spent fuel, their half-lives, and their




biological behavior as reflected in dose-equivalent conversion factors




(rem per curie ingested or inhaled).  Table B-2 lists the geochemical



characteristics (i.e., retardation factors and solubility limits) of each




of these radionuclides.








     Each model repository contains the waste from 100,000 metric tons of




heavy metal (MIHM) charged to reactors.  We assume the waste is in the



form of unreprocessed spent fuel, since this assumption requires




examination of the behavior of both the fission products and the




transuranic nuclides.  The composition by weight of the spent fuel is




assumed to be 95.5 percent uranium isotopes, 0.9 percent plutonium




isotopes, 0.1 percent other transuranic isotopes, and 3.5 percent fission




products.  We assume all the  spent fuel has aged for  approximately 10




years after discharge from  the reactors.  More detailed  information  on  the



repository  inventory is given by Arthur D. Little, Inc.  (ADL 79).








     Continued  integrity of the waste containers assures  retention  of the




radioactive wastes  in the repository.  If the  containers  fail,  release  of




the radioactive material depends  on either  the release  rate of  nuclides




from the fuel matrix (waste form) or the  solubility  of  the  radioactive



elements as they  come  into  contact with the groundwater.  For our
                                   187

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             Table B-l




Characteristics of Radioactive Waste
Radionuclide
Pu-238
Pu-239
Pu-240
Pu-242
Am-241
Am- 243
Np-237
Cs-135
Cs-137
1-129
Tc- 99
Sn-126
Zr- 93
Sr- 90
C- 14
(Spent Fuel)
Initial Quantity
in Repository (ADL 79)
(curies)
220,000,000
33,000,000
49,000,000
170,000
17,000,000
1,700,000
33,000
22,000
8,600,000,000
3,800
1,400,000
56,000
190,000
6,000,000,000
28,000
Half-Life
(years)
89
24,400
6,260
380,000
458
7,650
2,140,000
3,000,000
30
16,000,000
210,000
100,000
950,000
29
5,730
               188

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                       Table B-2
   Solubility Limits and Retardation Factors  (SMC  82)
                       Solubility Limit
                            3
 Radionuclide          (Ci/m )     (ppm)     Retardation Factor
    Pu-239
    Pu-240
    Pu-242
    Am-241
    Am-243
    Hp-237
    Cs-135
    Cs-137
      1-129
    Tc-  99
    Sn-126
    Zr-  93
    Sr-  90
      C-  L4
6.0E-5
2.2E-4
4.0E-6
160
10
7.2E-7
na
na
na
2.0E-5
3 . OE-2
na
na
na
0.001
0.001
0.001
50.0
50.0
0.001
na
na
na
0.001
1,0
na
na
na
100
100
100
100
100
100
1
I
I
1
10
1
1
1
na  =     Solubility limit not used or not known
ppm = parts radionuclide per million parts solution by weight.
                             189

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 reference cases, containers last 100 years in salt and 500 years in other



 media.  For the release rate,  we assume that, after failure of the



 container, a constant fraction of the remaining radionuclide dissolves per



 unit time in the repository groundwater.   For our reference cases, this


                                             -4
 fraction is taken to be 1 part in 10,000  (10  ) per year.   The



 solubility limits become important  when there is not enough water to



 dissolve the quantity of the radioactive  element that could otherwise be



 released.   This is usually the case for technetium,  uranium,  neptunium,



 and plutonium,  and it also applies  to americium in some situations.







 B.I.2  Descriptions of Generic Repositories



     Our population risk assessment report (SMC 82)  and the Arthur D.



 Little study (ADL 79)  describe our  model  repositories for  five different



 geologic media—bedded salt, granite,  basalt,  shale,  and salt dome.   To



 demonstrate our analytical method,  we  will discuss only the bedded salt,



 granite, and basalt models,  since,  for our purposes,  the model shale



 repository  behaves  very  similarly to  basalt  and the  salt dome repository



 behaves  similarly  to the  bedded  salt.







     Each repository  is  2  kilometers wide and  4 kilometers  long.   About



 one-fourth  of the repository is mined  for the  waste,  the rest being  left



 as walls and pillars.  The mined portion is  5  meters  from floor  to roof.



After  the wastes have  been placed in the repository,  the mined areas are



 backfilled.  We assume that the backfill cannot  be refilled to the



 original density, which results in void space  amounting  to  20 percent  of



 its total volume, or 2,000,000 cubic meters.





                                   190

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     The model bedded salt repository (Figure 5-1 on page 89) is 460




meters (1500 feet) below the surface.  Above and below the repository




layer are 50 meters of salt, 50 meters of impermeable rock, and a 30-meter




thick, porous, water-bearing medium (aquifer).  The characteristics of the




overlying aquifer are shown in Table B-3.  There are 330 meters of




undefined sedimentary overburden between the top of the upper aquifer and




the surface.









     The granite  repository (Figure  5-2 on  page  90) is also  modeled as




460 meters below  the  surface.  The granite  formation  continues  downward




indefinitely,  so  there  is no  lower aquifer.  There  are 230 meters  of




granite above  the repository  and then an aquifer identical to  the  one




modeled above  the salt  repository.  Above the  aquifer are 200 meters of




overburden.









      The oasalt repository  (Figure B-l)  is  also  positioned at  460  meters




underground,  with 100 meters  of basalt both above and below.  The  basalt




is bounded  by overlying and underlying 30-meter  thick aquifers at  330  and




560 meters,  respectively.









      In granite and basalt, the void spaces in the backfill fill with




groundwater moving from the aquifer  through fractures in the bulk rock and




also  through the  sealed shafts and  boreholes.   We assume that all the




water becomes connected throughout  the mined volume and shafts and can be




considered  as one large volume.   In  the salt repository, the water may
                                    191

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                                                    -Surface
	Surface  	
 — Deposits	
                       	     	   	*iTf>
«* Aquifer                                          t^
                                                    460 meters
  FIGURE  B-l:  REFERENCE  REPOSITORY IN  BASALT
                            192

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                      Table B-3

               Aquifer  Characteristics
Distance from repository
    to overlying aquifer (meters)
Distance along aquifer to
    man's environment (meters)

Permeability in aquifer
    pathway (meters/year)

Transverse diffusivity (meters)

Gradient in aquifer pathway

Effective porosity in aquifer pathway

Thickness of aquifer (meters)

Width of aquifer  over repository  (meters)
 230.0  (granite)
 100.0  (bedded salt)
 100.0  (basalt)

1600.0
  31.5


   6.0

   0.01

   0.15

  30.0

4000.0
                            193

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enter through  imperfections in shaft and borehole seals.  Once  the salt




repository has been closed, the salt begins to  flow plastically under the




weight of the  overlying salt and the other geologic formations, so that




most of the void space in the backfill is eliminated.  Only a small amount




of the water may remain trapped in the salt in  the form of brine pockets.




We assume that each brine pocket is in contact  with two waste canisters



and contains 0.06 cubic meters of brine.









B.2  RISK ASSESSMENTS FOR POPULATIONS




     In this section we estimate the number of  serious health effects




(fatal cancers and genetic defects) that could  result from releases of




radionuclides from a repository.  The three objectives of our analysis




are:  (1) to estimate the probabilities of accidental events, (2) to




address the overall impact on man,  and (3) to provide information for




dec is ion-making.









     We considered mechanisms and initiating events that would release




radionuclides from the model repository and their subsequent transport




through the geosphere and the biosphere (SMC 82).  Releases were




categorized according to three general pathways to the biosphere: releases




to air,  releases  to land surface,  and releases  to aquifers (and ultimately




to surface water).   A number of initiating events were then identified




that would allow radionuclides to enter one or more of these media:
                                   194

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     1.   human intrusion (drilling for resources) — releases to land
         surface and to aquifers
           a.  direct hit on a canister
           b.  intrusion into the repository without hitting a canister

     2.   faulting — releases to aquifers
           a.  destruction  of canisters
           b.  disturbance  of repository  without  destroying canisters

     3.   breccia pipe formation — releases  to  aquifers.

     4.   volcanoes  — releases to air  and  land  surface

     5.   meteorite  impact — releases  to  air and land surface


     For each repository  type we examined a  number of pathways in the

biosphere through which  people may  be  exposed to radiation.  The pathways

for releases  to air were inhalation,  eating  contaminated foods,  exposure

to external radiation from contaminated air, and exposure to external

radiation from contaminated land surface. The same pathways apply to

releases to land surface since the nuclides  deposited on land surface will

be resuspended into the air.  We also calculated  the population risks from

use of the streams into which the aquifers discharge.  The pathways  for

releases to an aquifer and  subsequent discharge from a stream are drinking

water, eating fish, eating milk, meat, and  irrigated foods, breathing air

contaminated through resuspension, being  expoaured  to external radiation

from contaminated  air, and  being exposured  to  external radiation  from

contaminated land  surfaces.  We also  calculated risks  from eating  seafood

contaminated by entry of nuclides into the  ocean  from  the  rivers,  but

these risks were very small.  We calculated the fraction of each nuclide

ingested or  inhaled by humans for each subpathway and  then determined  the

resulting  doses to each of  10 organs:  ovaries, testes,  red marrow,
                                   195

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thyroid, lower gastrointestinal tract (LLI), lung, liver, kidney, bone,




and total body (considered representative of other soft tissue).  We used




the dose equivalent conversion factors (retn per curie) of INREM-II (KI 78,




DU 79) to calculate the organ doses.  The ovary and testes doses were




converted to genetic health effects, and the doses to the other organs




were converted to fatal cancers.  The genetic and fatal cancer




dose-to-health effect conversion factors are based on the work of the BEIR




Committee (NAS 72).  For all the situations we considered, the number of




potential genetic effects was always small relative to the projected




number of fatal cancers; therefore, this discussion concentrates on the




latter category of health effects.









B.2.1  Assessment Methodology




     The consequences we assessed for a population are the number of fatal




cancers produced by a release over the period of time considered.  We




examined the probability and consequences of each event occuring at a




number of different times.  The probabilities and consequences of all the




possible events and event times can be combined in two ways.   First, we




can present the results in a graph in which the ordinate of a point on the




graph gives the probability of obtaining at least the number of




consequences shown on the abscissa.  (For example, Figure B-2 illustrates




total fatal cancers caused by releases from a granite repository in the




first 10,000 years after repository sealing—it shows a probability greater




than 1 in 10 of a few hundred fatal cancers in this time and less than a




1 in 100,000 probability of much more than a thousand fatal cancers.)
                                   196

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                                             GRANITE
                                             REPOSITORY
              101      102      103      104      105
                   health effects over 10,000 years
10'
FIGURE B-2:  PROBABILITY OF POPULATION RISKS OVER 10,000 YEARS
                              197

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Alternatively, we can multiply each consequence by its probability and sum


the products, obtaining an expected value for the total number of fatal

cancers resulting from all events over the time period considered.

(Figure 5-3 on page 99 displays the sums of the product of the


probabilities and consequences for each type of event for the bedded salt

and granite repositories—a total for each type of repository is then

obtained by summing over all events.)  In the remainder of our discussion

of population risks, we will use the latter value to illustrate and

compare the harm associated with different characteristics of disposal

systems.




B.2.2  Releases

     Table B-4 indicates the major repository characteristics that control

the amount and rate of radionuclide release.  Releases to air and to land

surfaces are direct and rapid; the major controlling factor is the

fraction of the waste affected.  Releases through an aquifer are indirect


and slow; the major controlling factors are the fraction of the waste

affected and the rate of flow of contaminated water through the aquifer.

The flow rate depends on the permeability, porosity, cross-sectional area


of the flow path and the hydraulic gradient driving the flow.  In the

immediate vicinity of the repository, a major component in the hydraulic

gradient is the buoyancy resulting from heat produced by the waste; any

natural gradient between aquifers adds to the thermal buoyancy gradient.

The maximum value of the thermal buoyancy gradient is about 0.25, which

gradually decreases to zero as the waste cools (SMC 82).  Table B-4 gives


the particular characteristics associated with each of the release modes


we analyzed.
                                   198

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                                    Table B-4




                     Reference Case Characteristics (SMC 82)
Normal Groundwater Flow (to surface water via aquifer)
Permeability of host rock (m/yr)
Porosity of host rock
2
Cross-sectional flow area (m )
Hydraulic gradient

Drilling-direct hit of a canister (to land surface)
Fraction of repository inventory affected
Probability (yr )
Drilling-does not hit canister (to land surface)
Fraction of repository inventory affected
3
Volume of repository water involved (m )
3
Volume of repository water released (m )
Probability (yr )
Faulting (to surface water via aquifer)
Fraction of repository directly disrupted
Fraction of repository inventory affected
•3
Void volume of repository affected (m )
Permeability in fault zone flow path (m/yr)
Porosity in fault zone flow path
2
Cross-sectional area of fault zone (m )
Hydraulic gradient

Granite
3xlO~4
lxlO~4
8xl06
thermal

4xlO~6
2.5xlO~6
1.0
2xl06
200.0
0.0025

0.003
1.0
2xl06
3150.0
0.1
4000.0
thermal

Bedded Salt

—
—

4xlO~6
2xlO~5
6xlO~5
0.06
0.06
0.02

0.003
0.05
'IxlO5
31.5
0.1
4000.0
0.1 plus
thermal
Basalt
3xlO~3
lxlO~4
8xl06
0.1 plus
thermal
4xlO~6
lxlO~5
1.0
2xl06
200.0
0.01

0.003
1.0
2xl06
3150.0
0.1
4000.0
0.1 plus
thermal
Probability (yr  )
2x10
                                                         -8
                                                                 2x10
                -8
5x10
    -7
                                    199

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                                  Table B-4 (continued)
                         Reference Case Characteristics (SMC 82)
                                                        Granite
          Bedded Salt
             Basalt
Breccia Pipe  (to surface water via aquifer)
   Fraction of repository directly disrupted
   Fraction of repository inventory affected
   Permeability in breccia  zone  flow path  (ra/yr)
   Porosity in breccia zone  flow path
                                           2
   Cross-sectional area of  breccia zone  (m )
   Hydraulic  gradient

   Probability after first  500 yr (yr  )
             0.004
             0.016
            3150.0
             0.2
            3xl04
            0.1 plus
             thermal
            1x10
                -8
Volcano (to air and land surface)
   Fraction of repository inventory affected
   Probability (yr  )
 0.004
 0.004
 0.004
6X10'10
Meteorite (to air and land surface)
   Fraction of repository inventory affected
   Probability (yr  )
 0.001
    -11
4x10
 0.001
    -11
4x10
 0.001
4x10
                                       200

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     Normal Groundwater Flow




     The rock surrounding a granite or basalt repository is permeable, and




water moving through it could transport radioactive material to the




overlying aquifer.  The buoyancy effect resulting from the decay heat of




the waste drives water through rock fractures in accordance with the



permeability shown in Table B-4.  In this model, water flows upward into



the aquifer over the entire area of the repository and is replenished




through fractures in the rock around the repository.









     The release mechanism we considered for a bedded-salt repository is




more complicated.  Water may enter the repository from the upper aquifer




through imperfections in the sealed openings, especially the main  shaft




seal.  At the same time, plastic deformation of the salt caused by




lithostatic pressure from the overburden eliminates these void spaces.




This phenomenon tends to squeeze any water from the void spaces back  up




the shaft seals.  If the canisters have failed, the water squeezed back up




the shafts will contain some radionuclides from the waste.  Contaminated




water could then be released from the shafts into the overlying aquifer.




However, in the situations we felt were likely, too little water  leaked




into the repository to cause any releases (ADL  79, SMC  82).
                                   201

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     Drilling




     Human  intrusion  (drilling)  involves  either  a  direct  hit  on  the  waste




or  intrusion into any water contacting  the waste.  A  direct hit  results  in




transportation of a fraction of  the waste directly to the land surface and




opens direct communication between the  repository  and aquifer(s)  through




the borehole.  Release mechanisms for drilling intrusions into the water




are similar to those  for direct  hits on the waste.  For a granite or




basalt repository, the material  brought to the surface is a small fraction




of  the pool of contaminated water in the  entire  repository.   For  a




bedded-salt repository, we assume that  only one  of a  large number of small




brine pockets around  the waste would be affected.  We further assume that




intrusion into a brine pocket would cause ejection of the entire  contents




of  the pocket directly to the land surface because of the high lithostatic




pressure of the salt  (ADL 79).








     Faulting




     Fault movement can open a high-permeability pathway  for  release of




radioactive material to the overlying aquifer.  The heat  produced by the




decaying radioactive material in the repository  (thermal  buoyancy) then




drives the flow of water from the repository to  the aquifer.  In  bedded




salt and basalt,  the hydraulic gradient between  aquifers  also affects the




flow.   In granite,  recharge is from the surrounding bulk  rock; in salt-and




basalt,  recharge  proceeds principally through the  interconnection of the




lower  aquifer with the upper aquifer.
                                   202

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     If waste is directly disturbed by the fault, radionuclides from




damaged canisters can enter the water flowing to the aquifer at a rate




that may be limited either by the release rate from the waste matrix or by




the solubility of the radionuclides.  Also, if canisters fail for other




reasons—permitting radionuclides to leach or dissolve into the repository




water—and a fault occurs later, this contaminated water can be




transported to the overlying aquifer.








     Breccia Pipes




     A breccia pipe is a permeable formation resulting from the collapse




of rock into cavities formed by the dissolution of underlying  rock.




Breccia pipes might occur near bedded salt-repositories, but they are  not




relevant to the other geologic media we studied.  The release  mechanisms




to the overlying aquifer are the same as  in  faulting, but  the  flow  area  is




larger than the flow area of the fault, and  the  permeability of  the flow




path is greater than that for faulting.









     Meteorites and Volcanoes




     Meteorites and volcanoes that disrupt a repository  would  pulverize




parts of the repository and  its host  rock and would  eject  radioactive




material into the  atmosphere and onto the land  surface.  We assume  that




material ejected  into the air would  become uniformly mixed into  the




troposphere and would then  settle  onto  the surface  of the  earth.  Nuclides




would reach people through  the  pathways previously  described in




Section B.I.  Either type of event has  an extremely small  probability.
                                   203

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B.2.3  Reference Case Results




     Table B-5 presents the projected number of fatal cancers resulting




from the release of radionuclides by the release modes identified in




Table B-4.  For the bedded-salt repository, drilling events that miss the




canisters but hit a brine pocket constitute the dominant type of




initiating event, causing about 200 fatal cancers, primarily from releases




of araericium-241 (44 percent) and americium-243 (55 percent).  For the




granite repository, about 750 fatal cancers are projected in the first




10,000 years, due mainly to drilling into the repository water and to




releases due to normal groundwater flow.  The major causes of health




effects are, again, americium-241 (10 percent) and americium-243




(&9 percent).  For all model repositories the risks from faulting,




meteorites, volcanoes, and breccia pipes are insignificant.  The dominant




initiating event in the basalt repository is also drilling into the




contaminated water in the repository; this could result in 3,000 fatal




cancers over the 10,000 years, but releases from normal groundwater flow




would add another 1,400 projected effects.
                                   204

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                                  Table B-5

                Projected Population Risks Over 10,000 Years:

                              Reference Cases*



                                   Projected Health Effects
Repository Routine Drilling
Type Release Faulting (No hit) (Hit)
Granite 10 + 750 +
Bedded Salt 0 + 180 8
Basalt 1,400 3 3,000 2
Breccia Volcano;
Pipe Meteorite Total
+ 760
+ + 190
+ 4,400
* From Table 7-4, SMC 82

# "No hit" means the drill does not hit solid waste but only repository water
     while "hit" indicates the drill does hit solid waste.

 + = Less than 1 projected fatal cancer

— = not applicable
                                   205

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B.2.4  Sensitivity Analyses




     To this point, we have examined our reference cases for calculating




releases due to natural groundwater flow and several initiating events.




Using our assumed values, our risk assessment estimated numbers of health




effects for the reference cases.  However, for a broader insight into the




possible consequences of radionuclide releases to the biosphere, we have




varied the values of several parameters of the repository models.  These




variations involved changes in canister life, different release rates




(sometimes called "leach rates") from the waste form, different




retardation factors for radionuclides in the aquifer, and different




solubility limits of the radionuclides.









     Different Canister Life and Waste Form Characteristics




     We varied the values for canister failure time and waste matrix leach




rate from the reference repository characteristics.  The projected health




effects in Table B-6 show the effect of these changes.









     The primary effect of longer canister lifetimes is to lengthen the




period over which risks from human intrusion and normal groundwater flow




can be lowered or eliminated.  Over 10,000 years, canister lifetime has




much less effect on the health risk than does the waste-form release




rate.  There is a slight decrease in the risk value when increasing the




canister lifetime from the reference age to 1,000 years.  However, a




canister lifetime of 5,000 years, rather than 1,000 years, significantly




reduces the projected number of fatal cancers over 10,000 years.
                                   206

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                                Table B-6


              Projected Population Risks Over 10,000 Years:
           Various Canister Lives and Waste Form Release Rates
Canister Life

     Reference Case (100 years)
                    (500 years)

     1000 years

     5000 years

Waste Form Release Rate

     Reference Case
     (10-4 yr'1)

     High Estimate
     (10~2 yr"1)
     Low Estimate
     (10~6 yr'1)
                                       Projected Health Effects
                                    Granite    Bedded Salt  Basalt
  760
  575
  120
  760
2,500
   10
190
         4,400

 90      3,900

 40        180
190      4,400
 200      18,000
  50
50
 —  =  not  applicable
                                   207

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      Release-rate changes have much larger effects on the resulting risks


 than  canister  life changes,  and release-rate changes have more effect for


 the granite  and basalt  repositories than for the bedded-salt repository.


 The risks  for  the bedded-salt repository do not  increase much above those


 for a release  rate of  10   parts of remaining waste per year (ppy)


 because  of the solubility limits of the  americium isotopes.   Risks  for the


 granite  and  basalt repositories do not  increase  much as the  release rate

                   _3
 increases above 10  ppy  because there  is no more waste to dissolve


 during the 10,000 years.
     Finally, although not  shown  here,  studies  of simultaneously  varying


 the waste matrix  leach rate and canister  lifetimes reinforced the previous


 observation that  the  leach  rate has more  effect on the  overall risk than


 does the canister lifetime.




     Solubility of Radionuclides


     We examined  the model  repositories for unlimited solubility  of all


 the radionuclides so that entry of all nuclides into groundwater  would be


 limited only by the leach rate.  Table B-7 shows  that solubility  limits


 are extremely important in  limiting projected releases  from the model


 repositories and  resulting  fatalities, especially for bedded salt.   In the


model bedded salt repository, the most significant  solubility limit is


 that of the americium isotopes in brine pockets.
                                   208

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                               Table B-7
             Projected Population Risks Over 10,000 Years:
           Various Retardation Factors and Solubility Limits

                                  	Projected Health Effects
                                  Granite    Bedded Salt      Basalt
    Solubility Limits
        Reference Case *              760          190         4,400
        No Solubility Limits        2,800       14,700        12,600

    Retardation Factors
        Reference Case *              760          190         4,400
        All  Retardation Factors
          Equal to  1               38,000          200        5 million
* see Table B-2
                                   209

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      Retardation Factors  in the Aquifer




      Table B-7  also presents the  projected  health  effects  for  two




 different  choices of retardation  factors: our  reference  case assumptions




 about retardation factors  (see Table B-2),  and a case  that assumes that




 all  retardation factors are equal to one  (no retardation).  Lower




 retardation produces many  more health effects  from expected releases or




 from faulting at granite and basalt repositories.  Releases from bedded-




 salt repositories depend only slightly on retardation, because releases to




 an aquifer are  much smaller than  releases directly to  the  land surface.








      Genetic Health Effects




      In all types of repositories  analyzed, the number of projected




 genetic effects  are at least  one  order of magnitude less than the number




 of projected cancers, because most of the radionuclides  involved do not




 concentrate in  the  gonads.  For the reference  case release modes—over




 10,000 years following repository closure—we  project  15, 2, and 150




 genetic effects  from releases from granite, bedded salt, and basalt




 repositories, respectively. In comparison, we  project 760, 190, and 4,400




 fatal cancers for  these three repositories.









 B.2.5  Conclusions




     Several broad conclusions can be drawn from these performance




assessments.  First, major changes in the geochemistry at a site can




affect long-term risks much more than major changes in the engineered




barriers.   For example,  neglecting geochemical retardation for a granite
                                   210

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repository increases the consequences from about 800 health effects to




38,000.  In comparison,  assuming that the waste form dissolves very




quickly raises risks to a little more than 3,000, while assuming a zero




lifetime for the waste canister increases risks only to about 1,000.




Thus, it appears that efforts to identify a repository site with




appropriate characteristics can have greater benefits than efforts to



improve engineering controls.









     Second, comparing the two types of engineering controls, variations




of waste-form release rate consistently have more effect on  long-term




risks than variations of canister lifetime.  Improvements  in waste  form




appear to provide more benefits than improvements in waste canisters.








     Finally, good engineering controls, particularly good waste  forms,




can overcome poor site characteristics.  Our generic model of a basalt




repository assumes that relatively  large amounts of groundwater are




available to dissolve and  transport waste.  In  spite of  this disadvantage,




our basalt model can achieve risks  comparable  to those at  the  low end  of




the range for our granite  model  if  the waste  form used with  basalt is




about  an order  of magnitude  better  than  that  used with granite.
                                   211

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B.3  DOSE ASSESSMENT FOR INDIVIDUALS

     We assessed the annual dose equivalent (hereafter referred  to as

annual dose, for convenience) that could be received by  individuals who

breathe contaminated air or drink contaminated water from a well drilled

close to a repository.  These pathways are less complex  than  those in  the

population risk assessment because they do not include the ingestion of

foodstuffs or water from a river.  We used the model repositories and

geology described in Section B.I.2, and we examined several possible

pathways through which individuals could be exposed.  For each case, we

calculated the annual doses to individuals and, for those cases  involving

releases through a borehole, the extent (i.e., area) of  the aquifer that

would be contaminated as a result of the releases.  We chose  four of the

most significant scenarios to describe in this Appendix:

     1.  releases of solid waste to the land surface from striking waste
         while drilling;

     2.  releases to groundwater from striking contaminated repository
         water while drilling;

     3.  releases to groundwater from a fault movement that allows
         releases of water from within the repository, but that does not
         directly disrupt the waste and;

     4.  releases to the aquifer from natural groundwater flow.


     The probabilities and consequences of the four initiating events

differ markedly.  Over 10,000 years, it is quite likely  that  someone will

inadvertantly drill into a repository (assuming that institutional

controls have no effect beginning 100 years after disposal),  thereby

opening a small pathway through which relatively small quantities of

radionuclides could be released.  Fault movement, on the other hand, is
                                   212

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unlikely at a well-chosen site—but, if a fault should develop, it would



open a large pathway through which large quantities of radionuclides could




be released.  Finally, natural groundwater flow is inevitable in the




non-salt repositories beginning at the time of closure.  However, the




consequences will be small at an offsite well due to the slow movement of




the radionuclides.








B.3.1  Pathway Characteristics




     The preceding section described the events and processes we




considered—for the individual dose analyses—that may allow radionuclides




to reach an aquifer or the earth's  surface.  In this  section we  trace  the




pathways that those events and processes open from the waste to  an



individual.








     Geospheric Pathways




     Drilling.  Drilling  in  search  of  resources will  leave porous material




in  the  resulting  borehole through which contaminated  water from the




repository  can  flow.  This contaminated water flows upward to  the




overlying  aquifer mainly  because of the heat produced by radioactive




decay.   In  the  cases  of  the  basalt  and bedded-salt repositories, it is




assumed that the  underlying  aquifer is connected to the repository—




resulting  in a  hydraulic  gradient,  which (conservatively) is assumed  to




drive water to  the overlying aquifer.   Upon reaching the aquifer,  the




 contaminated water moves downstream from its point of origin,  the




 borehole.   Simultaneously, it very slowly spreads transversely across the
                                    213

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aquifer, contaminating a parabolically shaped  area.  At  some  future  time,




according to our model, an individual might unknowingly  drill  a well




within the contaminated area to obtain drinking water.   At  the time  the




person has ingested the water, the pathways caused by drilling are




complete.








     Faulting.  The faulting event is postulated to open communication




between the repository and an aquifer along a  line that  runs  the  full




length of the repository—involving a much larger amount of radionuclides




than drilling.  The faulting event does not release radionuclides to the




land surface, since we assume the implementing agencies  would  not permit




selection of a site where an aquifer is under  such pressure as to be




artesian.  (We also assume that the fault movement is much  too small to




transport radioactive waste directly to the surface.)









     We assumed the fault would create a 10-meter wide zone of high




permeability.  The fault could cross the aquifer in any  direction.  A




fault in the direction of the aquifer flow was chosen for analysis since




it would produce the highest individual doses  from faulting,  because waste




from many individual canisters along the fault would contribute to the




dose rather than just a few if the fault were oblique to the canister row.









     We considered two faulting scenarios.  In one, the  fault  strikes and




breaks a row of canisters;  in the other,  the canisters have failed before




the fault develops, and the contents of two or more rows of failed
                                   214

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canisters have combined in the porous volume of the repository backfill.




However, only the pathway from a fault intersecting repository water is




presented because these releases are much larger than releases restricted




to direct effects on canisters.  This pathway then terminates when the




contaminated aquifer water is ingested by someone.








     Normal Groundwater Flow.  We also analyzed the natural groundwater




flow pathway through the repository  to the  overlying aquifer.  We  looked




only at  the basalt repository, because this allowed us  to  estimate the




upper-bound dose rate  from natural groundwater  flow.  (The groundwater




flow rate  through the  basalt  repository  is  the  greatest normal  flow rate




associated with  any of the types  of  repositories  we analyzed.)   There  is




very  little water flow through  the  bedded-salt  repository since the




plasticity of  the salt allows it  to  seal cracks and  fissures  that may




develop.  In  the case  of  the  granite repository,  there  is no  lower aquifer




 to produce a  hydraulic gradient—so the  only driving  force to take




 nuclides to  the  overlying aquifer is the heat produced  by the decaying




 waste.









      Atmospheric Pathway




      This pathway is straightforward and also is a result of drilling.




 Material reaches the surface either as a solid (a core  sample)  or a liquid




 (contaminated water from the repository).
                                    215

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      The waste material brought to the surface may be removed by the




 drilling crew, it may be covered by other residue from the borehole, or it




 may lie at the top of the pile of residue.  The latter scenario provides




 the most conservative estimate and is the one we used.  The material is




 assumed to be a point source.  We did not calculate the direct radiation




 dose from this release,  which may be very large, because it depends on the




 way the material is handled,  on the length of time the drillers remain




 exposed, and on other factors.   We considered this incident to be an




 industrial accident not  pertinent to this work.  However, if the material




 becomes airborne it could be  spread beyond the drilling site and could




 eventually expose people in the general population.   In this analysis, the




 pathway ends when the material  is inhaled and has exposed an individual's




 lungs.   We also consider exposure from submersion in contaminated air and




 direct  exposure from the contaminated surface;  however,  these doses were




 found to be very  small compared to  the inhalation pathway and are not




 discussed in this Appendix.









 B.3.2   Dose Analysis




     A  computer code  called MAXDOSE-EPA was  used to  perform the individual




 dose analyses  (SE 81).   The results  of our study (GO 82)  are complex




 because  of  the  various pathways,  the  many  variations  of  the parameters,




 and the  number  of data points corresponding  to  the variety of years after




 the initiating  event  for which we analyzed.  Besides  this,  there  is much




data concerned with areas contaminated to  various  levels.   Rather than




attempt  to  summarize  all of these results  in this  section,  we present




results  for only those events that occur 1,000 years after  the repository






                                   216

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is sealed (except for shaft seal leakage that begins at 500 years, when




the canisters fail).  The method used to calculate the annual individual




dose rates leads to a rate equal to the exposure in the 50th year of




chronic exposure, which is how all our individual dose equivalent rates




should be interpreted.  Chapter 5 of Goldin, et al., (GO 82) presents the




entire dose rate and contaminated land area results for the granite and




bedded-salt repositories.









     A drill that hits solid waste brings 15 percent of the waste of one




canister to the  surface.  The largest dose  rate  is  incurred by breathing




contaminated air close to the drilling site  (20 meters) 10 years  after  the




event.  The dose rate is about  11 rera per year (rem/yr) to  the lungs,




mostly from americium-241.  At  longer times  after  the  drilling,




plutoniunr-239 and -240 become dominant as the dose rates  fall from about




4 rem/yr at  1,000 years  after the event  to  about 20 mrem/yr  at 10,000




years.








      The  same direct-hit drilling  event  will also allow nuclides to be




 released  to  the aquifer  via the borehole.   The  largest dose rate from




 drinking  contaminated aquifer water is  about 600 rem/yr to red  bone




 marrow,  again 20 meters  from the original  point of release but  1,000 years




 after the event.  No areas are  contaminated enough to give more  than 0.5




 rem/yr until 1,000 years when  americium-241 has traveled the 20-meter




 distance and contaminated 0.33 hectares (3,300 square meters).   The




 contaminated area then increases to 6 hectares as  the nuclides spread  and,




 after 10,000 years, decreases because of radioactive decay.






                                    217

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     Releases caused by a fault intersecting a granite repository are—as




would be expected—much larger than those associated with drilling.  The




doses range from 40 rem/yr at 10 years and 20 meters from the repository




to 3.4x10  rem/yr at 100 year.s and 20 meters.  The latter dose rate




slowly decreases after 100 years to 2,600 rem/yr at 10,000 years.




Tin-126 dominates the dose rate for 20 years following the faulting event;




americium-241 then dominates until 5,000 years, when americium-243 takes




over.  At 10,000 years, plutonium-239 delivers the greatest doses in this




scenario—which is very unlikely to occur (see section B.3.3).









     Doses from natural groundwater flow are not considered in




MAXDOSE-EPA.  Instead, we used an equation from section 4.1.1 of GO 82 to




describe the release rate from the repository, and then we used




equation 3-7 from the population risk assessment report (SMC 82) to relate




the radionuclide flow rate from the repository to the nuclide flow rate at




the well used for drinking water.  We then related this flow rate to a




radionuclide concentration and converted the concentration to a dose rate,




using the dose conversion factors found in Table 3-4 of GO 82.  The




results of this analysis are presented in Table B-8.  Within 100 meters of




the repository there are potentially high dose rates.  However, beyond 100




meters the potential annual doses are, with one exception, below 1 rem per




year.  The dominant nuclide close to the repository is americum-243 while




cesium-135 generally dominates at greater distances—this is indicative of




the longer half-life and greater mobility of cesium-135.
                                   218

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                         Table B-8

 Dose Equivalent Rates (rem/yr) from Drinking Groundwater
          Contaminated by Normal  Groundwater  Flow

Through a
Basalt Repository
Years after Closure
Distance from
Repository (m)
20
100
1000
2000
4000
8000
10,000
15,000
20,000
1000
1.5E+1*
94%
Sn-126
1.5E+1
94%
Sn-126
9.8E-1
74%
Cs-135
O.OE+0
0%
None
0 . OE+0
0%
None
O.OE+0
0%
None
O.OE+0
0%
None
O.OE+0
0%
None
O.OE+0
0%
None
2500
1.5E+5
68%
Am-243
1.3E+1
94%
Sn-126
8.1E-1
78%
Cs-135
8.0E-1
78%
Cs-135
7.8E-1
80%
Cs-135
O.OE+0
0%
None
O.OE+0
0%
None
O.OE+0
0%
None
O.OE+0
0%
None
5000
8.5E+4
75%
Am-243
l.OE+1
94%
Sn-126
5.9E-1
82%
Cs-135
5.9E-1
83%
Cs-135
5.8E-1
84%
Cs-135
5.6E-1
87%
Cs-135
O.OE+0
0%
None
O.OE+0
0%
None
O.OE+0
0%
None
10,000
3.4E+4
72%
Am-243
2.5E+4
70%
Am- 14 3
5.7E+0
94%
Sn-126
3.3E-1
89%
Cs-135
3.3E-1
90%
Cs-135
3.2E-1
92%
Cs-135
3.2E-1
92%
Cs-135
3.1E-1
94%
Cs-135
O.OE+0
0%
None
*1.5E+1 = 15 rem/yr
  94%   = 94% of the 15
 Sn-126.
rem/yr is from
                            219

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B.3.3  Probabilities of Failure Modes


     In this section we provide information useful for considering the


probabilities of some of the potential individual doses just discussed.


It should be noted, however, that the following probabilities of



inadvertant human intrusion assume that all institutional controls—both


active and passive—will be ineffectual beginning 100 years after


disposal.  Thus, we believe that these probabilities for human intrusion



may be conservatively high.






     Direct Hit on Waste


     From the information given in the ADL report (ADL 79), the probability


of a drill hitting a waste canister is estimated to be 2x10   hits per



year in a bedded-salt repository, 3x10   hits per year in a granite


repository, and 1x10   hits per year in a basalt repository.






     Drilling into Contaminated Repository Water


     According to the ADL study (ADL 79), the frequency, and therefore  the



probability, of drilling into the granite repository is once every 400


years, or 2x10   per year.  The probability of drilling into a basalt


                           -2
repository is given as 1x10   per year.  The probability of drilling



into a brine pocket in a salt repository may be conservatively estimated



to be the same as the drilling frequency into any part of a salt


repository—even though it would likely be lower because brine pockets  do


not exist throughout the repository.  However, using the more conservative


approach, ADL estimated 21 drilling attempts in the first 1,000 years—

                                      _2
which results in a probability of 2x10   hits per year.
                                   220

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     Fault Movement




     ADL (ADL 79) sets the probability of a fault occurring as 2xlO~8




per year for bedded salt and granite and 5xlO~7 per year for basalt.




However, we believe that only 10 percent of the faults would be lengthwise




(i.e., parallel or within 18 degrees of the longest repository axis).




These would produce the effects studied here, but there is a 90 percent




probability that the effects would be less than those given here.









B.3.4  Probability of Exposure




     Air




     The area around the point where the waste material is brought to the




surface following a direct hit on a waste canister—where the air and




ground are contaminated enough to result in a level of 500 mrem/yr—was




found to have a radius of 170 meters.  Since the radioactivity persists




for hundreds or thousands of years and the radioactive material is assumed




not to be reburied, it is virtually certain that someone would eventually




encounter some small level of exposure.








     Water




     The probability of a person encountering a contaminated area of  the




aquifer can be estimated by the probability of drilling for drinking  water




in the contaminated area, the area of the contamination, and the length of




time the area will remain contaminated.  There is no  simple way to  display




the many results of these calculations; however, an example will




illustrate the procedure.
                                   221

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     As was discussed in Section 5.3 of this document, the average well


                                               -4
drilling density in the United States is 6.4x10   well per hectare per



century.  Through calculations performed by the computer code MAXDOSE-EPA,



it was found that the area contaminated above 500 mrem/yr at 10,000 years



is 8 hectares.   The probability of a well being drilled into the



contaminated area is then (6.4x10  ) times 8 — or about 5x10   per



year.
B.3.5  Conclusion



     Considering the failure model probabilities and the associated dose



equivalents, we conclude that the chance of individuals receiving large



doses from disruptive events in a repository is small.
                                   222

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